pu consumption in advanced light water reactors
TRANSCRIPT
eGENuclear Lnergy
AdvancedReactorProgramsGeneralElecmcCompany6835ViaDe/OroM/CSanJose,CA95/I9.I315408365.6600
XL-P2A37-94003 January 15, 1994
U.S. Departmentof EnergyDOE San FranciscoOperationsOfficeNuclear Division1301 Clay Street, Room 700NOakland, CA 94612-5203
Attention: Kashmira Mall
Subject: Contract No. DE-AC03-93SF19681"Pu Consumptionin Advanced Light Water Reactors"Transmittal of Phase 1C Report
Enclosedis the GE Nuclear Energy Phase 1C Report "NEDO-32314 - Study of PuConsumptionin Advanced Light Water Reactors; Evaluationof GE's AdvancedBoiling Water Reactor Plants, Compilationof Phase 1C Task Reports'.
If there are any questions, please contact me at (408) 365-6468 or(408) 925-1714.
Edward EhrlichProject ManagerGE ALWR Pu ConsumptionStudy
XL-P2A37-94003 2 January 15, 1994
Attachment: GE Nuclear Energy Report NEDO-32314 - Study of PuConsumptionin Advanced Light Water Reactors; Evaluation ofGE's Advanced BoilingWater Reactor Plants, Compilation ofPhase 1C Task Reports,dated September 15, 1994
cc: Aundra RichardsDOE-SANDave PinesDOE-SANRobertNeuholdDOE-HQ NE-45ErnieCondonDOE-HQ NE-45
NEDO-32314RFP DE-AC03-93SF19681
January15, 1994
@GENuclear Energy
i
SanJose, Ca/ifornia
Study of Pu Consumption inAdvanced Light Water ReactorsEvaluation of GE Advanced Boiling Water Reactor Plants
Compilation of Phase 1C Task Re _orts
NEDO-32314RFP DE-AC03-93-SF19681
JANUARY 1994
TITLE: STUDY OF Pu CONSUMPTION IN LIGHT WATER REACTORSEvaluation of GE Advanced Boiling Water Reactor PlantsCompilation of Phase 1C Task reports
Preparedfor theUnited States Departmentof Energy
UnderContractNo. DE-ACO3-93SF19681
• MAS]ERGENuclearEnergy
AdvancedReactorProgramsSanJose,Ca/ifomia95119-7315
i_TRlt_JTION OF ]'HIS DOCUMENT 1_ _jNL.1Mrl F._L_'_93-426..02
DISCLAIMER
This reportwas preparedas an accountof work sponsoredby an agency of the UnitedStatesGovemment. Neitherthe United StatesGovernmentnor any agencythereof,norany of their employees,norany of theircontractors,subcontractors,or theiremployeesmakesany warranty,expressor implied,or assumesany legal liabilityor responsibilityforthe accuracy, completenessor usefulnessof any information,apparatus, product orprocessdisclosed,or representsthat its use wouldnot infringeprivatelyowned rights.Referencehereinto any specificcommercialproduct,process,or serviceby trade name,trademark, manufacturer, or otherwise, does not necessarily constitute or imply itsendorsement, recommendation,or favoring by the United States Govemment or anyagencythereof.The viewsand opinionsof authorsexpressedhereindo not necessarilystateor reflectthoseof the UnitedStatesGovernmentorany agencythereof.
Study of Pu Consumption in Advanced Light Water Reactors
Evaluation of GE-ABWR
TABLE OF CONTENTS FOR PHASE 1C REPORT
Phase IC WBS
SUMMARY REPORT OF PHASE 1C EVALUATIONS
1.0 CORE AND SYSTEM PERFORMANCE
1.1 Reference Spent Fuel Design 1.1-11.1.1 Normal Operation 2.1 1.1.1-11.1.2 Transient Response of Reference Fuel Design 2.2 1.1.2-11.1.3 Fuel Characteristics after Irradiation 1.2 1.1.3-1
1.2 Alternate Core Designs for Pu Disposition 2.1 1.2-11.2.1 Alternatives for 100 MT in 25 Years 2.1 1.2-21.2.2 Altematives for 50 - 100 MT in 25 Years 2.1 1.2-51.2.3 Alternatives for 100 MT in more than 25 Years 2.1 1.2-7
1.3 Relationship between Pu Enrichment, Discharge Exposure, 1.3 1.3-1Disposition Time, Isotopics and Number of Reactors
2.0 FUEL CYCLE
2.1 MOX Fuel Fabrication Requirements for Various Spent 1.7 2.1-1Fuel Scenarios
2.2 MOX Fuel Handling and Disposal2.2.1 Criticality Analyses for Storage, Handling & 1.3 2.2.1-1
Repository2.2.2 Spent Fuel Disposition in Repository 1.4 2.2.2-12.2.3 Spent Fuel Proliferation Resistance 1.4 2.2.3-
2.3 Qualifying and Licensing MOX Fuel2.3.1 Review of MOX Fuel Licensability 1.6 2.3-12.3.2 Program Plan for Lead Fuel Testing 1.6 2.3-112.3.3 US Infrastructure for Lead Fuel Testing 1.5 2.3-202.3.4 European Infrastructure for MOX Testing 4.6 2.3.4-1
Table of Contents for Phase 1C Report (Continued)
Phase IC WBS
2.0 (Continued)
2.4 MOX Fuel FabricationFacility Requirements 1.72.4.1 Process Simulation 1.7 2.4.1-12.4.2 Weapons Pu Interface: Input Pu Specifications 1.1 2.4.2-12.4.3 Layout, Cost, Schedule, Rate of Pu Processing 1.7 2.4.3-12.4.4 First-of-a-Kind Technologies 1.8 2.4.4-1
2.5 Waste Stream Characterization/Management 1.9 2.5-1
3.0 TRITIUM PRODUCTION
3.1 MOX Core Design for Tritium Production 2.1 3.1-1
3.2 Tritium Target Design and Performance 3.4 3.2-1
3.3 Tritium Target Fabrication and Recovery Facility Requirements 3.1 3.3-1
3.4 ABWR Plant Operations for Tritium Production 3.3 3.4-1
4.0 INFRASTRUCTURE AND DEPLOYMENT
4.1 Non-U.S. Facilities Technology Evaluation4.1.1 Japanese MOX Fabrication Facilities 4.6 4.1.1-14.1.2 BNFL Facilities and Experience 4.6 4.1.2-14.1.3 Comparison of U.S. and Foreign (UK) MOX Fuel 4.6 4.1.3-1
Fabrication Facility Regulatory Requirements
4.2 Adapting Commercial MOX Fuel Fabrication Experience 4.6 4.2-1
4.3 Pu Disposition Complex Infrastructure 4.1,4.2 4.3-1
4.4 Transportation Infrastructure4.4.1 Transport Logistics for Tritium Production 3.1 4.4.1-14.4.2 Transportation of Plutonium Materials for MOX 1.10,4.3 4.4.2-1
Fabrication Facility4.4.3 Transportation of Nuclear Waste 4.4 4.4.3-14.4.4 Spent Fuel Transportation and Logistics 4.5 4.4.4-14.4.5 Comparison of U.S. and International Transport 1.10 4.4.5-1
Regulations
Table of Contents for Phase 1C Report (Continued)
Phase IC WBS
5.0 SAFETY AND ENVIRONMENTAL APPROVAL
5.1 Pu Disposition Complex Safety Approval with 6.2 5.1-1Tritium Production
5.2 Impact of Tritium Production on Environmental 6.2 5.2-1Approval
5.3 ABWR Disposition Complex Safety Approval 6.2 5.3-1Program
5.4 Environmental Permitting Plan and Schedule 6.2 5.4-1
6.0 DEPLOYMENT REQUIREMENTS
6.1 Development Requirements Overview 5.1 6.1-16.2 Development Requirements for MOX Factory 5.2 6.2-16.3 Development Requirements for Tritium Production 3.2,5.3 6.3-1
7.0 SAFEGUARDS AND SECURITY
7.1 Safeguards Requirements for Pu Transport 1.10 7.1.1-1
8.0 COST AND SCHEDULE
8.1 Cost and Schedule Analysis 6.1 8.1-1
Appendix A: Compliance of MOX Fueled GE9 Assembly with Amendment 22 of A-1NEDE-24011-P-A (GESTAR II)
Appendix B: Repository Considerations B-1
Appendix C: T2P2: A Computer Program for Estimating Tritium Target Performance C-1and Tritium Environmental Source Terms
Appendix D: Radiological Safety Requirements and Criteria for the Sellafield MOX D-1Plant
SUMMARY REPORT OF PHASE 1C EVALUATIONS
Contract No. DE-ACO3-93SFI9681, "Pu Consumption in Advanced Light Water Reactors"
The evaluations conducted during Phase 1C of the Pu Disposition Study have provided further
results which reinforce the conclusions reached during Phase 1A & 1B:
• 3E's ABWR was designed for a full core loading of MOX fuel and requires no reactormodifications or plant systems level changes to use MOX fuel.
• The ABWR design allows a wide flexibility in full MOX core design options to meeta wide range of disposition objectives.
• The technology for converting weapons Pu to MOX fuel has already beendemonstrated by DOE complex activities and by commercial operations.
• Existing DOE facilities can be adapted to building the MOX factory.
• No fundamental technical issues relative to nuclear safety, worker safety, publichealth and environmental impact or licensing have been identified which wouldpotentially delay either a MOX plant or reactor construction or startup.
• Institutional organizations and criteria which are needed to implement the MOX plantcan be established to support the project schedule.
• The infrastructure exists for near term (8-10 years) deployment in the U.S. of theABWR Plutonium Disposition Complex.
These conclusions clearly establish the benefits of the fission option and the use of the ABWR
as a reliable, proven, well-defined and cost-effective means available to disposition the weapons
Pu. This project could be implemented in the near-term at a cost and on a schedule being
validated by reactor plants currently under construction in Japan and by cost and schedule
history and validated plans for MOX plants in Europe.
Evaluations conducted during this phase have established that (1) the MOX fuel is licensable
based on existing criteria for new fuel with limited lead fuel rod testingl (2) that the applicable
requirements for transport, handling and repository storage can be met, and (3) that all the
applicable safeguards criteria can be met.
During this phase, visits were made to DOE's Complex 21 sites to assess the existing
infrastructure that might support the disposition process. Contact was established with LLNL
and LANL to determine the technical capabilities and interfaces that might be implemented at the
front-end of the MOX factory. Transportation requirements and infrastructure for all aspects of
the disposition process from fresh fuel to spent fuel were defined and evaluated. The
infrastructure for carrying out this disposition - management and staff, facilities, and procedures -all exist.
Evaluation of the capability to produce contract quantities of tritium was demonstrated in Phase
1A using a conventional urania fueled core. During this phase, a MOX core design was
developed and analyzed which shows that the same requirements can be safely, met with a MOX
fueled core such that tritium production and Pu disposition can be carried out concurrently.
Cost and schedule evaluatiops are ongoing and additional data that have been collected are
consistent with previously reported cost and schedule estimates.
The individual task results are summarized below.
CORE AND SYSTEM PERFORMANCE
A nominal candidate spent fuel MOX core design has an average Pu enrichment of 3.5% and an
average discharge exposure level of 37,000 MWD/MT(typical of the upper range of discharge
exposure for GE current 8x8 ABWR fuel design). Neutronics and safety analysis show large
margins to safety limits with no reactor system changes and no core uncovering under
accident conditions. For the baseline 25 year project term case, this design requires up to 6
reactors for dispositioning 100MT of Pu. For a disposition time of 60 years (the design lifetime
of the ABWR), only 2 reactors are required. For a disposition campaign of 40 years - the current
license term without relicensing - either the disposition amount could be lowered to 75 MT or a
third reactor added for a full 100 MT campaign.
With t,_e flexible capabilities of the ABWR, alternate core design options are available which
permit the plutonium to be dispositioned using fewer reactors. Options include discharging the
fuel at a slightly lower exposure or increasing the plutonium enrichment. It is possible to
disposition 100 MT of plutonium with two reactors in 36 years in a core design with 5%
enrichment and 37,000 MWD/MT burnup. Another core design option requires only one reactor
in a 54 year campaign to disposition the same amount. This option is a core design of 5%
enrichment and 30,000 MWD/MT burnup which produces discharge isotopics and bundle
radiation levels comparable to the current average BWR discharge exposure.
Explicit relationships between these variables, in particular the effect of disposition time, can be
seen in the figure below and are discussed further in the report. The cost tradeoffs for these
options are continuing. It should be noted that generically, for higher enrichment LWR designs,
licensing delays and technical development risks are increased.
100 Mt of Pu at 40 GWDIMt
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FUEL CYCLE
MOX Throughput Requirements
The mixed oxide fuel fabrication throughput requirements have been established for four spent
fuel scenarios for a 3.5% Pu enriched MOX fuel that is discharged at 38,000 MWD/MT. The
four scenarios with the resulting fabrication requirements are shown on the accompanying table:
Soent Fuel Scenari9,
Characteristic 1 2 3 4
Available Plutonium, (MT) 100 100 50 50
Time to Disposition, (Yrs.) 25 60* 25 60*
ABWRs required, (Number) 6 2 3 1
Fabrication Plant Design:
MOX Throughput, (MT/Yr.) 220 80 110 40
*ABWR design lifetime
These throughput requirements include a 15% excess production capability margin. The MOX
throughput requirements for the alternate core designs using 5% Pu enrichment are currently
being evaluated.
MOX Fuel Licens#Lg Evahtation and Lead Fuel Testing
GE's new fuel designs are licensed using a process agreed upon between GE and NRC which
requires GE to examine specific documented criteria in detail. The criteria of this document were
considered in detail and indicate that the available data base is sufficient for a preliminary license.
This evaluation included a consideration of all salient MOX fuel properties, in particular as to
how they differ from the properties of urania fuel and the impact of this difference on fuel
performance and licensing criteria. The sole remaining issue concerns the verification and
qualification of a fabrication process to demonstrate that MOX fuel with an acceptable
microstracture can be made and that it performs as expected in in-reactor experiments. No need
for full scale assembly tests was identified, in particular because prototypical fuel designs are
unlikely to be compatible with a predominantly urania core and also because such isolated
assemblies do not provide system level response information. Nevertheless, if there is a need to
irradiate full scale MOX assemblies, the infrastructure for doing such testing is readily available.
A short full MOX core confirmatory test activity is already reflected in the Pu Disposition
Project schedule.
MQX Fuel Hcmdlmg and Disposal
The work scope elements during Phase 1C called for an examination of the spent nuclear fuel
(SNF) characteristics relative to proliferation resistance, handling, storage and repository
requirements. Although the specific requirements are yet to be fully enunciated for some of these
categories, the evaluations show that the requirements which exist at this time can be met. In
particular, it was found that the proliferation resistance as measured by the attractiveness level of
the weapons plutonium is degraded with each process step even prior to irradiation as fuel, and
that the fission option proposed here has better and proven proliferation resistance compared to
other non-fission disposition options. For handling, storage and disposal, all applicaLle criteria, in
particular subcriticality requirements, are met while employing existing technology and
configurations used for commercial SNF.
MOX Fuel Fabrication Facility Reqmrements
A systems analysis for establishing the requirements of a mixed oxide fuel fabrication plant that
will provide the capability needed for the plutonium disposition mission vvas carried out. The
analysis approach includes ,_ dynamic simulation of the entire disposition process from the
accumulation of excess plutonium metal to the final disposal of the spent fuel in long term
storage. All system interfacing requirements among the storage facility, plutonium shipping, fuel
fabrication, fresh fuel shipping, reactor operation, spent fuel shipping and final long term storage
can be evaluated. A detailed process simulation model of the fuel fabrication plant was
developed to establish process performance requirements. The dynamic analysis was planned to
provide insight on the optimum strategies for fuel fabrication and to minimize life-cycle waste
accumulation. This dynamic model is currently being evaluated for consistency of numerical
results. Further model development could include elements for establishing the requirements for
real-time material accounting and the requirements for minimizing radiation exposure of workers.
This system analysis approach and dynamic modeling could be implemented in the project phase
to optimize the details of the various activities.
TRITIUM PRODUCTION
The core design for tritium production using conventional urania fuel was presented in the Phase
1A report. During this phase of the study, a MOX fueled core has been designed that meets the
tritium production requirements. This core has a core-average Pu enrichment of 5.9%, uses four
tritium target rods per assembly as before, and utilizes the MOX fuel to an exposure of 28,000
MWD/MT. All the nuclear and thermo-mechanical design criteria for normal operation have been
met. To meet a requirement to design within the already existing target rod database, it would be
necessary to discharge all the target rods every year. An alternative would be to extend the
irradiation data base which would allow the target goals to be met with a variety of fuel cycles of
longer duration which would be more compatible with commercially attractive electricity
production fuel cycles of 18-months or partial core reloads with more frequent refueling.
IN RASTRUCFURE AND DEPLOYMENT
Planned MOX Fuel Fabrication Facifities in Foreitm Ctnmtries
As part of the infrastructure evaluations, the MOX fuel fabrication capability in Japan and
England was considered with a view to technology assessment. The Japanese MOX program is
still in a planning stage. It calls for reprocessing of LWR spent fuel and production of about 100
MT of MOX fuel per year by the year 2002. A reprocessing facility is under construction in
Rokkasho Mura. The MOX fuel fabrication factory is currently in the,design phase. The
planned MOX facility will be fully automated, and is also being designed to accommodate non-
remote maintenance, if this is required. The input feed material to this plant is planned to be a
50-50 master blend of urania and plutonia powder. Transportation casks for international
shipments have been designed and fabricated and are in use. Casks for local shipping are
currently being designed. It is believed that safeguards will be implemented primarily through
IAEA inspection and standards.
A detailed description of the planned MOX fuel facility of BNFL at Sellafield which will use
plutonium from reprocessed commercial SNF is provided in this report. In addition to the
facility descriptions which indicate that the technology is ready and is easily adapted to
dispositioning weapons plutonium, a comparative evaluation of the licensing and safety
requirements for the Sellafield plant vs. a Greenfield facility in the U.S. has also been provided.
Ada_vtit_ Exist#Tg MQX Fuel Fabrication Technology
Although MOX fuel has not been fabricated in significant quantities in the U.S. in nearly two
decades, other countries are proceeding to implement plans for fabrication of MOX fuel from
reprocessed plutonium. Since this technology exists and is being implemented, the changes
needed to adapt this technology for processing weapons plutonium were defined. In most of the
areas such as shielding, worker exposure, handling, and maintainability, weapons plutonium
would be easier to process than reprocessed plutonium. Two areas were identified which require
additional evaluations in adapting this already available technology: the first concerns a re-
evaluation of the constraints - in the form of allowable qaantities of plutonium in any given
area/container- arising from criticality requirements and its effect on throughput, and the second,
surveillance/accountability instrumentation based on gamma activity which may need
modification or alternate instrumentation, as the weapons plutonium g activity is far lower than
in the case of reprocessed plutonium.
Plutonium Disposition Infrastructure in the United States
It was concluded in Phase 1A that the infrastructure appeared to be established for deployment
of an ABWR Pu Disposition Complex in the United States. The brief surveys of each of the
DOE sites conducted under Phase 1C support this conclusion.
The Department of Energy (DOE) already has thesites and capabilities, and the flexibility with
these sites and capabilities, to deploy an electric power producing, full MOX-fueled ABWR Pu
Disposition Complex. For study comparison purposes, the reference case for deployment of the
Plutonium Disposition Complex in the United States is a new "Greenfield," in which facilities are
constructed and located all together on a hypothetical site at Kenosha, Wisconsin. The purpose
of the infrastructure portion of this study was to determine the extent to which the "complex"
could utilize the existing capabilities at one or more of the existing DOE and/or commercial sites.
The study included visits and a collection of data for the following sites:
• Idaho National Engineering Laboratory (INEL)• Nevada Test Site
• Oak Ridge Reservation (ORR)• Pantex Plant
• Savannah River Site (SRS)° Hanford Site
• Lawrence Livermore National Laboratory (LLNL)° Los Alamos National Laboratory (LANL)
It is clear that considerable cost effective, installed capability is available within the DOE
community now for meeting the Pu disposition needs in the near term with one or more electric
power producing, full MOX-fueled ABWR plants. These capabilities can be implemented in the
short term with effort ranging from minor refurbishing to upgrading of existing facilities, with
only a few requirements, such as the reactor, being Greenfield efforts at all sites. It is anticipated
that a minimum cost deployment will be to locate the entire Pu Disposition Complex at one site.
SRS, ORR and INEL already have in place significant applicable elements. It is also possible to
take advantage of unique capabilities which exist at these and other sites and create a distributed
"complex," with some additional cost for transportation between sites.
WASTE CHARACTERIZATION AND TREATMENT
Waste stream characteristics were updated where new information modified previous estimates.
Treatment and disposal options were also identified for each waste stream. It was concluded that
existing and planned treatment and disposal technology will be adequate, and that the types and
quantities of wastes generated are typical of normal reactor operations.
SAFETY AND ENVIRONMENTAL APPROVAL
The safety approvals and environmental permitting activities required for the ABWR Plutonium
Disposition Complex were examined. The New Production Reactor (NPR) safety approval was
used as a model for the development of a detailed ABWR Pu Disposition Program safety
approval schedule. This schedule assumed that a single Integrated Safety Analysis Report
(ISAR) will be submitted in stages to DOE and supporting government agencies for approval to
support critical program decision points. The environmental permitting process required for a
Record of Decision (ROD) was also examined for both a Greenfield site and an existing DOE site.
SAFEGUARDS AND SECURITY
Work continued on assessing the impacts of safeguards and security requirements for plutonium,
tritium and enriched lithium on the configuration and operation of the ABWR Plutonium
Disposition Complex. The results of these additional studies indicate that because the number of
shipments is relatively small ( -- 1 fresh fuel shipment/month for the 2 reactor case and less for
enriched lithium or tritium), transportation of controlled materials is unlikely to be a controlling
factor in the configuration of the complex. However, as discussed in Phase 1A, maintaining the
uncertainty in material accountability at acceptable levels must be addressed given the large
quantity of plutonium to be processed.
COST AND SCHEDULE
Assuming a national commitment, and the application and use of existing ABWR submittals,
existing DOE site-specific environmental data and procedures, and the use of existing DOE and
MOX fuel fabrication technology, it is concluded that the overall ABWR Pu Disposition
Complex schedule issued in the ABWR Phase 1A report can be achieved. Further refinement of
the ABWR Plutonium Disposition Complex costs and schedules was initiated. The structures
and improvements account (EEDB account 21) of the baseline ABWR capital cost estimate was
reviewed against current information available from construction of the two ABWRs in progress
in Japan and similar cost studies done for the GE Simplified Boiling Water Reactor (SBWR).
New cost estimates, cash flows and schedules are being developed for disposition of 50 or 100
MT of plutonium in 25 years, 40 years and for the reactor lifetime of 60 years. Development of
preliminary revenue calculations and review of cost and schedule data from an existing European
MOX fuel fabrication facility owner/operator was begun. Verification of environmental and
safety approval schedules was completed. Evaluation of the cost tradeoff from the use of
existing facilities at DOE sites vs a Greenfield site was initiated.
1. CORE AND SYSTEM PERFORMANCE
GE's 1300 MWe Advanced Boiling Water Reactor (ABWR) has been designed to utilize full
core loading of Mixed Uranium-Plutonium Oxide (MOX) fuel. This design characteristic and
large power rating allows the ABWR a broad array of core design options to optimally dispose
of weapons plutonium. Whether the objective is to maximize the return on investment or to
minimize initial capital cost, the ABWR with its well established cost and schedule data base,
provides the best means of LWR fission based disposition.
Phase 1A of this study, concluded in May 1993, presented three disposition alternatives.
Subsequent DOE reviews have indicated that the spent fuel option presented the most optimal
means of disposition. Core design efforts during Phase 1C have been focused on this option.
Specifically, the following Agreements and Commitments were issued:
(a) GE to focus on core design options that will result in the fuel being ,lischarged as
spent fuel with isotopic compositions typical of current BWR commercial spent
fuel.
(b) GE to evaluate disposition options where the reference amount of Pu to be
dispositioned will remain at 100MT and the disposition time is limited to 25
years, but information may also be provided for a range of amounts from 50 MTto 100 MT.
(c) GE to evaluate disposition options for the reference amount of 100 MT of Pu and
disposition time of 25 years, but information may also be provided for mission
duration up to the design lifetimes of the reactors. Emphasis will be on 40 year
design life, but 60 year case can also be considered.
The first subsection following describes the Reference Spent Fuel Case. Both neutronic and
transient results are given for this case. The next section shows various spent fuel alternatives
for differing dispositon goals. The final section has two parts. First, a generic discussion is
presented on the relationship between the quantity of Pu that has to be disposed, the disposition
time, Pu enrichment and exposure. Following this, the relationship between spent fuel isotopics,
Pu enrichment and exposure is examined.
1.1.0-1
1.1 REFERENCE SPENT FUEL DESIGN
The general approachutilized in the core and fuel nuclear design for the Reference Plutonium
Spent Fuel option was to establish the bundle average enrichment to enable the fuel to attain the
highest allowable batch average discharge exposure consistent with the thermal-mechanical
limits imposed on the GE9 fuel product line which is approximately 38,000 MWd/metric tons.
The batch size utilized in this design is 232 bundles in equilibrium (a batch fraction of 26
percent) which results in a batch average discharge exposure of 37.0 GWd/metric tons. The
resulting maximum residence time for these bundles is, assuming the reference capacity factor of
75 percent, six years. This discharge exposure is consistent with the upper range normally
associated with urania spent BWR fuel.
The reference plutonium spent fuel designpresented here can disposition the required plutonium
inventory in just over nineteen years operation of six ABWR power plants. A summary of the
important fuel cycle parameters and plutonium consumption rates is shown in Table 1.1-1.
Table 1.1.1. Key Parameter Summary, Pu Spent Fuel Case
Numberof reactors 6Cycle Length,EFPD 392.2DischargeExposure,MWd/metric tons 37081ReloadBatch Size, bundles 232Plantcapacityfactor 75%Pu Loadingkg/bundle 5.33Pu Consumption rate metric tons/yr 5.19MOX fuel rod usage, rods/year 58000
1.1-1
1.1.1 NORMAL OPERATION
The important features of the core design for the reference Plutonium Spent Fuel Option are
summarized in the following sections. The reference fuel bundle and core design are described
in the following sections. A description of operating limits follows, showing thermal margins,
reactivity margins, reactivity coefficients and the operation of the core through an equilibrium
cycle.
1.1.1.1 Reference Bundle Design
The bundle design for the spent fuel option contains plutonium in all sixty power producing
rods. This design also has a relatively large number of gadolinia bearing burnable poison rods.
This bundle design closely resembles the reactivity characteristics normally associated withenriched uranium fuel.
The bundle axial and radial enrichment distribution is given in Figure 1.1.1-1 along with the
gadolinia and plutonia concentration distribution. Values of enrichment for uranium are read in
hundreds of a percent. For instance, the "071" refers to 0.71 w/o U:35. Only natural uranium is
used. Two reasons for this are that a slight contribution from the U235aids the reactivity
coefficients and natural uranium is thought to be easier to use in the fuel fabrication facility,
since quality control on the material is slightly higher for natural than for tails. The plutonia
concentrations (of isotopic composition given previously) are similarly given in hundreds of
weight percent. The minimum plutonia concentration is 1.00 w/o and the maximum is 4.20 w/o
PuO2. Use is made of only one gadolinia concentration (1 w/o). There are a total of seven pellet
types required for this option, three of which are gadolinia rods.
The infinite lattice radial powe_ peaking is essential in determining the peak power producing
rod. The distribution of relative power peaking (normalized to unity across the lattice) is shown
at beginning of life for the 40 percent void case in Figure 1.1.1-2. The maximum infinite lattice
relative power peaking factor is shown as a function of exposure for the 0 percent, 40 percent
and 70 percent void history cases in Figure 1.1.1-3.
1.1.1-1
Figure 1.1.1-1 GE9 Bundle Design For the Reference Pu Spent Fuel Alternative
i ii i i
iiiiii iii
• - . 3 }: : :: - : Bundle Average.w v v _.. U23sEnrichment = 0.69 w/o
_ Q __ _ __ _ _ @ Total IN/Loading = 5.33 gg
Q Q Q _ _ O _ (_ Average Fissile Fraction = 3.48 w/oMass of Heavy Metal = 179.0 Kg
_______ ActiveFuel Length = 381.0cm
®®®®®O®®®®®®®@®®®®®®®®®Q
Enrichment> 071 071 071 071 071 071 071
Plutonia > 100 160 230 28{ 280 320 420
Oadolinia > 100 100 100
I.I.I-2
Figure 1.1.1-2 Beginning of Life Lattice Power Distribution For the
Reference Pu Spent Fuel Alternative
1.302 1.273 1.287 1.308 1.309 1.288 1.274 1.305
i iii iiiii
1.273 0.788 0.706 0.760 0.761 0.706 0.789 1.276ii
1.287 0.706 0.697 0.763 0.763 0.697 0.707 1.290
"i.... .'._i!::
1.308 0.760 0.763 _i_! 0.763 0.761 1.311 i!il _,
1.309 0.761 0.763 _!_i_!i_::0.764 0.761 1.312
1.288 0.706 0.697 0.763 0.764 0.697 0.707 1.291i
1.288 0.789 0.707 0.761 0.761 0.707 0.789 1.277ii
1.305 1.276 1.290 1.311 1.312 1.291 1.277 1.307
I.I.I-3
Figure 1.1.1-3 Lattice Power Peaking
11 i iiiiii i i iii i I
1.35!_ .....
_0_ Voidlao - _ 40% Void
oooo 70% Void
•_ !.20 ......
1.15 ......•
%qb%%
% qb qt, % qb
1.05 _._,,,," . - .
I'000 5 !0 15 20 25 30 35-- 'dlO 45 50 55
. LatticeAverageExposure(GWD/ST)Figure 1.1.1-4 Hot Uncontrolled K-infinity
1.20,.1 /--. '.....1.15_ // _.,__%\\ .... 0 % Void
J / / -_',_, " ' 40% Void
110_ //.." " " .__, "'" 70%Void
10.90 "-..0.85
, ...... ,, ,.,, '--_ '-_O.8j.
0 10 20 30 40 50LatticeAverageExposure(GWD/ST_
1.1.1-4
The exposure-dependent k. are given in Figure 1.1.1-4 for the uncontrolled lattice. This figure
shows the k.. for three void histories. As seen from Figure 1.1.1-4, the design of the lattice is
such as to emulate a UO, lattice. This is accomplished by means of the widely dispersed
gadolinia in the interior rods.
1.1.1.2 Equilibrium Core Design
The equilibrium core design philosophy was to simulate the reactivity distribution of an
equilibrium UO_ core in order provide simple operation. Also, the fuel fissile inventory was
matched to the residence time of each batch in order to minimize any discharged plutonium
inventory. As stated previously, the target discharge exposure is the maximum allowed by the
thermal mechanical limits on the fuel. A single fuel nuclear design was utilized in an
equilibrium batch of 232 bundles. The detailed core design layout is presented in Figure 1.1.1-5.
The numbers shown in the beginning-of-equilibrium-cycle core map represent the relative
number of the cycle since fuel loading. For instance, the number ',1" refers to fresh fuel (loaded
this cycle) and the number "4" refers to bundles which are about to start their fourth cycle.
A single nuclear design of fuel is loaded into the equilibrium cycle. The important fuel bundle
parameters were summarized previously. A control cell core loading strategy that contains 37
control cells was utilized. Due to the improved hot to cold reactivity swing characteristics of the
ABWR core, it was possible to design the fuel with a large cold shutdown margin and still
maintain sufficient hot excess reactivity. The hot excess reactivity dictated the use of 37 controlcells.
The important parameters of the equilibrium cycle design are summarized in Table 1.1.1-1.
Examination of the results reveals that all thermal, reactivity and energy requirements aresatisfied.
1.1.1-5
Figure 1.1.1-5 Equilibrium Cycle Loading Pattern
_ iii iiiiii il ii ii i i iiii iii ii
4,
1.1.1-6
Table 1.1.1.1. Equilibrium Cycle Key Parameter Summary,
Plutonium Spent Fuel Case
Cycle. Length, EFPD .... 392Cycle Energy, GWd 1,539,,,, ,,,,,
Cycle Exposure, MWd/metric tons 9861
CoreMass, metric, tons. 149.1Reload Enrichment, W/9 U-235 3.70Reloa_dBatch.Siz e, bundles .......... 232Maximum MAPRAT 0.88
-- ,,, -- ,,,,,
Maximum CPRRAT 0.77-- ,, ,=, =,,, ,, , ,
Maxi_mum LHGR, KW/ft ( LHGR limit = !.4.4 ) 13.0MCPR_( OLMCPR = 1.25 ) 1.63Minimum Cold Slautdown Margin "' 1.9,,=, - ,,
Hot Excess Reactivity at BOC ........ 0.5
1.1.1.3 Core Thermal Margins
The critical power ratio and MAPLHGR thermal margin performance are plotted as a function
of cycle exposure in Figure 1.1.1-6 and Figure 1.1.1-7. Operation within MAPLHGR limit
assures the mechanical integrity of the fuel rods is maintained by limiting their power output in
an appropriate manner throughout their lifetime. These results demonstrate ample margin to
core thermal limits.
1.1.1-7
Figure 1.1.1-6 Equilibrium Cycle Maximum MAPRAT vs Exposure
• 1.10
1.05
1.IX) .i i
0.95
0.90
o,o__" ----0.75
0.70
0.65
0.60 1 2 3 4 5 i_ 7 II . 10
C_ck_, GWD_
Figure 1.1.1-7 Equilibrium Cycle Maximum CPRRAT vs Exposure
1.10
1.05
1.00 i
0.95 ....
0.90 _
_T 0.95
0.00 ....+., ,..._
o__ _- -o.7o----______0.95 ....
Oii6 0 S ipO 240 _00 4e 0 6. 0 600 7il 0 840 .... 9dlO 10'0
1.1.1-8
1.1.1.4 Reactivity Limit Summary
The reactivity performance of the plutonium spent fuel option design is summarized in Figure
1.1.1-8 and Figure 1.1.1-9. Due to the improved hot to cold reactivity swing of the ABWR N-
lattice, there is abundant cold shutdown margin; therefore, there is little or no impact of the
mixed-oxide fuel utilization on core design from cold shutdown margin considerations.
1.1.1.5 Reactivity Coefficients
The core dynamic void coefficient and Doppler coefficients of reactivity are plotted as a
function of cycle exposure in Figure 1.1.1-10 and Figure 1.1.1-11. The dynamic void
coefficient is somewhat larger than the upper generic limit for ABWR while the Doppler
coefficient is within the generic limits. The significance of these results is discussed in detail in
Section 1.1.3.
1.1.1.6 Core Performance Description
The core performance characteristics as a function of exposure through the cycle are given in
Figure 1.1.1-12 through Figure 1.1.1-17. The core maps in these figures show the control blade
patterns in the core expressed in terms of notches (which are 3-inch sections of blade) withdrawn
from the top of the core. Those cells which have no numbers represent cells in which there are
no blades inserted. The thermal limits and reactivity margins associated with the given exposure
are noted in the summary included with each figure. As seen from these figures, all thermal and
reactivity margins are met. The resulting core average power and exposure profile are also
given. Typical of the ABWR, the power profile shifts towards the top of the core during the last
quarter of the cycle. Since the reactors core design itself provides sufficient margins, it is not
necessary to axially grade the fuel assembly to accommodate the shift in power.
1.1.1-9
Figure 1.1.1-8 Equilibrium Cycle Hot Excess Reactivity
2.00
1.75
i
1.00
% AK ,_0.75
0.50 .... _X
0.2S _,Oo 1.o _o a_ 4.0 s.o 0.o 7.0 e.o o.o lo.o
Cyck Exposure. OWD/st
Figure 1.1.1-9 Equilibrium Cycle Minimum Cold Shutdown Margin
&0
6.6
6.0 ------- _ _ _..___ _
4.0 ........
3.S_hu°Id
tdown a.0M_usm
_5i
P-0
I.S Design1.0 Limit
1.00.5
00 1.0 2.0 S.0 4.0 5.0 e.0 7.0 8.0 9.0 lU.0Cyck_Sxlmure,OWD/_
l.l.l-lO
Figure 1.1.1-10 Equilibrium Cycle Dynamic Void CoefficientI i
0
="1
='2i I Ili I I I I I
='3
-4
m 5 ii , ,
m 6 =--
Void -7
Coef -8
('I_) -o , ,,-10 _ ,,,
-11
_,, .-.-:_.,_------ -- ___-13 ___-14
-1So 1.o 2.o 3.o 4.0 s.o e.o 7.0 e.o o.o lo.oCycleExixm_,GWO/st
Figure 1.1.1-11 Equilibrium Cycle Doppler Coefficient
l.l.l-ll
Figure 1.1.1-12 Equilibrium Cycle Core Data Summary at 0 MWD/stI I I II l lI I III I
Cycle Exposure,MWD/st 0.0
Cycle Energy,MWD 0.0
Numberof Full I_wer Days 0.0I III I
AverageVoidFraction 0.4214IIIIIIII
2 Con: Row, Mlb/hr ' 1.151
MaximumChannelPeaking 1.34762 2 CoreAxial PowerPeak 1.1928
I I
I RAPLHGR 0.7830I II IIII
2 2 MaximumCPRRAT 0.7374
HotExcessReactivity,% 0.83
2 2 Cold ShuRIownMargin 5.18H IIIII I
I
2
[_N - Numberof 3 inchincrementsthatthecontrolbladeis withdrawnfromfullyinserted
26 26
20 , \l..... 20 _
I0 _ I0
| II --
4 ,, 4
2 J 2
°o 0.2 0.4 0.6 0., _.0 s_ _.4 s.6 s., 7,.0 °o 2 4 6 s _0 n 14 _6 n2022 u 26 2, 3o
Core Axial AverageRelative Power Core Axial AverageExposure,GWD/st
1.1.1-12
Figure 1.1.1-13 Equilibrium Cycle Core Data Summary at 2000 MWD/st
I t
I I
CycleExposure,MWD/st 2000.0
CycleEnergy,MWD 344082tt ii it
Number of Full POwerDays 87.6tltt ttt t
CmeAverageVoidFraction 0.4301i II i I
6 6 Core Flow, Mlbjhr 1.151" ' 1.30 5MaximumChannelPeakingt
6 4 6 Core AxialPowerPeak 1.2005
MaximumRAJPLHGR 0.8272
4 4 I MaximumC'PRRAT 0.7119
[ Hot F.x.ceu Rmctivity. % N/A
6 4[ 6 Cold ShuldownMargin N/A
6 6
[_N Numberof 3 inchincrementsthatthecontrolbladeis withdrawnfromfully insertedi
26 26,
' I24 _ _ 24 _t_
• I22 22
18 18
16 16
14 14
Z n 12
1o IO8 II
2 .A 2 "_'rl!
00 0.2 0.4 0.6 0.l 1.0 1.2 IA 1.6 1.8 2,.0 00:2 4 6 8 I0 12 14 16 II 20 22 24 26 21130
Core Axial AverageRelativePower CoreAxialAverageExposure,GWD/st
I.I.I-13
Figure 1.1.1-14 Equilibrium Cycle Core Data Summary at 4000 MWD/st
iiiii i i i ii I ii i i ii i ii i iii i illl
CycleEnergy,MWD 688164iiiii Hill I i
28 Numberof Full PowerDays , 175.3i l I ii I iiii i I
- CoreAverage VoidFraction 0.434 Ii I i ii i iii iiii
0 0 0 Core Flow, Mlb/hr 1.151..... u_Chnel ......Maxim Peaking 1.3581
28 0_ 0 CoreAxial Power Peak 1.2133"RAPLHGR....Maximum 0.8725
" ' ' ' CPRRAT ' 'o o o Maximum 0.7166, , ,,,,,|,,
Hot ExcessReactivity,q6 1.46_I , ' ,,,, , !
28 ] 0J ' 0 28 Cold ShuMownMargin 4.95
o o o
28 28I _ _ II' II I
[_N Numberof3bichincrementsthatthecontrolbladeiswithdrawnfromfullyinsertedmind
26 ........ 26%
22 , N .... 2o "_, I
16 .... -- 16 ......
1 z |,...... , ,!It
e!!
t , I]4 ' ' jpr
2 ,_,,,; _ = ........ _ _;im _ _--00 0.2 0.4 0.6 0.8 1.0 1_ 1.4 1.6 IJl 2,0 00 2 4 6 | IO 12 14 16 18 20 22 2,1 26 28 30
Core AxialAverageReladvePower Core AxialAverageExposure,GWDIst
1.1.1-14
Figure 1.1.1-15 Equilibrium Cycle Core Data Summary at 6000 MWD/st
i ii liili iL I I I i li i I I il i
II I II II I I I 6000 ICycle Exposure, MWD/st .0I
Cycle E_e:Ny,MWD I032245
....... "- FuU.....my 9Numberof Power s 262.i ill i i inn u
CoreAverageVoidFraction 0.4460IIII [ III I ii III
o ' o ConsFlow,Mib/hr 1.151r MaximumChannelPcaddng 1.3532
0 0 0 Core Axial Power Peak 1.2820.... _L'dOR .....Maximum 0.8823
i iii
-o o MaximumCPI_AT 0.6797nlll,,i ii i
Hot Exc=u R_cfivitT, % N/AI i II I I .........
0 -0 0 Cold Shuldown Margin N/Ali ii
o o'm li
_N Numberof 3 inchIncrementsthatthe controlbladeis wilhdrawnfromfully inserted
2' _ %_,[ 22_o2' __%, II
i_ ,, _
'll ' z II' " tiil _ ,o
, li , tl
' I' ' I4 _ 4 ! ,Oo 0o.4 o.6 o.n t.o 1.2 n.4 1.6 u zo o s 4 6 8 :o _2:4 _6182on24 _ n3o
Core Axial AverageRelative Power Core Axial AverageExposure,GWDlst
I.I.I-15
Figure 1.1.1-16 Falullibrium Cycle Core Data Summary at 8000 MWD/stIIII II I i ii iiiiii [ _ I I IIIII III i[ i mlllllSlll II n IIII I I JI [I I I I
.....,, ....... .....' ,, "76327
Numberof FullPowerDays ,_k_0.6I I II III I
_ Fraction 0.3819core vo.I II I _
0 Cole Row, ldlb/lu" 1.151III III EllHII If[ m|mllm|i
MaximumChmnelPeaking 13895IIIIII _ II I --
_j 12 • 112 C,me AxialPowerPeak 1.2303III I illl ii I _ ]1 I
Maximum_OR 0.8666I nil
0 0 kt_tm_ CI_,_T 0.7_3II IIiiiii i i i iiiiii iiii I
Hot ExcessReactivity,% 0.58Ill i ilUI liililII _ I J
!2 12 Cold ShutdownMargin 4.59ill i i _ I iii iii ii IlUl lull II I
__
0
l L ! I II
B N Numberof 3 inchincremenuthatthecomrolbladeiswithdrawn fullyinsertedfrom
26 ..... 26
24 _- 24
18 , - 18 -
16 16 ,
..... 14
.... n 1_no _,no8 j 8 .....
j,6 6
4 ..... 4
J : ......2 _,d_"" ' ""- "'-
°o o.2 0.4 o.6 o.s _.o _a 1.4 n.6 n.n zo °o s 4 6 o non n4s6nn2o=u 26a 3o
CoreAxialAverageRelativePower CoreAxialAverageExposure,OWD/st
1.1.1-16
Figure 1.1.1.17 Equilibrium Cycle Core Data Summary at End of Cycle
I illIII I IIIIII IIIii ..... IIIIiiiiii II I _111 _ii
I I II ii I I _ III
CycleEzposure,MWDIst- _ I I
Cy_ r_sy, _ 15y_oooI l I II ] II I] IIII I _I_ I
Number of Full Power Days S92 0J I I I . I . _-_ IIIIIIIII I I • 1
cm Aver,eVom_ 0.373sI flit It II + I
Flow,Mlb/lu' 1.151__ .... __ Piing .
I III III1[ III I -- L I IIIII iii II I I I I III I
CornAxial Powa Peak 1.2464
0.'P967iiii I IllII
Maximum_AT 0.7524[ IIII I
_ Ho!a,_ _,cUvi_.,_ 0.0Cold Shuulown Margin 3.90i Illll -. I i i iii I
I II
r I I
I I I I
B N - Number of3 InchIncrementsthatthecomml bladeiswlthdmwn fromfullyinserted
26 r ......
%,.- *_b,.. I24 ...... 24 -
__ _ __. J
,, ..... ,, I. ,, i ,,, . ki
,, !,, |
II
, , + , II, ,_ Ii4
00 ...... 000.2 _4 06 [O i.O il i.l it ii lo 2 4 + i IO nl Ii 16 II lo 22 Ii I ill 1
Core Aria] Average Relative Power Core Axial Average Exvosure, OWI)/st
1.1.1-17
1.1.2 TRANSIENT RESPONSE OF REFERENCE FUEL DESIGN
The StandardSafety Analysis Report(SSAR) for the ABWR was submittedto the NRC in 1987.
The NRC review is approachingcompletion. All majortechnicalissues areresolved (SECY-89-
153 and SECY-90-016) and approvedby the NRC on June26, 1990, Final design approvalwas
expected but was rescheduledfor 1994.
The ABWR was designed for use of generic fuel: Therefore,much of the available behavioral
analyses for the ABWR, includingthe SSAR, are applicableto the presentstudy. In that work,
GE has evaluatedthe entire spectrumof events in nuclearsafety and operationalanalysis areas to
establish the most limiting or design basis events in a meaningful manner. The scope of thesituations analyzed includes anticipated operational occurrences, off-design abnormal
(unexpected) transientsthat induce system operationaldisturbances,postulatedaccidentsof low
probability(e.g., the sudden loss of integrity of a major component), and finally, hypothetical
events of extremely low probability(e.g., an anticipatedtransientwithout the operationof the
entirecontrol rod drive system). In the event analysis, all essential protection sequences were
evaluated until all requiredsafety actions were successfully completed. The event analysis
identifiedfront line safety systems and theiressential auxiliaries,operatoractions, and limits to
satisfy the requiredsafety actions. A partiallist of the events examinedin given below.
• Events that Decrease Core Coolant Temperature
- Loss of Feedwater Heating
- Runout of One Feedwater Pump- Feedwater Controller Failure to Maximum Demand
- Opening of One Bypass Valve
- Opening of al Control and Bypass Valves
- Inadvertent Opening of One Safety-Relief Valve
- Inadvertent RHR Shutdown Cooling
• Events That Increase Reactor Pressure
- Fast Closure of One Turbine ControlValve
- Slow Closure of One Turbine Control Valve
- Pressure Regulator Downscale Failure
1.1,2-1
- GeneratorLoadRejectionwith BypassOn
- GeneratorLoadRejectionwith Failureof One Bypass Valve
- GeneratorLoadRejectionwith Failureof All Bypass Valves
- TurbineTripwith Bypass On
- TurbineTripwith Failureon One Bypass Valve
- TurbineTripwith Failureof All Bypass Valves- InadvertentMSIV Closure
- Loss of CondenserVacuum
- Loss of AC Power
- Loss of All FeedwaterFlow
- FeedwaterPiping Break- Failure of RHR Shutdown
• Events That Decrease Reactor CoolantSystem Flow Rate
- Tripof Three Reactor InternalPumps
- Tripof All Reactor InternalPumps
- Fast Runbackof One ReactorInternalPump
- Fast Runbackof All ReactorInternalPumps
- Seizure of One ReactorInternal Pump
- One Pump Shaft Break
• Reactivity and Power Distribution Anomalies
- Rod Withdrawal Error During Refueling
- Rod Withdrawal Error During Startup- Rod Withdrawal Error at Power
- Control Rod Misoperation
- Abnormal Startup of One Reactor Internal Pump
- Fast Runout of One Reactor Internal Pump
- Misplaced Bundle Accident
- Rod Ejection Accident
- Control Rod Drop Accident
1.1.2-2
• EventsthatIncreaseReactorCoolantInventory
- InadvertentHPCFStartup
The ABWR responseto transients is reportedin Chapter 15 of the SSAR. Response to design
basis accidents is reported in Chapter 6. Finally, compliance with the ASME code for
overpressureevents is reportedin Chapter5. In developing the ABWR two bases were used for
the analysis -- design basis and licensing basis. For the design basis of the NuclearBoiler GE
chose the most conservative core design that wasknown at the time, including the possibility of
the use of mixed-oxide fuel in the future. This was the so-called Core Z design used in the plant_'
design development. In the licensing basis, reportedin the SSAR for limiting transients, GE
used a reference fuel design, typical of today's UO_ fuel offerings, called Core A. Thus, as
alternativemixed-oxide fuel designs used in this presentstudywere developed, GE alreadyhad
• a guide as to which transientsand accidentswould be themost limiting, and whethertheremightbe problemsaccommodatingthe designs.
Table 1.1.2-1 compares the nucleardynamic parameters of Core A, Core Z, and those of the
high burnup mixed-oxide core at the end of cycle, which is the most limiting point for
pressurization transients.
Table 1.1.2-1. Comparison of Nuclear Dynamic Parameters
I ........ hB__Par_eterl Core _, _ i iCore Z Hid ,utnup CoreVoid Coefficient, ¢/% -8.4 -11.6 _ -11.2
....Doppler coefficient, ¢/°C -0.31 -0.43 - -0.63
1.1.2.1 Transient Response
1.1.2.1.1 Determination of the Plant Operating Limit
The purpose of the transient response analysis is to set plant operating limits to avoid the
possibility of departure from nucleate boiling for events expected to occur during the plant
lifetime. For BWRs this has been measured by a parameter called Minimum Critical Power
1.1.2-3
Ratio (MCPR), which is the ratio of the power at which a departure from nucleate boiling is
expected to occur on the hottest rod in a fuel bundle to the bundle's current power level.
Starting from the nominal situation, i.e. MCPR = 1.0, it is necessary to factor in statistical
considerations which account for manufacturing and measurement tolerances, including
uncertainties in the thermal/hydraulic correlations developed by test programs, to measure the
departure from nucleate boiling. This sets what is called the Safety Limit MCPR (SLMCPR).
For the ABWR, this has been calculated to be 1.07, meaning that if the most limiting fuel
bundle in the core is predicted to have anMCPR.of.l.07, there is a 0.1 percent probability that a
rod will have experienced a departure from nucleate boiling. That is the acceptance criterion
approved by the U.S. NRC for BWRs.
Next, all possible transients which can occur (generally due to a single equipment failure or
single operator error) are analyzed to determine the most limiting one in terms of the change ofcritical power ratio which can occur. This so-called ACPR is added to the SLMCPR to
determine the plant Operating Limit MCPR (OLMCPR). The steady-state fuel performance
analysis is then compared to the OLMCPR to determine the amount of operating margin.
In the prior studies of ABWR with Core Z and Core A, GE determined that the most limiting
transient is the Load Rejection with Bypass Failure (LRWOBP) event. Even though this event
involves multiple failures, a long-standing GE-USNRc agreement exists to consider this event
as one of those in Chapter 15 to be compared to the MCPR safety criterion. The primary nuclear
dynamic parameter driving this transient is the void coefficient, since it is the void collapse from
the primary system pressurization which results from suddenly shutting of the steam heat sink
which cases the power rise in the core. In the Chapter 15 analyses, pressurization transients
were analyzed with GE's one-dimensional kinetics code, ODYN, which is capable of following
the effects of the traveling pressure wave through the core as the event progresses. In this code,
basic cross-sections from the three-dimensional steady-state core analysis are used, so it is not
possible to apply conservative multipliers to the key parameters - Void Coefficient and Doppler
Coefficient. Other transients were analyzed using the point model kinetics code REDY. While
this does not compare all transients on an "apples-to-apples" basis, because of the background
studies showing the pressurization transients to be limiting, the OLMCPR will be correctly
computed. The net effect of the Chapter 15 analyses reported in the ABWR SSAR is that
pressurization transients use the dynamic parameters derived from Core A, while the others use
1.1.2.4
the more limiting Core Z. Table 1.1.2-2 shows the results of the three most limiting SSAR
analyses all of which become pressurization-type transients.
Table 1.1.2-2. Results of Transient Analyses
!ii _i i ! _ _ i: i , ! _: _i i i i iD¢i_CdtiealPoWcrR_ti_!i'_,!ii!/!_,,_ili!_ii!i
Generator Load Rejection withFailure of all Bypass Valves 0.10 0.19Fast Closure of One TurbineControl Valve 0.10 0.10Fcedwater Controller Failure -Maximum Demand 0.10 0.16
Comparing the parameters of Table 1.1.2-1 confirms that the LRWOBP event will be the
limiting transient for the limiting mixed-oxide core, based on the background studies.
Therefore, it was the key one analyzed to determine the OLMCPR and available operating
margin for this study. Since two other events also showed comparable Acritical power ratios for
the UO2 core, they were also repeated, using ODYN, for this study. Table 1.1.2-3 lists the input
conditions used for the transient analyses. These are identical to those used in the SSAR.
1.1.2-5
Table 1.1.2.3 Input Parameters and Initial Conditions for S/'stem Response Transients1. ThermalPowerLevel (MWt)
Warranted Value 3926Analysis Value 4005
2. Steam Flow (kg/hr)Warranted Value 7.64x 106
Analysis Value 7.84x 1063. Core Flow (kg/hr)
Rated 52.2x106Maximum 59.0x 106
4. FeedwaterFlow Rate (kg/sec)WarrantedValue 2122
....... Analysis Value 21795. Feedwater Temperature (Celsius) 2176. Vessel Dome Pressure (kg/cm2g) 73.17. Vessel Core Pressure (kg/cm2g) 73.78. Turbine Bypass Capacity (%NBR) 339. Core Coolant Inlet Enthalpy (kcal/kg) 294.110. Turbine Inlet Pressure(kg/cm2a) 69.911. Fuel Lattice N
12. Core Leakage Flow (%) 11.6713. MCPR Safety Limit " 1.0714. Nuclear Characteristics Used in
ODYN Simulations EOEC*
15. Number of Reactor Internal Pumps 1016. Safety/Relief Valve Capacity (%NBR) at
80.5 kg/cm2g 91.3
17. Relief Function Delay (sec) 0.418. Relief Function Opening Time (sec) 0.15
19. Safety Function Delay (sec) 020. Safety FunctionOpeningTime (sec) ....... _. ....._- .....0.3 ...........
1.1.2-6
Table 1.1.2-3 (continued)
21. Set Points for Safety/Relief ValvesSafety Function 82.8, 83.5, 84.2,
84.9, 85.6Relief Function 80.5, 81.2, 81.9,
82.6, 83.3, 88.422. Safety/Relief Valve Reclosure Setpoint-
Both Modes (% of Setpoint)Maximum Safety Limit (used in analysis) 98Minimum Operational Limit 93
23. High Flux Trip (%NBR) 127.524. High Pressure Scram Setpoint (kg/cm:g) 77.725. Vessel Level Trips (m above bottom of
separator skirt)Level 8 (m) 1.73Level 4 (m) 1.08Level 3 (m) 0.57Level 2 (m) -0.75
26. APRM Simulated Thermal Power Trip (%NBR)Analysis Setpoint 117.3Time Constant (sec) 7
27. Reactor Internal Pump Trip Delay (sec) 0.1628. RIP Trip Inertia Time constant for Analysis
(see) 0.6229. Total Steamline volume (m3) 113.2
30. Set Pressure of RIP Trip (kg/cm_g) 79.1
1.1.2-7
Table 1.1.2-4. Sequence of Events for the LRWOBP
_i_i!i_ii_iiii!ii!!i!!iii_i_!i!:ii!!ji!_ii_i_i_iiiiiii_ii_!ii_i_i_i_!iiii!!iEventi,iii_,ii_i_iil!iiii_iii_ii_iii_,_:_: ili,;i:_ii:!ii,__i, :ii,Time (seconds)Turbine-GeneratorDetection of Loss of ElectricalLoad -0.015Turbine-GeneratorLoad Rejection Sensing Devices Trip toInitiate Turbinecontrol Valves Fast closure 0.0
Turbine Bypass Valves Fail to Operate 0.0Fast Control Valve Closure Initiates Reactor Scram and
Tripof Four RIPs 0.0Turbine Control Valves Closed 0.07
Safety/Relief Valves Open Due to High Pressure 1.3Safety/Relief Valves close 7.1Safety/Relief Valves Open Again to Relieve Decay Heat 8.9Safety/Relief Valves Close Again >15.0 (est.)
Table 1.1.2-5. Sequence of Events for the Fast Closure ofOne Turbine Control Valve
Time (seconds)simulate one Main Turbine Control Valve to close 0Failed Turbine Control Valve Starts to Close 0.0
Turbine Bypass Valves Start to Open 2.7Neutron Flux Reaches High Flux Scram Setpoint andInitiates a Reactor Scram 2.84
Water Level Reached Level 3 Setpoint. Four RIP'sare Tripped 7.65
1.1.2-8
Table 1.1.2.6. Sequence of Events for the Runout of all Feedwater Pumps
!ili!:_i!iiiiii'_i!iii!_:_iiiiii_:_!':}I:I,I!i{i__:_,_i__i:ii!_!Ev_n.ti!i,_ !_i:__: z:_i_i i_ ...._ _ _..... , _ Time (Seconds)
Initiate Simulated runout of all Feedwater Pumps (130% at SystemDesign Pressure of 74.9 kg/cm2g on Feedwater Flow) 0Level 8 Vessel Level Setpoint Initiates Trip of Main Turbineand Feedwater Pumps 18.45Reactor Scram and Trip of Four RIP's are Actuated by Stop ValvePosition Switches 18.46
Main Turbine Bypass Valves Opened Due to Turbine Trip ' ' 18.6 .SRV's Open Due to High Pressure . 20.2Safety/Relief Valves Close >25Water Level Dropped to Low Water Level Setpoint (Level 2) >40 (est.)RCIC Flow into Vessel (not simulated) >70 (est.)
Tables 1.1.2-4 to 1.1.2-6 show the key events occurring during the transients and Figures 1.1.2-
1A to 1.1.2-3D graphically display the time variation of key plant parameters. The ACPRs from
these events are given in Table 1.1.2-2. The most limiting event is the LRWOBP as expected;
the limiting delta-CPR is calculated to be 0.19; therefore the OLMCPR would be set at 1.26 for
this core design. Based on the steady-state results reported in Table 2.7-2, the minimum
operating margin would be 23 percent (CPRRAT=0.77 with OLMCPR=I.26 and MCPR=l.63).
GE has typically supplied fuel to currently operating BWRs with operating margins as little as
seven percent; therefore this design is acceptable.
The size of the margin (23 percent) for MCPR is still significantly larger than the 15 percent
margin requirements applied per section 1.1. This is the case even with the handicap of the
additional 0.09 delta-CPR compared to the ABWR SSAR result for the LWROBP transient
(Table 1.1.2-2). This significant margin has been maintained because this plutonium disposition
study is based upon the GE9 fuel design. This fuel design shows significant performance
advantages over the fuel design on which the ABWR SSAR is based, GE8.
1.1.2-9
Ol'+'l'l
PERCENTOF RATED
l.l.2.11 i
_I'Z:'I'I
UNITS
1.0 I _
' ,/// __ ''_'I''__
0.0 I /
1 = VOID REACTIVITY2 = _PPLER REACTIVIi_
M 3 = SC_H REACTIVITY•
M 4 = TOTALREACTIVITY
w
= \
-1.0 \
_ 1- X-2.0 I _ i i ! i I _ i0 1 2 3 4
TIRE (SEC)
Figure 1.1.2-1D Load Rejection With Bypass Failure
I = NEUTRONFLUX2 = PEAKFUELCENTERTENP3 = AVERAGESURFACEHEAT FLUX4 = FEEDWATERFLOW5 = VESSELSTEANFLOM
'150
!
// ..zoo _ •
' ,
! I ! !
0 . . . I . . . .4 8 12 16
TIHE (SEC)
Figure 1.1.2-2A Fast Cl_e of One Turbine Control Valve
1 = LEVEL (INCH REF SEP-SKIRT)Z = g R SENSEDLEVEL (INCHES)3 = N R SENSEDLEVEL (INCHES)
150 4 -- COREINLET FLOM(g)5 = PI.IHPFLOM3 (g)
100
•-. I--,'-', 4z
4
50 ! z
-- _
0 4 8 12 16
TIHE (SEC)
Figure 1.I.2-2B FastClosureofOne Turbine Control Valve
1 = VESSELPRESSURERISE (PSI)2 = STEM LINE PRESSURERISE (PSI)3 = TURBINEPRESSURERISE (PSI)
125 4 = RELIEF VALVEFLOM([)5 = BYPASSVALVE FLOM(Z)6 = TURBINESEAR FLOEI(Z)
6
75 _6
N =
°..
_ 6 • 3 t
.i i" |
-ZS _ _ I I I I I0 4 8 12 16
TIME (SEC)
Figure 1.1.2-2C Fast of One Turbine Control Valve
i
1.1.2-17
1 = NEUTRONFLUX2 ,, PEAKFUELCENTERTEHP3 = AVERAGESURFACEHE;,T FLUX4 = FEEDWATERFLOW5 = VESSELSTEAMFLOg
150
• 4 • 4 4
t .! 'i:1oo_.. i-_.-, I--
,,. b__ C_
F-
ly
!
I
0 I I I I _ I I ! I,0 5 10 15 20
TIHE (SEC!
Figure 1.1.2-3A of All Feedwater Pumps
1 = LEVEL (INCH REF SEP-SKIRT)2 = g R SENSEDLEVEL (INCHES)3 = N R SENSEDLEVEL (INCHES)4 = COREINLET FLOM(%)5 = PUHPFLOH3 (%)150
- \• i..-
# z
3 z
50 .aI
!
I
0
0 5 lO 15 20TIHE (SEC)
Figure 1.1.2-3B Run-up of All Feedwater Pumps
1.1.2-20
.//
./
•" .j"
_ i
1.0o.
-Y• /1 ' o-
, , __t/I't a,tu.u _
.. 1:,o,o,EAcT,v,T, --==_i_,v_ 2 = DOPPLERREACTIVII'Y
.- >- 3 - SCRAHREACTIVITY;., I.-.-
.,_ ,--,> 4 = TOTALREACTIVITYt_" I'"
I,,i,J
-1.0D
$i
I
m
m
I
i
-2.0 , , , , I , , , ,0 5 10 15 20
TIHE (SEC)
Figure 1.1.2-3D Run-up of All Feedwater Pumps
The large MCPR margin for the high burnup mixed-oxide fuel is the result of using the GE9
design.
1.1.2.1.2 Stability
From a safety point of view, the ABWR has provided protection from reactor instabilities which
might cause fuel damage. In BWRs density-wave oscillations have been observed under test and
operating conditions as the reactor power and flow are lowered along the rated rod line. With
this in mind, the ABWR has provided protection by excluding a zone of operation in the power-
flow map (see Figure 1.1.2-4). During power ascension by pulling control rods, if the operator
attempts to increase power above 25 percent with flow less than 40 percent (30 percent pump
speed) an automatic rod block is issued by the control system which prevents further rod
withdrawal until the minimum pump speed is achieved. During transients, if two or more RIPs
are tripped and the flow becomes less than 40 percent with power greater than 25 percent, a
Select Control Rod Run-In (SCRRI) command is issued by the control system to insert rods until
the power becomes less than 25 percent.
The basis of the above protection is to prevent the plant operators from getting the plant into an
operating state where instabilities have been known to occur. In spite of the above protection,
however, the ABWR also monitors the local power through the LPRMs and will issue a scram if
significant oscillations are detected.
The basis of the minimum pump speed limit has been determined from stability analyses
demonstrating a 95 percent confidence that oscillations will not occur at or above this point.
From many tests and analyses, GE has determined the primary drivers for instability. Once
again, the void coefficient is the most significant parameter for the purpose of the mixed-oxide
studies. Therefore, stability limits were checked to see if further restrictions in the operating
map would be necessary.
Stability is calculated by the use of frequency-domain codes which linearize the basic dynamic
equations for small perturbations and report the behavior of the reactor in terms of a second-
order feedback system. Commonly, a decay ratio is reported; this is the ratio of the second over
shoot to the first overshoot of the reactor response to a step change in input. Thus a decay ratio
1.1.2-22
<1.0 means the reactor is stable, and a decay ratio >1.0 means the reactor is unstable. In
calibration of GE's methods to test data, a 20 percent conservatism is needed to achieve a 95
percent confidence of avoiding instability; thus it is required to predict the decay ratio <0.8. The
least stable point in the operating domain of Figure 1.1.2-4 is the intersection of the rated rod
line with the minimum pump speed line. Thus, stability of the high burnup corewas calculated,
using the same procedures and methods documented in the SSAR for Core Z. The most limiting
void coefficient for the mixed-oxide core was taken from the mid-cycle point (-12.9
cents/percent, per Figure 1.1.1-11).
The results are shown in Table 1.1.2-7. It can be seen that at the current minimum speed (30
percent), i.e. core flow of 40 percent the decay ratio = 0.86. Therefore, a second case was run
with a slightly higher speed in order to determine how much further restriction would be
necessary. Referring to Table 1.1.2-7, it can be seen that a nominal increase in required
minimum pump speed will satisfy the stability criteria. This will not present any significant
operating restriction.
Table 1.1.2.7 Stability Evaluations_ii_Core Decay.Ratio l _:! i il Hot Channel,Deca_,Ratio
30% Speed/40%Flow 0.86 I 0.3935% Speed/45%Flow , 0.73 I 0.17Design Criteria <0.80 I <0.60
1.1.2.1.3 Anticipated Transients Without Scram (ATWS)
ATWS events are a special class of transients also analyzed in Chapter 15 of the SSAR. Because
of the low probability of occurrence, ATWS acceptance criteria are somewhat different thanthose for normal transients. These include:
a. Core coolability, determined by demonstratingfuel temperatures <2200Fahrenheit
b. Nuclear Boiler peak pressure not to exceed service Level C of the ASME Boiler
Code, i.e. 1500 psig.
c. Containment pressure not to exceed s_rvice Level A, i.e. 45 psig
1.1.2-23
i i i
I I I I I i I i I I i130,,-
TEN OF TEN INTERNAL PUMPS OPERATING -"
PERCENT PUMPSPEEO120 .-- ll_m3m_ ,x, .-.
0 0 NATURAL CIRCULATIONI 30
110 -- 2 40 ._3 5O4 60 7 8
100 S 70 100% POWER - 3926 MWt tt 1¢_ 6 _
6 80 ,,jr_7 90 10095FLOW - 115.1 mlb/hr
9O 8 99 4 -100% SPEED = 1500 rpm
PERCENT ROD LINE 3.lU
oo "I x ,O I_GI ON !11 2;.. a. B 100i,_ t-- FOR MOX F1EL-. z 70 C 8O B _
m D 60Ua: E 40 REGION IVm 60 F 20is. C
0 REGION III50
I
40 l J!REGION I I
3OE
#I REGION II
20 l
TYPICAL F STEAM SEPARATOR LIMIT10 STARTUP /
PATH /
0 v0 10 20 30 40 50 60 70 80 90 1O0 110 120
PERCENT CORE FLOW
Figure 1.1.2-4 Modified Power-Flow Operating Map for ABWR with MOX Fuel
In the SSAR, ATWS analyses were reported showing significant margins to the acceptance
limits. However, background studies for ABWR have shown that the peak pressure predicted
for ATWS pressurization transients is within 20 psid of that for the ASME overpressure
transient (reported below). Thus ATWS is not a limiting overpressure basis for ABWR.
1.1.2.2 Overpressure Analysis
In BWRs the number of safety/relief valves is normally set by requirements of the ASME code
for overpressure protection. This analysis is a special transient in which the MSIVs are suddenly
closed, the normal MSlV position scram is assumed to fail - the reactor scrams on high neutron
flux. In addition no credit is given for safety/relief valves opening in the power-actuated relief
mode; credit is given only for the spring mode. The acceptance criterion for this analysis is
Nuclear Boiler peak pressure less than service Level B of the ASME code, i.e. less than 1375
psig. Since this is an overpressure transient, the void coefficient will once again play a
significant role in the results. Therefore, this analysis was repeated for the limiting mixed-oxidecore.
Results show that the peak vessel pressure is <1272 psig, well within the acceptance criterion.
Even for a further delayed case using the second backup scram (high pressure scram) the result
is 1284 psig.
1.1.2.3 Accident Analyses
Response to design basis accidents is reported in Chapter 6 of the SSAR for Loss-of-Coolant
Accidents (LOCAs) and in Chapter 15 for Reactivity accidents. Reactivity accidents, such as
Rod Drop or Rod Ejection, which are dominated by the fuel Doppler Coefficient, are eliminated
in the ABWR by design features which preclude their happening. For the ABWR no analyses
were reported in the SSAR; therefore no further consideration was given to this class of events
for the present study.
1.1.2-25
1.1.3 FUEL CHARACTERISTICS AFTER IRRADIATION
The decay heat rate on a per bundle basis is shown in Figure 1.1.3-1. The average isotopic
content of the fuel bundles for the reference spiking case is shown in Table 1.1.3-1. The
isotopics are calculated five days after reactor shutdown.
Table 1.1.3.1. Average Bundle Isotopic Content (grams per bundle)Plutonium S )ent Fuel Case
Th-232 tr Kr-83 7.463 Eu- 153 24.808U-233 tr Ru-1Ol 168.020 Eu-154 9.074U-234 5.820 Cd-ll3 2.494 Eu-155 3.465U-235 382.964 Pm-148G 0.085 Gd-154 18.709U-236 144.963 Pm-151 0.002 Gd-155 0.094U-238 168000 Ru-103 6.601 Gd-156 332.846,,,,. .,,,.,,
Pu-239 1414.876 Sm-148 24.027 Gd-157 0.103Pu-240 1185.369 Xe-133 0.686 Gd-158 390.969, ,,.,
Pu-241 492.230 1-135 tr Tb-159 14.568Pu-242 223.414 Pr-143 1.936 Gd-160 210.204Pa-231 tr Rh-103 116.481 Nd-145 115.947
Np-237 27.778 Rh-105 0.023 Gd tails 3.704Pu-238 31.542 Cs-133 222.921 Fiss.Prod.
Balance 2378.898Np-239 2.321 Xe-135 trAm-241 28.525 Nd-143 133.578
,, ,,
Am-242 tr Tc-99 139.838Am-243 43.571 Xe-131 90.899,,. ,,,
Th-228 tr Pm- 147 29.849Th-230 tr Pm-148M 0.243Cm-242 10.085 Pm- 149 0.035Cm-244 13.062 Sm-147 17.438
Sm-149 0.519.,,
Sm-150 54.718, ,,,,
Sm-151 3.011Sm-152 34.573
tr - less than 0.0005 grams/bundle
1.1.3-1
Figure 1.1.3-1 Decay Heat Load/'or the Pu Spent Fuel Case
00000 ....
10000_ .... ,
iooo
wattsper __-_ ___._ __._..___bundle
00 ,, ii l
I0 500 lOGo 1500 2000 _ 3000 3500 4000 4500
• daysaftershutdown
1.1.3-2
1.2 ALTERNA_ CORE DESIGNS FOR PU DISPOSITION
The reference fuel design for discharging Pu as spent fuel will be the design reportedin the
previous section (Section 1.1) of this Phase 1C report. This design maximizes Pu utilization
with a core averageenrichmentof 3.48% and a dischargeexposure of 37 GWD/MT (typical of
the GE9 commercial assemblydesign). As reportedin Phase 1A, thisdesign requiresno change
to the reactor system. The fuel material has been well proven with Uraniafuel and requires
minimal developmenteffort. Fordispositionin 25 years (fromproject inception), 6 reactorsare
required,with the first reactorcoming on-line at the end of 7 years and each subsequent reactor
coming on-line every six months. The study assumes a 75% capacity factor. If the capacity
factoris the ABWR reference87%,only 5 reactorswill be required.
Due to the large capital cost required for the reference scenario, other options are described
which employ fewer reactors. It is important to note that the disposition time is the single most
influential variable in lowering the required number of reactors for disposition. Three cores are
described in the following sections. The first option is the reference spent fuel core, described
in Section 1.1. This option gives the most complete utilization of the excess Pu. The second
option utilizes a core design incorporating a 5% enrichment assembly and an exposure of 37
GWD/MT (with a possible redesign to accomodate 30 GWD/Mt). The third option is a core
design which employs 3.75% Pu enrichment and is capable of up to 30 GWD/MT (detailed for
Phase IA in Section 2.6).
The major parameters for the Option 2 core design which employs a 5% Pu enriched assembly
and a burnup of 37 GWD/Mt are given below in Table 1.2-1.
1.2-1
Table 1.2.1. Equilibrium Cycle Key Parameter Summary,
........... High EnrichmentCycleLength,EFPD ...........................392Cycle Energy, GWd 1,539Cycle Exposure, MWd/metrictons .' ..', 98.6.!Core Mass,metric tons 149.1i i i i ,
ReloadEnrichment,W/0U-235 5.0ReloadBatchSize,bundles..... 232,, ,,,,,,ii iIll,I I ,,, ,
Maximum MAPRAT 0.72-Maximum CPRRAT 0.75
Maximum LHGR, Kw/ft(LHGR limit= 14.4) I0.2MCPR ( OLMCPR = 1.25 ) .............. 1.67Minimum Cold ShutdownMargin 3.9HotExcess Reactivity at BOC 4.0
1.2.1 Alternatives for 100 Metric Tons in 25 Years
There are two major routesto disposing the 100MT of Pu in 25 years. One is to utilize the same
fuel design as the reference design and lower the exposure, another is to increase the enrichment.
Several different core design studies have been conducted which show both approaches are
possible. The following options are shown graphically in Figure 1.2-1.
Option 1 uses the Reference Spent Fuel core which is 3.5% enrichment and 37 GWD/MT
burnup. This design maximizes Pu utilization. Six reactors are required, with the first reactor
coming on-line at the end of 7 years and each subsequent reactor coming on-line every six
months. The study assumes a 75% capacity factor. If the capacity factor is the ABWR reference
87%, only 5 reactors will be required.
There are two possibilities for Option 2 (5% enrichment). Option 2A has 5% enrichment and a
burnup of 37 GWD/Mt. This design would require only 4 reactors for disposing 100 MT in 25
years at 75% capacity factor. Option 2B would use a redesigned 5% assembly to give a burnup
1.2-2
of 30 GWD/Mt and faster Pu disposition capability. This design would require only 3 reactors
for disposing 100 MT in 25 yeats at 75% capacity factor.
Option 3 utilizes the core design which employs 3.75% Pu enrichment and is capable of up to
30,000 MWD/MT. This design would require only 4 reactors for disposing 100 MT in 25 years
at 75% capacity factor.
1.2-3
100 Mt of Pu in 25 Years
Figure 1.2-1 Options for 100_t Pu Disposed in 25 Years
1.2.2 Alternatives for 50.100 Metric Tons in 25 Years
The system design andcore designs will be the same as the reference spent fuel and based on the
options outlined under the previous section. For disposing less than 100 MT in 25 years, fewer
reactors will be needed. These alternates are discussed below and shown in Figure 1.2-2.
Using Option 1 ( reference spent fuel) and 75% capacity factor, 50 MT of Pu could be disposed
using 3 reactors, 66 MT using 4 reactors and 83 MT using 5 reactors, in 25 years as shown in
Fig. 4. In all these cases Pu is utilized to the fullest.
The Option 2A design (5% enrichment and 37 GWD/Mt burnup) could dispose 50 Mt with 2
reactors or 75 Mt with 3 reactors. With the Option 2B fuel design (5% enrichment and 30
GWD/Mt burnup), 2 reactors could be used to dispose 50 MT of Pu in 21 years or 66 MT could
be disposed in 25 years.
Option 3 (3.75% Pu enrichment) could dispose 50 MT with 2 reactors and 75 MT could be
disposed using 3 reactors. Pu is not utilized as fully as with the Reference Fuel Design. Note
that Option 3 gives the same results as Option 2A. This will be discussed more fully in Section1.3.
1.2-5
50-100 Mt of Pu in 25 Years
mjl
I o " i
. Dwi i
I " |$ ..............................._................................................................................•................ _..............r'""_4" ,..................
i j, " : i
i , ,,_" •..io" Ii• _ _ !
0 i m "a"
w
0 ! ! [ _ _" i i [ i .
=1 [_... _ i ;__-,-'-- _ i i i [ i | i
_1_-_-----_--_ _................ L.............................. ' ......... '.---................ -- ................... '--..........................
" _ _ _ .... Option 1
i _ "" Options 2A & 3....................................... _ ............. _............................. .._ . = ..................
1 i - i ..... [.............. i .............................- ................... _ ............................. Option 2Bi : i i i i
: i
O
50 55 60 65 70 75 80 85 90 95 100
Mass of Pu Disposed (Mt)
Figure 1.2-2 Options for 50-100 Mt Pu Disposed in 25 Years
1.2.3 Alternatives for 100 Metric Tons in More Than 25 Years
The system designand core designs will be the same as the reference spent fuel and based on the
options already discussed for other alternatives. For disposal times of greater than 25 years,
fewer reactors will be needed. These alternates are discussed below and shown in Figure 1.2-3and 1.2-4.
Using the Reference Spent Fuel assembly with 3.48% Pu enrichment (Option 1), 2 reactors will
dispose 100 MT in 48 years of operation or 1 reactor could be used for disposing 50 MT during
this sa_aeperiod.
Option 2A (5% assembly enrichment, 37 GWD/Mt) will dispose 100 Mt in 36 years with 2
reactors or 50 Mt during the same 36 years with 1 reactor. Option 2B (5% assembly enrichment,
30 GWD/Mt) will dispose 100 MT of Pu in 54 years with one reactor. The design lifetime of
the ABWR is considered to be 60 years. Alternatively, 50 MT could be disposed of in 27 yearswith one reactor.
Using Option 3 fuel design (3.75% assembly enrichment), 100 MT could be disposed of in 36
years with 2 reactors or 50 MT during the same period with 1 reactor.
1.2-7
Figure 1.2-3 Options for 100 Mt Pu Disposed in Greater Than 25 Years
100 Mt of Pu, Variable Years
1. , OpUonI (
z _--[ _-- °PU°"2Bl/_!/ / /O ,f/ , V /
36 48 54
Number of Years
Figure 1.2-4 Options for 50 Mt Pu Disposed in Greater Than 25 Years
50 Mt of Pu, Variable Years
,./k/. / / / /./
Option 2A &'$ :iOption 2B . Optionl
I Option 3• r
i
,.... / / .//0-27 36 48
Number of Years
1.2-8
1.3 RELATIONSHIP BETWEEN Pu ENRICHMENT, DISCHARGE EXPOSURE,
DISPOSITION TIME, ISOTOPICS AND NUMBER OF REACTORS
This section is a generic discussion of the relationships that govern the meaningful factors in Pu
disposition. These relationships represent all reactors in the fission option.
1.3.1 RELATIONSHIP BETWEEN QUANTITY TO BE DISPOSED, TIME,
ENRICHMENT AND EXPOSURE:
There exists a direct relationship between the quantity of Pu that needs to be disposed,
disposition time, Pu enrichment, exposure and the number of reactors necessary for the
disposition process. By defining any three of these variables, the fourth will be predefined.
Since the reactor system represents a major initial (investment) cost, it is useful to explicitly state
this relationship although it is based on rather elementary principles.
Variables:
N: Number of reactor years available for disposition (does not include
time necessary for building and testing reactors)
Q: Quantity of Pu to be disposed (Mt)
X: Average assembly Pu enrichment (%)
E: Average discharge exposure (MWd/Mt)
C: Capacity factor
P: Reactor power level (MWt)
# of Reactors = Q*E ,
(0.01)*X% * N years *365 days/year * C * P
For the ABWR and the current Pu disposition guidelines:
C = 0.75 P = 3926
1.3-1
# of Reactors = O Mt * E Mwd/Mt ,
10747*X%* N years
The above equationis shown in the following:
MWd producedby reactor in given time = N years * 365 days/year* C * P MWt
Total heavy metal mass (Mt) = O Mt .0.01 *X
Total MWd from disposing given mass = 0 Mt * E MWd/Mt.0.01 *X
- The relationship between the various parameters is graphically shown in Figures 1.3-1 through
3. The number of reactors required is lowered with lower discharge exposure or with higher
enrichment or longer disposition times. It should also be noted that with lower number of
reactors, the amount of electricity produced (and therefore the revenue stream) is also lower but
the capital cost is also diminished..
The curves show hypothetical cases. The feasible design space for each variable is a subset of
the total system. As an example, a core design with 3.48% Pu enrichment and 37,000
MWD/MT exposure which satisfied all the requirements was reported in Phase 1A. Such a fuel
design would require 6 reactors for disposing 100 MT of Pu within 25 years from project
inception (or 18 years of operation). If the value of Pu enrichment were increased to 20% and if
the discharge exposure could be kept at the same value, the above relationship would indicate
that only one reactor is required. However, such a design with very high Pu enrichments
appears unfeasible in an LWR without extensive system level changes and an extensive re-
examination of the licensing basis. Even if such high Pu enrichment designs were feasible, they
may require extensive and unpredictable development costs but more importantly, as discussed
below, the spent fuel isotopics are none the better for their "high exposure" compared to a lower
Pu enrichment design where the fuel is discharged at a lower exposure.
1.3-2
100 Mt of Pu at 40 GWD/Mt
• J _! 1 J
6 , ; ...................;..... i ,i J
"J i i i iS "-....................t..... .... i I I _
% _ . . !
¢_ No l_,dditional _" ...._ ..........: ............-_...............................]-.........................i............................_..i................. _ ,.-- erable_ s . .,.! Kequires -'11<. _ Requires Consa ' : I
'_ _ Develo )ment Requ:red[ ' t • ,'% , Moderate i Developmen_: : i !_.............. ......._ .......................... Development i & Licensing Re, ,iew3 N ..............*- _,-....-i ........................T.........................I..............................................
E= , \! . ,
2 '_ i' _ ..................4J.......................--.'.-i-'..-,,.-,..-..:..._-.-,i.........................I _ ! "-.. !
I ! _ ! _ " " ,, ,, ..._,,,, ,...................................
i : i i _-
I i I _ i0
0 2 4 6 8 l0 12 14 16 18 20
Pu Enrichment (%)
Figure 1.3-1 100 Mt of Pu at 40 GWD/Mt
100 Mt of Pu at 30 GWD/Mt
Figure 1.3-2 100 Mt of Pu at 30 GWD/Mt
100 Mt of gu at 20 GWD/Mt
Figure 1.3-3 100 Mt of Pu at 20 GWD/Mt
1.3.2 RELATIONSHIP BETWEEN SPENT FUEL ISOTOPICS, Pu ENRICHMENT,AND FUEL EXPOSURE
It is generallyrecognized thatPu separatedfrom any LWRspent fuel could be ,_sedfor making
weapons, even though it may differ greatly from so-called "bomb-grade"material. A clearer
definition of the resistanceof the spent fuel to subsequentuse for weapons is needed and thiswould include the degree of difficulty in working with irradiatedfuel, the extent to which
diversion is possible without detection, the minimum number of assemblies that need to be
divertedfor a critical mass of plutonium, and the deterioration/shelf-lifeconsiderationsfor Pu
derived from spent fuel. The following discusses the isotopics of the spent fuel undervarious
core design options.
Pu is destroyed according to energy produced. For any one reactor design, a given number of
reactors will destroy a given amount of Pu over a given time period. Pu disposed out of
stockpile can change with different core enrichments and different burnups. Consider two
different core design options, one that uses a high Pu enrichment of 7% and high exposure of
40,000 MWD/MT and another with a lower enrichment of 3.5% and 20,000 MWD/MT. These
two core designs, based on the discussions of the earlier section, would require the exact same
number of reactors to dispose of a given quantity of Pu in a given disposition time. Although
the first design, in view of its "higher" exposure appears to better "utilize" the fuel and hence
lead to better denaturing, in fact the fuel isotopics at discharge is just about the same. This is
shown in Fig. 1.3-4. As a first approximation, assume the number of reactors and disposition
time are fixed and assume the range of fuel enrichments that could be used in an LWR without
considerably affecting the licensing basis (1 to 10%). For this case, the extent of Pu denaturing,
that is (Pu239 + Pu241)/Total Pu, could be expected to be very nearly the same for various core
designs. Isotopic compositions of the spent fuel do vary, but the extent of "denaturing" is aboutthe same.
If the number of Pu disposition reactors are restricted, using a higher.enrichment and higher
discharge exposure has a secondary advantageous effect. The MOX throughput needed for
higher exposure fuel (expressed in MT of MOX fuel) is lower compared to the needs of a lower
enrichment fuel which is discharged at a lower exposure. One item to remember is that beyond
1.3-6
about 5% Pu enrichment levels, additionaldevelopments/confirmatory tests may be required asthe fuel material deviates more and more from the conventional urania fuel.
1.3-7
Figure 1.3-4 Plutonium Isotopic Comparison
Pu FissileFractionI._
Oe_
Pu
0.4IIIIII i
0.3 [-------- MOx(3.5_)0.2 i-'- "" Mo,_.o_jI
O.1 BWR SpentFuel ExposureRange
00 5 10 15 20 25 80 S5 40 45 50
Exposure (MWD/ST_
1.3-8
2.0 FUEL CYCLE
2.1 MIXED OXIDE FUEL FABRICATION REQUIREMENTS FOR SPENT FUELSCENARIOS
The fabrication requirements for the fuel pins and bundles for each of the ABWR spentfuel scenarios being investigated are summarized in Table 2.1-I. (Note: detailed fuel pindesigns are given in Section l.O on Core and System Performance). Table 2.1-I identifiesand summarizes the amounts of the key strategic materials needed in each type of fuelbundle. The strategic materials are uranium, plutonium and gadolinium. The total amountsof each strategic material in each bundle is given. These materials are distributed amongthe 60 pins in each bundle.
Table 2.i-1 also summarizes the nominal heavy metal throughput requirements for the fuelfabrication plant for the ABWR spent fuel scenarios. In each scenario the fuel fabricationcampaign is planned to begin three years before the first reactor begins operation. Thedesign operating cycle length is 523 days (392 days at full power and 131 days formaintenance and refueling outage). For these multi-reactor scenarios, after the first reactor,another reactor is planned to begin initial operation every half cycle (262 days) until allreactors have started up. The duration of the campaign needed to completely transformeither 100 MT or 50 MT of weapons-grade plutonium into ABWR fuel bundles is alsoindicated in Table 2.1-1. The mixed oxide fuel throughput for the fabrication plant design,as indicated in Table 2.1-1, is specified to provide approximately 15 percent excessproduction capability.
2.1-1
,,, ,,,
Table 2.1-1
Characteristics of Mixed Oxide Requirements for the Various ABWR Spent FuelScenarios
, " ,,1, fm "" ' 'n,T i' ,,, ! i "' f ii ,, fl,'lJ , iI , ,i ,,,,, , , ,,,,, L ', , , ," ,' ,',", i
SPENT FUEL SCENARIOS
Characteristics .... 1 I 2 3 4
Plutonium, MT 100 100 50 50
Reactors, # 6 2 3 1
Time to Dispose of Pu, years 25 60* 25 60*
ABWR Fuel Bundle Design:Fuel Pins, # 60 60 60 60
Uranium, kg 172.7 172.7 172.7 172.7Plutonium, kg 5.3 5.3 5.3 5.3Gadolinium, kg 1.0 1.0 1.0 1.0
Fabrication Requirements (Material Usage):Uranium, MT/yr 168 56 84 28Plutonium, MT/yr 5.2 1.7 2.6 0.8Gadolinium, MT/hr 1 0.3 0.5 0.2
AsssemblyRequirements:MOX Fuel Pins, #/yr 58,320 19,440 29,160 9,720Bundles, #/year 972 324 486 162Campaign Duration, yrs 19.2 58.8 19.2 58.8
Fabrication Plant Design'MOX Fabrication, MT/yr 220 80 110 40
*ABWR design lifetime
2.2 MOX FUEL HANDLING AND DISPOSAL
2.2.1 Criticality Analyses for Storage, Handling & Repository
This section summarizes the criticality safety analyses conducted for theplutonium-basedfuel bundles. Both the freshfuel and spent fuel bundleswere includedinthe criticalityanalysisof shippingcontainer. In addition,preliminaryinvestigationof spentfuel storagein the repositorywas conducted.
2.2.1.1 Criticality Analyses for Fuel Shipping
For more than 20 years, the RA series shipping containershave been used byGeneralElectric Nuclear Energy (GENE) to ship BWR fuel elements to domestic andinternational customers. Extensive analyses have been performed to demonstratecriticalitysafetyof these containersfor the transportof a wide rangeof 7 x 7, 8 x 8, 9 x 9,and 10 x 10 BWR fuel assemblies. The analysis performed in this report focusesspecificallyon the GE9 bundledesignwith MOX fuel, as used in the referencespent fuelalternativefor Pu disposition.
2.2.1.1.1 Criticality Safety Requirements
There are two principal criticality safety requirementsfor shipping container: (1)classification as a Fissile Class I shipping container, as documented in 10CFR71 -Packaging and Transportation of Radioactive Material (Reference 2.2.1-1), and (2)qualification under the 1985 IAEA Regulations for the Safe Transport of RadioactiveMaterials (Reference2.2.1-2). These requirementsare brieflysummarizedin the materialwhich follows.
10CFR71 Requirements for Fissile Class I _;hinnint Container
The criticality safety requirements for a Fissile Class I shipping container, asdocumented in 10CFR71,are as follows:
"A Fissile Class I package mustbe so designed and constructed and its contents solimitedthat:
(a) Any number of undamaged packages would be subcritical in anyarrangement and with optimum interspersed hydrogenous moderationunless there is a greater amount of interspersed moderation in thepackaging, in which case the greater amount may be assumed for thisdetermination,and
(b) Two hundred fiRy (250) packages, if each package were subjected to...Hypothetical Accident Conditions ... would be subcritical if stackedtogether in any arrangement,closely reflected on all sides of the stack bywater, and with optimuminterspersedhydrogenousmoderation."
In addition, it is required that a single Fissile Class I container be subcritical withoptimumhydrogenousmoderationand when closely reflectedon all sidesbywater.
The HypotheticalAccidentConditionsreferredto hereare:
2.2.1-1
1. a 30 foot (9.15 m) freedroptest
2. a I meter free dropand puncturetest
3. a thermal exposure or fire test in which the container is exposed to800 °C for at least 30 minutes,and
4. an immersion test equivalent to at ! "_ cC _,-_ (15.25 m)of water for at least 8 hours.
IAEA Requirements for ArraYsof Shipping PackaEes
In 10CFR71, it is noted that unlimitedarraysof undamagedcontainers and 250unit water reflected arrays of damagedFissile Class I containers were requiredco bedemonstratedtc be subcritical ("Damaged" refersto the worst case condition resultingfrom the "Hypothetical Accident Conditions".) The designationof shipping containersinto "Fissile Classes" is differentfrom the current IAEA regulations which specify genericrequirementsfor arraysof containersas follows:
"An array of packages shall be subcritical. A number"N" shall be derived: =....assumingthat if packageswere stackedtogether in any arrangementwith the stack :....
closely reflectedon all sides by water 20 cm thick (or its equivalent)both of thefollowing conditionswould be satisfied:
a) Five times "N" undamaged packages without an_ing between thepackages would be subcritical;and
b) Two times "" _qagedpackages with hydrogenous moderationbetweenpackages tc nt which results in the greatest neutron multiplicationwould be suL _."
If unlimitedarraysof both undamagedand damagedpackages are demonstratedto............... --be subcritical, individual shipments of such packages are not required to be limited in
numberdue to criticality safety, but will be assigned a TransportIndex due to exposureratesmeasuredat the container.
2.2.1.1.2 Description of RA Series Shipping Containers
For more than 20 years, the RA series shipping containers have been used byGeneral Electric Nuclear Energy (GENE) to ship BWR fuel elementsto domestic andinternational customers. The RA series containers consist of rectangular steel innercontainers transportedin wooden outer overpacks. The wooden overpack containers aredesignedwith ethafoam and honeycombcushioningbetween the metal inner container andthe inside walls of the outer. The inner metal conta/ner has two internal ethafoam-cushioned channel sections each of which can hold a single fuel assembly.
The origin_ designed RA-1 inner container was modified in the 1970's toaccommodatea longer fuel assembly. This was accomplished by adding a largerend capto the body of the inner. The new designwas designated as the RA-2. Subsequently, andas a result of consideration for fabrication and handling, the longer bodied RA-3 (with a
2.2.1-2
shorterend cap) was introduced. Correspondingchanges in the outer wooden containerwere also madeto accommodatethe new innerdesigns.
The RA-3D inner metal container is constructed of stainless steel 321 with aminimum16-gaugeouter shell andstructuralandreinforcingcomponents. Insidethe innercontainer,•there is an innerbasket formed of two perforatedmetal channels. The inner..
basket is held inplace by the six 3 inch by.3 inch by 1/8 inch (7.62 cm by 7.62 cm by0.3175 cm) thick angled supports welded to the innerwall of the outer shell. Within theinnerbasket, fuel assembliesrest on additionalethafoamcushioning. At the upperend of
' the innercontainera removableend cap is attachedby.boltswhich screw,into threaded_.=_bolt holes welded onto the main body. The innercontaineris sealed by a lid and rubbergasket which are held in place by 14 stainless steel clamps. A pressurerelief valve isinstalled on the inner containerwhich is designed to pass up to 2 cfm (56.6 l/m) of air ifthe pressuredifferentialbetween the inside and outside of the containerexceeds 0.5 psi(3450 Pa).
The outer containeris a rectangularwooden box 33 inches high by 32 inches wideby 207 inches (83.82 cm by 81.28 cm by 525.78 cm) long. It is fabricatedof 2 inch by 4inch (5.08 cm by 10.16 cm) wooden studs,wood planks,and plywood and is lined with8.5to 9.0 inch (21.59 cm to 22.86 cm) thick phenolicresin impregnatedhoneycomband 3 ....to 4 inch (7.62 to 10.16 cm) thick ethafoampads. Cutoutsare madein the ethafoam andhoneycomb to accommodatethe handles and liftinglugson the innercontainer._SubjecttomeetinS the minimumpackagerequirements,the ethafoamand honeycombcushioningareotherwise arranged in the outer container to minimize vibrationaleffects on the fuelassembliesbeingtransportedin the innercontainer.
During the packagingand handlingof the RA-3D container,one or more BWRfuel assemblies are placed in the chambers in the inner container. (If only one fuelassembly is packed, the other chamber is usuallyfilled with a dummybundleto providebalanced loading.) Prior to being placed in the inner containerfor shipment,each fuelassembly is first preparedby installingplastic separatorsbetween rows and columnsof thefuel rods and byenclosingthe entirefuel assemblyin a thin plastic sheath.
The RA series shippingcontainersare currently licensed as Fissile Class I shippingpackages for the transport of 7 x 7, 8 x 8, 9 x 9, and specific 10 x 10 BWR fuelassemblies. The RA-3D shippingcontaineris also currently licensedas a Type-Apackagein accordancewith the Regulationsfor the Safe Transportof Radioa_ive Materials,1985edition (Supplement1990) of the InternationalAtomicEnergyAgency (IAEA) for generic9 x 9 BWR fuel assemblies.
2.2.1.1.:$ Analytical Technique
In this analysis,neutronmultiplicationfactors(k_'s or kefl_s)have been calculated
with the GEMER.4 (Reference 2.2.1-3) and the MCNP (Reference 2.2.1-4) Monte Carlocode. The following sections providebrief descriptionsof these two codes.
2.2.1-3
The GEMER.4 Code
GEMER.4 is an enhanced combinationof the geometry modeling capabilitiesofthe well knownKENOMonte Carlocode and the sophisticatedcross section handlingandneutrontrackingof GENE's IV_RITMonte Carlo code. MERIT is a derivative of theBattelle Northwest BMC code, and is characterizedby its explicit treatment of resolvedresonance in material cross section sets. The MERIT treatment uses cross sectionsprocessedfrom the ENDF/B-IV library.These cross sections areprepared in a 190 broad
.... groupformatand the groups inthe resonanceenergyrangehave the formof Breit-Wigner-resonance parameters. These parameters are used in explicit sampling to determinethe
value of the cross section at the neutron's energy. Since resonances are consideredexplicitly,fluxweighting of cross secamnsis unnecessaryand only one cross section set isrequired per isotope (and per temperature). Thermal scattering of hydrogen in water,paraffin, etc., is represented by the S(oql3)kernels in the ENDF/B library. The types ofreactionsconsidered in the Monte Carlo calculationsare fission, elastic scattering, inelasticscattering, and (n, 2n) collisions. Absorptionis implicitlytreatedby reducing the neutronweight through determiningthe non-absorption probabilityat each collision.
The geometry treatment in GEMER.4 includes the regular and generalizedgeometry options from the KENO-IV code and anenhanced complex embedded option ......_...which permitsgrouping of regular geometry regions inside of other such regions. In the.regular geometry treatment, a geometric configuration is generated by defining boxeswhich when stacked together in one, two, or three dimensional arraysmake up the totalmodel. Within each box, individualregions and their corresponding materials (limited toone material per region) are defined using nested simple geometry forms such asCYLINDERs, SPHEREs,and CUBOIDs. Within each box, each region must completelyenclose all previously defined regions except that successive regions may share commonboundaries. The GEMER.4 geometry package permits arraysof boxes to themselves beenclosed by the simple geometry forms so that modeling a close water reflectorcan beachieved byusing a simplewater filled CUBOID to surroundthe array.
In the generalized geometry treatment, geometric modeling is achieved using theequations for quadratic surfaces and by specifying the various materials that lie in theregions bounded by the surfaces. This option allows a description of very complicatedgeometry models, but it becomes cumbersomeand computationally inefficient when largenumbersof surfaces (such as would be required for a lattice of fuel rods) are necessary.Generalized geometry boxes can, however, be stacked in (three dimensional) arrayswithregular boxes.
In the complex embedded geometry treatment,arrays of regulargeometry boxesmay be placed inside of one or moreother boxes. For example, ifBox Type 1 describes afuel rod, Box Type 2 a Gad rod, Box Type 3 a water rod, and Box Type 4 the regionbounding an entire fuel assembly,then Box Types 1, 2, and 3 may be embedded in BoxType 4 to give the complete descriptionof the assembly. Box Type 4 may then itselfbeassembled into an arraywhich can then be surrounded by regions representingpackagingor water reflection.
2.2.1-4
The three geometry options describedabove may be used in any combinationtogeneratea geometrymodel. In the presentanalysis, the regulargeometry optionhas beenused to describethe outer regions of the shippingcontainer,and the complex embeddedoption has been used to describe the fuel assemblies and their immediatelyadjacentregions. A reflectiveboundaryconditionwas used to analyzeinfinitearraysof undamaged
' . and damaged,containers,while tile individual (inner container)geometry models werestackedusing the box arrayoption.
• The GEMER.4 code has beenvalidated by_comparison,against more than one._.hundredcritical. experiments. •.These critical experimentshave included a ..signiticant._..numberinvolving comparable lattices of fightwater reactortype low enrichedfuel rods.
"Fromthisvalidation,GEMER.4'sbias has been conservativelyestimated to be -0.003 forthe range of materials,water-to-fuelratios, and kefl_sand koo'sapplicableto this analysis.The minus sign in this value indicates that the neutron multiplicationfactors areunderpredicted.
This bias is a result of three primaryfactors. The first is the randomstatisticaluncertaintiesin the Monte Carlo calculationsof the benchmarks themselves. Typicalrandom errors (a) for these calculations are in the range of 0.0Ol to 0.005 andconsequentlythe averageof N such values has anuncertaintyon the order of 0.0005 to0.002 (o divided bythe squareroot of N).
The second significantcomponent of the bias is in the uncertaintyof the crosssection sets. As noted above, GEMER.4uses a single unique cross section set for eachisotope and hence the benchmarkcalculationsalso serve to benchmarkthe cross sectionsets. This uncertaintyin the cross section data is probablythe largestcontributorto thebias.
The thirdidentifiablecontributorto the bias is the cumulativeeffect of other code
and modelinglimitations.Theseincludeuncertaintiesdue to programmingapproximations,the broadenergygroup cross section structure(as opposed to the cross section data setitself), and inherentlimitationsof the Monte Carlomethoditself. The contributionto thebias of these types of errorsis the smallestof the three types, especiallysince these errorsare normallysmall and manyof them will average out as the diversity and numberofbenchmarkcritical experimentsincreases.
The MCNP Code
MCNP is a general-purpose, continuous-energy, generalized-geometry, time-...dependent,coupledneutron/photonMonte Carlotransportcode. It solves neutralparticletransport problems and may be used in any of three modes: neutrontransport only,photon transport only, or combined neutron/photontransport, where the photons areproducedby neutroninteractions. The neutronenergyregime is from 10"11MeV to 20MeV, and the photon energy regime is from 1 keV to 100 MeV. The capabilitytocalculatek-effective eigenvaluesfor fissilesystemsis also a standardfeature.
MCNP uses continuous-energynuclear data libraries. The primary sources ofnucleardata are evaluationsfrom the EvaluatedNuclear Data File (ENDF) system, the
2.2.1-5
EvaluatedNuclear Data Library(ENDL) and the ActivationLibrary(ACTL) compilationsfrom LivermoreNational Laboratory,and evaluations from the Applied Nuclear Science(T-2) Group at Los Alamos National Laboratory. Evaluated data are processed into aformat appropriate for MCNP by codes such as NJOY. The processed nuclear datalibrariesretainas much detailfromthe original evaluationsas is feasible.
.,
.... • _ Nuclear-data tables exist for neutron interaction, photon-interaction, neutrondosimetryor activation, and thermalparticlescatteringS(ot,13)kernels.._Over.500 neutron• .
_" " '.'interactiontables are available for approximately 100 --differentisotopes-ror,elements............. .-:-.Photoninteractiontables exist for all.elements fi'om.Zfl-through Z---94.•,Thedata in the ,.
-_ _ photon interactiontables allow MCNP to account for coherent and incoherent scattering,...... _:photoelectric absorptionwith the possibilityof fluorescent emission, and pair production. ,
Cross sections for nearly 2000 dosimetryor activationreactions involving over 400 targetnuclei in ground and excited states are part of the MCNP data package. These crosssections may be used as energy-dependent response functions in MCNP to determinereaction rates. Thermal data tables are appropriate for use with the S(c_,13)scatteringtreatmentin MCNP. The data include chemicalbindingand crystallineeffects that becomeimportant as the neutron'senergy becomes sufficientlylow. Data are available for lightand heavy water; berylliummetal, berylliumoxide,.benzene, graphite,,polyethylene,,and .,_zirconiumand hydrogenin zirconiumhydride.
2.2.1.1.4 Modeling
Modelint of the Inner Container
A geometrymodel of the RA-3D innercontainer is shown in Figure 2.2.1-1. Themetal in the shell, the innerbasket, and the angled basket supports are all includedand aremodeled as stainless steel. The perforatedinnerbasket is includedby modelingit as metalwith a reduced density (85% of the normal stainless steel density). The ethafoamcushioning between the fuel assembly and the innerbasket as well as the plastic sheathingaroundthe fuel assemblies are not included. Eliminatingthis internalmoderatingmaterialis conservativefor the following reasons:
1. Arrays of undamaged containers are over-moderatedby the cushioning materialand wood in the outer container. Therefore,the omission of moderatingmaterialswill result in increasingthe calculatedkoo'S.
2. For accident condition arrays, the-fire test (which is part.of the .Hypothetical_.Accident Conditions) completely burns away all internal flammable materials(except plastic separators), Even were this not the case, the accident arrays areanalyzed with interspersedmoderationwithinthe innercontainerwhich is variedtodetermine the optimum amount. The presence of additional ethafoam or theplastic sheaths around the assemblieswill therefore not result in greater koo'sbut
will only cause a slight change in the optimuminterspersedmoderatordensity.
2.2.1-6
Dimensions in inches
Fuel Rod Cladding Basket is 85% DensityZirconium Stainless Steel
Fuel Within Fuel Rod Inner Container and Supports areCladding Full Density Stainless Steel
Polyethylene Separators Other Open Areas areBetween Fuel Rods interspersed Water
Figure 2.2.1-1 RA-3D Inner Container Geometry Model
ModelinEof the Outer Container
The outer containeris modeled as shown in Figure 2.2.1-2. This model is alsoconservativesince it does not includeall of the moderatingcushioningmaterialsthat areactuallypresentin the package. Note in particularllhatportionsof the regions betweenthe innerand outer containersare empty (i.e. void) in the model. The model corresponds
.... in this regard to what is known as the "Minimum Packaging Model':.,.The Minimum.....PackagingModel also includesa50% reduced materialdensity of the ethafoamto permit.......some flexibilityin the arrangementof the cushioning.
Modelin_ of the Fuel Assembly
In the actual bundle design for the referencePu spent fuel alternative,there areseven distinctfuel rodtypes each with differentPu enrichmentandGad loading. Namely,the 28 peripheralrods,all without Gad,include4 difl%rentPu enrichments(4 with 1.0%,8with 1.6%, 8 with 2.3%, and8 with 2.8%);while the 32 interiorrods all contain 1%Gadandconsist of 4 rodswith 2.8% Pu, 8 with 3.2%, and 20 with 4.2%. Thus, the averagePu enrichmentis 2.057% and 3.775%, respectively, for the 28 peripheralrods and 32interior rods. The bundleaveragePu enrichmentis 2.973%. All rods contain 0.71% ofU-235.
A general practice in criticality safety analyses is to carry out k-effectivecalculations with a simplifiedfuel assemblymodel that contains only one or two fuelenrichments in the bundle. The purposes of this approach are (a) to simplify thecalculation and, more importantly,(b) to provide criticality safety results that would beapplicable to a spectrum of fuel bundle design that falls within the enrichmentspecifications. Otherwise, time-consumingcriticality safety analysiswould have to beperformedwhen a new fuel elementdesignis conceived.
The fuel assemblymodel used in this study is simplifiedto two enrichmentzones.The 28 peripheralrods were assumedto contain no Gad and have a Pu enrichmentof2.8% while the 32 interiorrods have a Pu enrichmentof 4.2% and a Gad enrichmentof1%. All rods contain0.71% of U-235.
Fuel assemblieshave been modeled in complex embeddedgeometry. The modelconsists of constructinga box correspondingto each type of unit in the fuel assemblyandembeddingthese units in anotherbox with the samedimensionsas the fuel assembly. Thefollowing assumptionsand parametershave beenused in this scheme:
1. The diameterof the fuel regionwithin the fuel rodshas been"assumedto equalthenominalinsidediameterof the Zirc cladding(0.419 inches). For determinationofthe fuel nuclide densities, the fuel pellets have been assumedto have a density ofthe maximum theoretical value of 10.96 g/cm3, which is greater than typicaldensityfactorsfor realUO2 pellets. The resultingdensity is then averagedover theinside of the claddingby applyinga smeardensity of 0.9622 (square of ratio ofpellet outerdiameter,0.411 inches,overcladdinginnerdiameter).
2.2.1-8
Dimensions in inches
m Ethafoam i I Void
_._ Honeycomb Wood
Figure 2.2.1-2 RA-3D Outer ContainerGeometryModel
2. Thethicknessof the clad was assumedto be the nominal(0.032 inches).
3. An active fuel length of 150 inches (381 cm) was assumedfor the MOX and Gadrodsfor all cases.
4. The plasticinsertswereconservativelymodeledascylindricalshells(0.020 inchthick) surroundingeachindividualfuel rod suchthatthe amountof plasticwasgreaterthan thatactuallypresent. The plasticshells extended over the full length.....of the fuel rods.
5. The gadoliniumdensity was calculated as the specifiedrpercentage of the MOX =_density. To add conservatism, the specified weight percent of gadolinium isfurthurreducedby 7.5% in the model. For example, for 2 weight percentgad,1.85 percentwas actuallyused. The displacementof MOX by gadolinium wasneglected. The averaged(or "smeared") fuel nuclidedensitiesused in this analysisare listedin Table2.2.1-1.
6. All structuralcomponents in the fuel assembly except for the cladding wereconservativelyignored.
Table 2.2.1-1Nuclide Densities (Atoms/Barn-Cm) For Fresh MOX Fuel Rods
Nuclide 2.8%Pu_o Gad 4.2% Pu/l% Gad
U-235 1.6572E-04 1.6572E-04U-238 2.2237E-02 2.1914E-02Pu-239 6.0386E-04 9.0580E-04Pu-240 3.6617E-05 5.4926E-05Pu-241 1.9272E-06 2.8908E-06O-16 4.6090E,02 4.6556E-02Gd - 3.1275E-04
2.2.1.1.5 Analytical Procedure
The criticality safety criteriafor shippingcontainers meeting the requirements.forFissile ClassI containers and the 1985 IAEA Regulations can be summarized as follows:
1. !ndiv.idual undamaged container: An individual undamaged container must besubcritical when optimallymoderated and fullyreflected by water.
2.2.1-10
2. Infinite array of undamaged containers: An infinite array of undamagedcontainers must be subcriticalwith optimum interspersed moderation betweencontainers.
3. Array of damage4,containers:
3a. FissileClassI: An arrayof 250 containerseach subjectto the HypotheticalAccident Conditionsmust be subcritical when.closely reflected by water ..andwhen arrangedin the most reactiveconfiguration.
3b. 1985 IAEARegulations: An array.of" 2N" containers,each subjectto theHypothetical Accident Conditions, must be subcritical,when closelyreflected by water and when arrangedin the most reactive configuration.The numberN sets the TransportIndex and is defined as the allowablenumberof containerswhichmay be transportedin any shipment.
It can be seen that, for the arrayof damagedcontainers,if the allowablenumberofcontai:_,_rsto be shipped is taken to be infinite,then it is clear that the requirementsfor Fissile Class I packages is a subset of the IAEA Regulations set ofrequirements. Therefore,requirement3b is used for the analysisof infinite arrayofdamaged containers.
Based on these requirements,compliance with the FissileClass I requirements andthe 1985 IAEA Regulationsare typicallydemonstratedin the following manner:
1. A single container is demonstratedto be subcritical by showing that two fuelassembliesare subcriticalwhen optimallymoderatedand fullyreflectedbywater.
2. An infinite arrayof accident condition containers is demonstratedto be subcriticalin the following steps:
a. The accident condition container is defined based on prior HypotheticalAccident Condition tests of real containers as consisting only of the innermetal container with all cushioning and burnable components removed.The absence of the sealing gasket between the lid and body of the innercontainermeans that water in-leakage mustbe considered.
b. To show that infinite arraysof damaged containers with various densitiesof interspersedwater insidethe container is subcritical.
3. An infinite arrayof undamaged containersis demonstratedto be Subcritical in the ..following steps:
a. To show that the undamaged container is over moderated bydemonstratingthat (1) placing water between the undamaged contair_ersdecreasesthe kooof the arraysand (2) placing water within the containers
decreasesthe kooof the arrays.
b. To show that an infiniteclose packed arrayof the undamaged containersissubcritical.
2.2.1-11
Extensive criticalitysafety analyses performedpreviouslyfor BWR fuel bundles inthe RA series containers have shown that the accident arrayshave significantly highermultiplicationfactors than any of the undamaged container arraysor the single containerand hence are the limiting cases in the criticality safety analysis. Therefore, only theaccidentarraysare investigatedin this study.
2.2.1.1.6 Results for Fresh Fuel Bundles
This section documentsthe resultsof criticalitysafety analysis.for.an infinite_array.of damaged containers. Fresh fuel bundles shown in Section 2.2.1.1.4 were used in the -calculation of neutronmultiplication.
Forthe RA series containers,the HypotheticalAccident Conditionsresulted in thewooden outer container and all of the internal ethafoam, honeycomb, rubber,and plasticbeing burnedaway. With the rubbersealing gasket in the innercontainergone, in-leakageof water during the immersiontest was a certainty. With the destruction of all of theburnable materials, arraysof damaged containers are no longer over-moderatedand theaddition of interspersed water may cause the array koo to increase. Other than thedestruction of the burnablematerials, the HypotheticalAccident Condition tests did notresult in any changes in the fuel assemblies or the inner container that could havesignificant impact to the criticalitysafety results. Minor changes in geometry due to thedrop test and the fire test actuallymade it moredifficult ratherthan easier to achieve theclose-packed accident arraysthat are assumed in criticality safety analyses. There was, ofcourse, no loss of neutron absorbingmaterials in the inner metal container or in the Gadrods. Froma criticality safety perspective, the key issues are water in-leakage (which isassumed to be optimum anyway) and damage to the fuel assemblies or inner containerleading to a morereactive configuration.
The results of the analysis are presented in Tables 2.2.1-2 and 2.2.1-3 based onMCNP calculations. These results are for the accident condition container arraysfor therod configurationwhich contains 1%Gad in the interior rods. Table 2.2.1-2 shows arrayneutron multiplicationvalues as a function of varying interspersedmoderatordensity forthe case with Pu enrichments of 2.8% and 4.2% in the peripheral and interior rods,respectively. The results indicate that the damaged arrayis significantlysubcriticalfor allmoderator densities. The maximum kQois found to be 0.7881 nominal, or 0.7903
includinga 2o uncertainty,at a moderatordensity of 0.075.
Table 2.2.1-3 shows array neutron multiplication values for the case with auniform Pu enrichment of 4.2% in all fuel rods. As expected, the increase of Puenrichmentfrom 2.8% to 4.2% for the peripheralrodsraises the neutron multiplicationbyabout0.04 - 0.06, dependingon the moderatordensity. However,the results indicatethatthe damaged array is also significantly subcritical for all moderator densities. Themaximumkoois found to be 0.8406 nominal, or 0.8430 includinga 20 uncertainty, at a
moderatordensity of 0.075.
2.2,1-12
Table 2.2.1-2Neutron Multiplication Factors from MCN-P
for Infinite Arrays of RA-$D Containers in Accident Condition ........(Fresh Bundle, 2.8/4.2% Pu Enrichment, 1% Gad)
FractionalWater Nominal 1oDensity K-infinity Uncertaintyo.000 0.6668 0.00110.050 0.7834 0.00110.075 0.7881 0.00110.100 0.7816 0.00120.125 0.7671 0.00120.150 0.7525 0.00131.000 0.5431 0.0041
Table 2.2.1-3Neutron Multiplication Factors from MCNP
for Infinite Arrays of RA-3D Containers in Accident Condition(Fresh Bundle, 4.2/4.2% Pu Enrichment, 1% Gad)
FractionalWater Nominal 1oDensity K-infinit.x Uncertainty0.000 • 0.7064 0.00110.050 0.8289 0.00110.075 0.8406 0.00120.100 0.8405 0.00110.125 0.8328 0.00120.150 0.8247 0.00121.000 0.6062 0.0023
Comparison Between MCNP and GEMER.4
Tables 2.2.1-4 and 2.2.1-5 show a comparisonof neutron multiplicationvalues asa function of varying interspersed moderator density between the MCNP results andGEMER.4 results. For the case with 2.8%/4.2% Pu enrichment,MCNP shows lowerneutron multiplication factors than GEMER.4. The differences range from 0.01 at amoderatordensity of 0.150 to 0.04 at a moderator density of 0.050. The case with auniform4.2%Pu enrichmentalso shows similarpattern. The cause for these differencesis
2.2.1-13
not clear at this time. More detailed study is needed to understandand resolve thedifferences.
Table 2.2.1-4Comparison of Nominal Neutron Multiplication Factors
Between MCNP and GEMER.4for Infinite Arrays of RA-3D Containers in Accident Condition
(Fresh Bundle, 2.8/4.2% Pu Enrichment, 1% Gad)
FractionalWaterDensity MCNP GEMER.40.050 0.7834 0.81790.075 0.7881 0.8138O.1O0 O.7816 O.79830.125 0.7671 0.78210.150 0.7525 0.7635
Table 2.2.1-SComparison of Nominal Neutron Multiplication Factors
Between MCN]Pand GEMER.4for Infinite Arrays of RA-3D Containers in Accident Condition
(Fresh Bundle, 4.2/4.2% Pu Enrichment, 1% Gad)
FractionalWaterDensiW MCNP GEMER.40.050 0.8289 0.86830.075 0.8406 0.8633O.100 0.8405 0.85400.125 0.8328 0.83750.150 0.8247 0.8208
2.2.1.1.7 Results for Spent Fuel Bundles
This section documentsthe resultsof criticalitysafety analysis for an infinite arrayof damaged containers for spent fuel bundles. These calculations were performedusing ageometrymodel which is similarto the one used for the analysisof freshfuelbundles. Theprincipal differences are (1) the removalof Gad (burnout in spent fuel), (2) change of Pucompositions that reflect the designed fuel bumup of 38,000 MWd/t, (3) inclusion ofbundle channel (0.1 in thick) and, (4) no plastic separatorbetween fuel rods.
The results of the analysis are presented in Table 2.2.1-6 based on MCNPcalculations. These results are for the accident condition container arraysand show array
2.2.1-14
neutron multiplicationvalues as a function of varying interspersedmoderatordensity forthe case with the spent fuel bundle that has as-builtPu enrichmentsof 2.8% and 4.2% inthe peripheraland interior rods, respectively. The results indicatethat the damaged arrayis significantlysubcriticalfor all moderator densities. The maximum_ is found to be
0.7652 nominal,or 0.7696 includinga 20 uncertainty, at a moderatordensityof 0.100.
Table 2.2.1-6Neutron Multiplication Factors from MCNP
for Infinite Arrays of RA-3D Containers in Accident Condition(Spent Fuel Bundle, 2.8/4.2% Pu Enrichment at Fresh, No Gad)
FractionalWater Nominal 1oDensity K-infinity Uncertainty0.000 0.5487 0.00360.050 0.7325 0.00280.075 0.7588 0.00250.100 0.7652 0.00220.125 0.7598 0.00270.150 0.7491 0.00331.000 0.6145 0.0029
2.2.1.1.8 Conclusion for Fuel Shipping
The criticalitysafety requirementsfor classification as a Fissile Class I shippingcontainer and for qualificationunderthe 1985 IAEA Regulations for the Safe Transport ofRadioactiveMaterials have been appliedto the ILA-3Dcontainerwith Pu-based GE9 8 x 8fuel assemblies. These have included requirements for the subcriticality cf singleundamaged containers, infinite arrays of undamaged containers, and infinite arrays ofdamagedcontainers.
Foran infinite arrayof damaged containers,it has been shown that when optimallymoderatedbywater, the system of such packages is subcriticalwith the specifiedPu-basedGE9 fuel assembly for Pu disposition application. The maximum koo including 2ouncertaintyis 0.79 for fresh fuel bundlesand 0.77 for spent fuel bundles.
Previous criticality safety analyses have shown that accident arrays havesignificantlyhigher multiplicationfactors than any of the undamaged container arraysorthe single containerand hence for the RA-3D are the limitingcases in the criticalitysafetyanalysis. Therefore,although analyses for a single RA-30 container and an infinite arrayof undamaged containers were not performedin this studs, meeting the criticality safetyrequirements for these conditions is not perceived to be a problem based on priorexperience.
L
2.2.1-15
2.2.1.2 Criticality Analyses for Spent Fuel Storage in Repository
Monte Carlo calculationswere performedon discretelymodeledspent fuel storagearrays to determinethe eigenvalue of the Yucca Mountain repository filled with spent
I MOX fuel containers in normalandaccident(water flooded)conditions.l
A spent fuel storage container is designedto hold 10 BWR fuel assembliesfor longterm repository storage. The container is a 15 foot steel drum with a 200 mil wallthickness and an outside diameterof 28 inches. It is buried verticallyin the ground, withadjacent containers spaced 15 feet away in a respective tunnel. Individual tunnels arespaced 126 feet apart. Figure2.2.1-3 shows a cross sectional view of the proposed single.,.containergeometrywith 10 BWR fuel assembliesthat was used in this analysis.
The spent fuel assemblies are discretely modeled for the criticality analysis,incorporatingend-of-life heavy metal isotopics. The plutonium content of the averagespent bundleis depleted 35%, relativeto the average fresh bundle,with appreciablyhigherburnout of Pu-239, and appreciablebuildupof Pu-240.
The numeric models used reflectiveboundaryconditions on the sides and bottomof the container arraysto simulateinfinitetunnels, both in length and number. The waterflooded scenario incorporates a 10 foot tall waterregion above the containers.
The 10 bundle container array generates approximately 3500 watts of heat 10years at_erfinal irradiation, when the bundles are installed in the repository for long termstorage.
The normal environment (non-flooded) Monte Carlo analysis yields a repositoryeigenvalue of 0.288 ± 0.003 (1-sigma). The accident (water flooded) environmentanalysisyields a repository eigenvalu_of 0.896 ± 0.002 (1-sigma). Replacing the water inthe central water rod position of each of the ten assemblies in each container with B4Cyields a water floodedrepository eigenvalueof 0.790 ± 0.003 (1-sigma).
The comprehensive Monte Carlo criticality analysis indicates that long termrepository storage of spent fuel MOX assemblies, utilizing the indicated repositorygeometry scheme, poses no criticalityconcerns.
2.2.1-16
References for Section 2.2.1
2.2.1-1. Title 10, Code of FederalRegulations, Part71, United States of America.
2.2.1-2. "Safety Series No. 6, Regulations for the Safe Transport of RadioactiveMaterials, 1985 Revised Edition_, published by the International AtomicEnergy Agency (IAEA) Vienna, Austria.
2.2.1-3. • "GEMER.4 User'sManual,',J. T. Taylor,GE Nuclear Energy, November .1989.
2.2.1-4. "MCNP u-A General Monte Carlo Code for Neutron and PhotonTransport," J. F. Briesmeister,Editor, LA-7396-M, Los Alamos NationalLaboratory,September 1986.
2.2.1-18
2.2.2 ACCEPTABILITY OF SPENT FUEL FROM Pu DISPOSITION INREPOSITORY
I. Introduction
The Nuclear Waste Policy Act of 1982 made DOE responsible for developing an
underground repository for the highly radioactive waste from civilian and DOE sites.
Amendments to this act in 1987 directed DOE to investigate only the Yucca Mountain site
for this repository. Currently the utilities have signed the standard contract given in 10CFR
Ch.III Part 961, Standard Contract for Disposal of Spent Nuclear Fuel (Reference 1).....
Article VI.A.1 of this contract provides the criteria for spent fuel acceptance. The
applicability of this contract as stated in Part 961.2 extends to "spent nuclear fuel or high-
level radioactive waste, of domestic origin, generated in a civilian nuclear power reactor."
It is assumed tha: the spent fuel generated by the disposition program will qualify under the
category of "spent fuel generated in a civilian nuclear power plant" and the rest of this
section briefly examines whether the general criteria for acceptance of spent fuel are
satisfied when MOX fuel is used instead of the standard urania fuel.
The general criteria for acceptance given in Article VI of 10CFR Part 961 (which in turn
refers to Appendix E of this Part), as well as the information which the purchaser of the
contract is to provide, are specified in this and succeeding Articles. These criteria have
been evaluated for potential applicability to spent fuel deploying MOX fuel. It has been
found that all the general criteria are met by the spent fuel to be discharged from the
disposition program.
II. General Criteria for Disposal:
The general criteria stated in Appendix E of 10CFR Part 961 is attached as Appendix B _f
this (Phase 1C) report. The spent-fuel from the disposition program will meet all the
requirements enunciated for "standard fuel" of this Appendix. These criteria do not place
any limit on the fuel bumup. Appendix B of this report also contains data on the discharge
exposures of spent fuel, both those already discharged as well as potential fuel discharges.
The average discharge exposure at this time is close to 30000 MWD/MT while the expected
average for future discharges in the year 2010 is 36000 MWD/MT.
All waste forms to be dibposed of must be extensively characterized so that their behavior
during disposal is understood. Currently, there is an extensive characterization program
ongoing that will continue for several years on LWR urania fuel. It will be necessary to
2.2.2-1
show that the MOX fuel behaves the same way. Ongoing studies include oxidation,
dissolution, cladding and gaseous release. A preliminary assessment for these areas is
given below.
III. Waste Acceptance System Requirements
DOE document DOE/RW-0351P (Reference 2) issued by the Office of Radioactive Waste
Management in January 1993 and Yucca Mountain Site Characterization Project Change
Directive CR No. DCP-060, dated 2/5/93 (Reference 3) describe the functions to be
performed and the technical requirements for a Waste Acceptance System for accepting
spent nuclear fuel (SNF) and high-level radioactive waste (HLW) into the Civilian
Radioactive Waste Management System (CRWMS). As a starting point, it is worthwhile
noting that the only significant difference between the standard spent fuel from urania
fueled assemblies and those from the MOX fueled disposition reactor, consists of the
higher fraction of transuranics, specifically Pu isotopes, in the discharged bundle. The
decay heat for up to 300 years is dominated by the abundance of fission products which in
turn depend upon the bundle exposure. Thus, the decay heat is essentially the same for the
urania fueled bundle as for the MOX fueled bundle taken to the same exposure.
Table F1.1.1 of Reference 3, reproduced in Appendix B, defines the Waste Acceptance
Criteria. These requirements have been reviewed in detail and only the following criteria
are considered to be affected by the presence of higher amounts of transuranics:
10 CFR 60.135 "(a) High-Level Waste Package in general.
"(1) Packages for HLW shall be designed so that the in situ chemical, physical, and
nuclear properties of the waste package and its interactions with the emplacement
environment do not compromise the function of the waste package or the performance of
the underground facility or the geologic setting.
(2) The design shall include but not be limited to considerations of the following
factors: solubility, oxidation/reduction reactions, corrosion, hydriding, gas generation,
thermal effects, mechanical strength, mechanical stress, radiolysis, radiation damage,
radionuclide retardation, leaching, fire and explosion hazards, thermal loads, and
synergistic interactions."
10 CFR 60.43 "License conditions shall include items in the following categories:
(1) Restrictions as to the physical and chemical form and radioisotopic content of the
radioactive waste"
2.2.2-2
10CFR60.131 "Criticality control. All systems for processing, transporting,
handling, storage, retrieval, emplacement and isolation of radioactive waste shall be
designed to ensure that a nuclear criticality accident is not possible unless at least two
unlikely, independent, and concurrent sequential changes have occurred in the conditions
essential to nuclear criticality safety. Each systems shall be designed for criticality safety
under normal and accident conditions. The calculated effective multiplication factor (Keff)
must be sufficiently below unity to show at least a 5% margin, after allowance for the bias
in the method of calculation and the uncertainty in the experiments used to validate themethod of calculation.
10CFR 72.124 "Criteria for nuclear criticality safety. (a) Design for criticality
safety. Spent fuel handling, packaging, transfer, and storage systems must be designed to
be maintained to be subcritical and to ensure that, before a nuclear criticality accident is
possible, at least two unlikely, independent, and concurrent or sequential changes have
occurred in the conditions essential to criticality safety. The design of handling, packaging,
transfer, and storage systems must include margins of safety for the nuclear criticality
parameters that are commensurate with the uncertainties in the data and methods used in
calculations and demonstrate safety for the handling, packaging, transfer and storage
conditions and in the nature of the immediate environment under accident conditions.
Of the requirements listed above, 10CFR 60.131 and 10CFR 72.124 have been
addressed in a previous section of this report where it is shown that subcriticality is
maintained for proposed geometries under consideration for repository disposal. 10 CFR
60.43 and 10 CFR 60.135 requirements primarily pertain to making sure that the
engineered barrier system (EBS) performs as intended. Containment of the waste within
the EBS is expected to be complete for 300 to 1000 years and limited to less than 1 part in
105 per year of the 1000 year inventory beyond this period. The nuclides of interest in this
regard are 14C, 85Kr and 3H and short lived isotopes such as 90Sr, 137Cs during the
containment period and 99Tc, 129I, and 135Cs during the post-containment period. Of
these, Sr, Cs, Tc and I are of interest because they are expected to move to the fuel grain
boundaries or the pellet-cladding gap region. The inventory of these isotopes should be
comparable for MOX fueled and conventional urania fueled discharges for the same level of
exposure although there are small differences in the fission product yields from Pu.
Typical urania fueled assemblies could contain up to 1% Pu at discharge. While the Pu in
the spent fuel from MOX fueled assemblies would be higher, there are considerable data
from MOX fuel developed for Liquid Metal Fast Breeder Reactors which indicate that the
movement of these fission products is not correlated with Pu enrichment. Therefore, the
2.2.-.-3
release rates of spent fuel from MOX fueled assemblies should be no worse than that fromthe conventional assemblies.
IV Other Disposal Criteria
Heat load from a single waste container (or canister) would appear to be a key
criterion for disposal. With regard to spent fuel, the repository is likely to be required to
package and dispose of a variety of fuel configurations. Container materials would have to
be selected based on material properties as well as its corrosion characteristics. The higher ....
Pu content of the PDR-SNF is judged unlikely to affect the corrosion characteristic of the
container material. A key parameter in this regard is the decay heat load of the containerand the manner in which the waste form will be distributed within the container. One
approach, proposed by LLNL, would seek to keep the emplacement hole walls above the
unconfined boiling point of water in the unsaturated zone (about 97C at the repository
elevation) while others have proposed keeping the package as cool as practical. In any
case, it is likely that the spent fuel will be "packaged" within the outer container to achieve
the required objectives. While "geometric tailoring" of the waste products (SNF) within
the container to achieve these objectives has been proposed, "receipt tailoring" or managing
the waste inventory to obtain the required levels of waste heat has also been suggested. In
the latter case, a single container may contain not only PDR spent fuel SNF waste but other
SNF as well. By such combination of measures, it will be possible to maintain the heat
load within adequate limits.
Studies and field work are still ongoing and the set of requirements for the
acceptability a given spent fuel form or composition might undergo changes in the future.
The design information for the repository given below was taken from DOE document
DOE/RW-0198, Nuclear Waste Policy Act, Site Characterization Plan Overview, Yucca
Mountain Site, USDOE-OCRWM, December 1988:
Repository Total Area 2100 Acres
Effective Repository Area 1400 Acres
Repository Heat Limit 57 kW/Acre
Container Separation along Tunnel 15 feet
Separation between Tunnels 126 feet
Container Area allocation 1890 sq. ft.
Number of Containers per acre 23
Heat Limit per container (average) 2.5 kW
2.2.2-4
However, a recent private communication (T. Doring, B&W, (702-794-1857) indicated
that the heat load requirement is undergoing further review and has not been fixed at thistime.
The PDR-SNF assemblies will be able meet the design constraints above equally as well as
conventional urania fueled SNF assemblies. No specific requirement on the allowable
quantities of actinides or fission products per container was found.
V. Spent Fuel Characteristics from Pu Dispositign
The isotopics of the spent fuel from spent fuel for the reference case where the fuel is taken
to an exposure level of 37,000 MWD/MT are discussed in Section 1.2.3 of this report.
After interim storage for 10 years prior to its placement in the repository, the heat load is
mainly due to long-lived fission products. The heat load per assembly as a function of time
following discharge is also discussed in Section 1.2.3. Ten years after discharge, the heat
load per assembly is less than 400 watts. Based on this figure and assuming that storage
commences immediately after 10 years of storage, a maximum of 6 assemblies could be
stored in each container. This is not different from the disposal scheme proposed for
typical commercial spent fuel as the heat load contribution is principally from the fission
products which are related to the exposure. The added transuranic actinides, principally in
the form of the additional Pu in the assembly (compared to typical commercial spent fuel)
does not add materially to the heat load.
VI. Conclusions
Based on the information available to-date, it is judged that the spent nuclear fuel
assemblies from the disposition reactor could be.stored in the permanent repository and that
all the applicable requirements will be met. The slightly higher transuranics in the spent
fuel does not preclude its disposal in the permanent repository.
References
1. 10CFR Ch.III Part 961, Standard Contract for Disposal of Spent Nuclear Fuel
2. DOE document DOE/RW-0351 P, issued by the Office of Radioactive Waste
Management, January 1993.
2.2.2-5
3. Yucca Mountain Site Characterization Project Change Directive CR No. DCP-060,
dated 2/5/93
2.2.2-6
D 2.2.3 SPENT FUEL PROLIFERATION RESISTANCE
I. Introduction
One of the work scope elements for this phase of the study called for an assessment of the
ability of the reference spent nuclear fuel (SNF) to meet proliferation resistance
requirements. A set of criteria/requirements in this regard is yet to be developed. Work
was however initiated in anticipation of a set of criteria becomingavailable and this section
summarizes the results of these evaluations.
II. Proliferation Resistance through Various Steps in the Disposition
Process
It was pointed out in Phase 1A studies that the proposed disposition option
•increases proliferation resistance incrementally during each stage of the disposition process:
a. Destruction of the pit shape - Requires reworking the materialb. Conversion to Oxide - Requires stripping the oxygenc. Downblending with Urania powder - Requires separation of the Uraniad. Sintered to pellet form - Urania separation is more difficulte. Loaded into fuel pins and assemblies - More difficult to divert in view of the sizef. Dispositioned as SNF - Requires extensive infrastructure, remote handling
Table 1, taken from DOE Order 5633.3A shows that attractiveness level of special
nuclear materials in various physical and chemical forms and it is seen that the
attractiveness level decreases progressively through the disposition process.
III. Safeguardability of SNF
We judge the safeguardability and diversion resistance of the spent fuel from the
disposition, process to be comparable to the SNF from conventional urania fuel for the
following reasons:
(a) Both conventional SNF and SNF from the PDR will require an extensiveinfrastructure, remote handling and processing equipment to separate theplutonium.
(b) Diversion of dispositioned SNF or commercial SNF are both easily detectable.
(c) As illustrated in Figure 1, for exposures higher than about 10000 MWD/MT, thePu isotopics of the SNF from disposition is virtually the same as SNF fromconventional urania fuel when the Pu enrichment of the MOX fuel is about 3.5%.Above this enrichment level, to achieve the same isotopic "degradation" wouldrequire proportionately higher levels of fuel exposure. These results from Figure 1are not expected to differ significantly between different LWRs.
2.2.3-1
Pu Fissile Fraction
239pu+241p UJL_
Pu
_iI uo_ " '!_ MOx(3.5%)......M°'_7"°'_l
O"I BWR Spent Fuel Exposure Range, i , I , I,, I ,, I I , ,,i _ I, •
0 5 10 15 20 25 30 35 40 45 50
Exposure (_D/S'O
Figure J,, Plutonium Isotopic Comparison
It should however be pointed out that no criterion governing the dependence of
proliferation resistance on Pu isotopics is available. It is also not clear that a basis for such
a criterion exists in view of the significant fission cross-sections of all Pu isotopes for high
energy neutrons.
The only aspect where the SNF from conventional urania fuel and MOX fuel SNF
differ slightly concerns the amount of Pu in a single assembly. Typically, high burnup ,.
SNF from conventional urania LWR assemblies has between 0.75 and 1% Pu as a result of
breeding from U238_.The Pu enrichment in the SNFfromdispositionedfuel is more nearly .... .
on the order of 1.8% for the reference fuel design. For example, the amount of Pu in a
single SNF assembly (for the reference design which uses 3.5% Pu enrichment), is 3.3 kg
of which 1.4 kg is Pu 239 and 1.2 kg is Pu 240. In disposition options where the initial Pu
enrichment is higher, the dispositioned SNF will contain more Pu. No criteria are available
relating the amount of Pu ir_ a single assembly to proliferation resistance, nor is it clear
whether there is a basis for such criteria. Given the small differences in the amounts of Pu
per assembly, it is judged that the proliferation resistance of SNF from the dispositioned
fuel and commercial fuel is comparable.
Conclusions:
Available (open literature) historical data base on SNF from commercial urania fuel
would appear to demonstrate that SNF possesses a very high level of proliferation
resistance with the implementation of appropriate safeguards. It is therefore concluded that
the proposed method of dispositioning Pu as SNF constitutes a verifiable and demonstrated
means of making this material proliferation resistant with existing technology.
2.2.3-2
2.3 QUALIFYING AND LICENSING MOX FUEL
2.3.1 REVIEW OF MOX FUEL LICENSABILITY
The considerations relative to lead assembly testing to qualify MOX fuel were examined in detail
during Phase 1B of this study and reported in Section 2.2 of Reference 1. During this phase of
this study, a more detailed examination was conducted with the following objectives:
(a) Complete a review of conformance to NRC requirements to license MOX fuel design inthe PDR
(b) Develop a program plan to test MOX fuel assemblies in existing research or industrial
reactors on an accelerated schedule consistent with existing NRC requirements, and
(c) Analyze U.S. Industry capability to support either BWR or PWR lead assembly testing.
Based On the evaluations givenin the Phase 1B report and further developed during Phase
1C, it is concluded that:
• A preliminary license for the use of MOX fuel could be granted based on the extensive
data already available for this fuel without any lead tests
• Limited lead tests, primarily in the form fuel of rods rather than fuel assembly
irradiations, might be required, with the specific objective of verifying the performance of MOX
fuel fabricated by the recommended process
• The lead test program should be used to validate the interface between the processes now
being developed for the destruction of the pit shape and its subsequent processing to fabricateMOX fuel
• The lead test program should be used to validate any new production techniques
including verification of procedures arising from automation/safeguards implementation
• All anticipated questions relative to final licensing could be resolved with either further
evaluations of the existing data base or from lead tests within 4 years of the program inception.
• Full scale MOX assembly tests are not needed for licensing. Limited full MOX assembly
tests might be conducted for confirmatory purposes.
PHASE 1B EVALUATIONS SUMMARY:
In Phase 1B it was concluded that although an extensive literature and data base are
2.3-1
available for Mixed (uranium-plutonium) Oxide (MOX) fuel properties and its performance in
LWRs, there are a number of reasons to undertake an early program of Lead Testing (LTA) in the
plutonium disposition project. The reasons included:
a. The need to develop and demonstrate a fuel fabrication process that complements the
proposed processes for destruction of the pit shape. A process of hydriding followed by
dehydriding has been proposed (Ref. 2) to destroy the pit shape. Currentlyexperimental
verification of this process is under way at LANL and LLNL.-Although additional work remains to
be done, it is clear that any proposed MOX fabrication process should complement the work
already being done in this area and still yield high quality MOX pellets.
b. Although urania-gadolinia pellets have been fabricated in large quantities in the past and
no difficulties are anticipated in fabricating MOX - gadolinia pellets, there is little data on
" fabrication of MOX mixed with the Gadolinia or other burnable poisons such as Erbium oxide or
Europia. There is a need to verify the fabrication parameters for this fuel.
c. There is a general need to reestablish the MOX fabrication technology in this country as
no large scale MOX fabrication has been carried out in more than 20 years. In this context, it is not
only necessary to reestablish the old technology which could probably be accomplished very
quickly but to define and incorporate safeguards and automation techniques to improve
productivity and lower worker exposure. The LTA program could serve as a vehicle to validating
any new production technique.
d. While nuclear analysis codes are readily available incorporating the full complement of
cross-section information needed for the design of a MOX core and the nuclear performance of
MOX fuel has been verified using partial core loads of this fuel in LWRs, additional benchmark
data for the nuclear performance of MOX fuel with burnable poison would help to better calibrate
the methods already in-place. Both self-shielding effects and poison burnout profile as a function
of radius within the fuel rod are considered important information that need to be benchmarked by
experiments.
e. It is desirable to obtain axial, radial power profiles and thermo-mechanical performance
data for full length rods from a MOX fuel assembly to provide confirmation of the code predictions
with experimental data.
2.3-2
A preliminary LTA (Lead Test Assembly) program was proposed with the following major
objectives: (1) to validate the fuel fabrication process, (2) to verify the thermal and mechanical
behavior of the MOX fuel thus fabricated, (3) to verify fuel rod nuclear performance and finally (4)
to verify the integral assembly performance. The plan proposed consisted of a three pronged
approach, first verification of fuel fabrication, second verification of individual fuel rod
mechanical, thermal and nuclear performance principally by conducting testsin research reactors
and finally verification of assembly performance through full scale MOX assembly testing. It was
pointed out that the last of these, full scale MOX assembly tests, was not considered essential for
validation and licensing of MOX fuel in PDR. This is because tests of isolated MOX assemblies in
existing reactors did not provide any additional data on overall system response as this - the overall
system response - will clearly be decided by the preponderance of urania assemblies in the core. In
addition, it was concluded that such full scale MOX assemblies will most likely be of a different
design so as not to perturb the already licensed envelope of the existing reactor core, in particular
the performance of adjacent urania assemblies. Therefore, tests of full scale MOX assemblies are
only of limited value.
During the present phase of this study, a more critical review of these aspects to MOX
qualification was undertaken. A step by step review of what it takes to license MOX fuel to NRC
requirements was conducted. The details of this review are presented in Appendix A while the
overall summary is included in this section. It is concluded that the most important aspect of
licensing the MOX fuel, considering that a vast MOX fuel data base already exists, is to fabricate
MOX fuel of sufficient quantity with the desired microstructure by a validated process and to show
its thermal-mechanical performance by rod tests in a research reactor. Based on this review, the
program plan presented during Phase 1B was updated. Finally, the industry capability to support
lead testing was evaluated.
The Standard Safety Analysis Report (SSAR) for the ABWR was submitted to the NRC in
1987. The NRC review is approaching completion. All major technical issues are resolved
(SECY-89-153 and SECY-90-016) and approved by the NRC on June 26, 1990. Final design
approval was expected but was rescheduled for 1994.
As presently conceived, and as evaluations to-date have shown, GE's ABWR will be used
for Pu disposition without involving any system level modifications whatever. The only change
relates to the use of MOX fuel instead of the standard urania fuel. New fuel licensing is carried out
by a process of documenting the compliance with the criteria of Amendment 22, a process
2.3-3
described below in more detail. Line-by-line compliance status of the Amendment 22 criteria is
included in Appendix A.
In developing the ABWR, two bases were used for the analysis -- design basis and
licensing basis. For the design basis of the Nuclear Boiler, GE chose the most conservative core
design that was known at the time, including the possibility of using mixed-oxide fuel in the
future. This was designated the Core Z designused for the plant design development. In the
licensing basis, reported in the SSAR for limiting transients, GE used a reference fuel design,
typical of today's UO2 fuel offerings, called, Core A. Thus, asalternative mixed-oxide fuel
designs used in this present study were developed,-GE already had a guide as to which transients
and accidents would be the most limiting, and whether there might be problems accommodating the
designs.
A report of the description of the fuel licensing acceptance criteria for the GE9 fuel design
is specified by Amendment 22 of the GESTAR II document (General Electric Standard Application
for Reactor Fuel, NEDE-24011-P-A). The amendment contains the basis for generic compliance
of the GE9 fuel design with those criteria. The reference fuel design for Pu disposition in fact uses
this standard GE9 fuel design with the exception that the fuel will be MOX fuel instead of thestandard urania fuel.
The proposed method of disposing Pu by the use of its conversion to MOX fuel and its use
in an ABWR entails no change whatever to the reactor system design except for the use MOX fuel
in the standard GE9 fuel design. Therefore, a detailed evaluation of Amendment 22 with respect to
this change should, as a start, establish the licensability of the ABWR system and the use of MOX
fuel. In this evaluation, two different approaches are possible:
a. To describe a specific MOX fuel design and establish the licensability of this particular
fuel design, or
b. To establish the licensability of MOX fuel in a generic manner, similar to the generic
licensing of Urania fuel under the GE9 fuel design, that is, to establish the licensability of MOX
fuel over an umbrella design range.
It should be noted that in either case, the satisfaction of these criteria will require cycle-
unique analyses which must be performed after the core loading for that'cycle has been specified.
For these cases, the generic information contained in this section will be supplemented by plant
cycle-unique information and analytical results. In general, this cycle-unique information will be
documented in a separate cycle-unique reload licensing report for each reload.
As reported in Phase 1B, if the intent is to obtain approval of MOX fuel for the government
owned PDR for operation on a government reservation, rather than to obtain approval for
2.3 -4
commercial operation of a BWR with MOX fuel, it is likely that the fuel safety review will be
conducted as part of the SSAR review, and the GESTAR process will not be used formally. Even
if the PDR is owned and operated by an independent power producer with the Government
providing the fuel, it appears likely that a special MOX fuel licensing review will be conducted
since present government policies do not envision private licensing of MOX fuel. However, the
technical elements are expected to be identical to those of the GESTAR Amendment 22 compliance
review process. Therefore evaluation of the GESTAR steps for licensing MOX fuel is relevant tothe PDR.
When the NRC is notified that the generic analyses for the new MOX fuel design are
completed and all criteria are satisfied, the fuel design is "licensed". No prior NRC review and
acceptance is required. If a specific criteria is not met, prior NRC review and acceptance is
required; however, the review is limited to the area of noncompliance. The reference UO2 fuel
design for the ABWR meets the acceptance criteria of Amendment 22 as documented in Appendix
4D of the ABWR SSAR. No further NRC review is required if the Combined Construction and
Operating License applicant utilizes this fuel design. It is the current NRC position that any change
to the reference fuel design for initial core application (e.g.. MOX designs) will require a change to
the SSAR and NRC review. However, if the NRC has been previously notified that the MOX
design meets the Amendment 22 criteria, this review should be only a formality.
An assessment has been completed to determine the impact on the licensing process for an
ABWR assembly with MOX substituted for the UO2 fuel. It is concluded that the MOX fuel
design will have to be treated as new fuel and compliance with Amendment 22 acceptance criteria
has to be demonstrated. There are three areas where the use of MOX fuel is expected to have an
impact on the licensing process. These are:
1. Amendment 22 requires the use of NRC approved analytical models and analytical
procedures. This will require prior NRC review of the analytical methods if the current NRC
Safety Evaluations for these methods do not address MOX applications.
This area was treated in some detail in the Phase 1B report. No changes to any of the
thermal-hydraulic or structural models are needed. While no changes to the basic methods already
in place for nuclear analysis are needed, calibration of the diffusion/transport models used in
nuclear analysis would be needed. This calibration will be performed by comparison with the
results of the _ _re exact Monte-Carlo solutions. The results of preliminary studies in this area
were reported in Phase lB. Additional calibrations require more detailed analyses which will be
2.3-5
performed in Phase 2. Once the calibrations are in place, utmost small changes might be indicatedto the Pu enrichment.
2. Amendment 22 requires that new design features be included in Lead Use Assemblies
(LUAs). If MOX is considered a new "feature" LUAs will be required before MOX is licensed.
As presently designed, MOX fuel assemblies are not considered to presentany new
"feature". The assembly thermal-hydraulic and mechanical response is expected to be identical in
every respect to the already licensed GE9 bundle design. The effect of using MOX fuel is internal
to the pin and the technology elements that need to be tested relate to the behavior of the fuel rather
than the bundle. This area is clarified in more detail in Appendix A. In summary, it is concluded
that while MOX pin tests are needed with a view to confirming the behavior of the fuel fabricated
by a representative process and which has undergone the necessary quality assurance steps, LUA
tests are not considered necessary.
3. Amendment 22 includes a general criteria that encompasses "new-fuel-related licensing
issues identified by the NRC." For MOX fuel this could result in the NRC request for new criteria
to address the concern(s).
It is this change, the potential changes to the behavior of the fuel, which might result in
new limits, that needs to be assessed in detail. This is the area that is covered in detail in this
section. It should be pointed out that typically LWR fuel at discharge contains up to 1% Pu that is
bred from the uranium during the course of irradiation. The reference fuel designs recommended
by GE for Pu disposition generally are small extrapolations in that the average Pu enrichment is
kept below 5% in all cases and for the reference case is below 3.5%. The wealth of data available
from MOX fuel irradiations fabricated from reprocessed fuel as well as the data generated in DOE
sponsored programs on MOX fuel containing up to 20% Pu for Liquid Metal Reactors in the past,
provide the necessary foundation on which we can confidently predict the behavior of MOX fuel in
the ABWR. This is supplanted with fuel pin tests to be carried out under the proposed LTA
program.
Therefore, from a licensing perspective, the two areas that need to be addressed are, first,
the calibration of nuclear methods for MOX applications and second, the effect of changes in
physio-chemical properties of the fuel in going from urania to MOX and its impact on fuel thermo-
mechanical behavior. A summary of the evaluations in these two areas is given below:
MOX FUEL ROD NUCLEAR PERFORMANCE: CODE CALIBRATION
2.3 -6
Confirmation of the core nuclear performance might require further calibration of the codes
now in use for urania fuel. The principal objective will be to conduct selected rod and assembly
tests to provide the required data for calibration of the diffusion/transport nuclear codes, in
particular for the rate of poison burnout as a function of radius in a MOX fueled rod. In addition,
it is expected that extensive cross-calibration will be done using Monte-Carlo codes.
MOX FUEL ROD THERMAL-MECHANICAL PERFORMANCE ,.
Fuel rod thermal-mechanical performance could be expected to be affected by the change
from urania to MOX fuel. A technical review of the available urania-plutonia mixed oxide fuel
properties and performance information has been conducted to qualitatively assess differences,
with respect to urania fuel, that may affect fuel rod thermal-mechanical performance relative to
design and licensing criteria. The results of this review given below indicates that for the range of
Pu enrichments considered in this study, the thermal-mechanical response of the fuel is predictable
using the existing data base and the fuel is expected to perform as well as the standard urania fuel,
The specific fuel properties and performance characteristics investigated include:
Theoretical density
Elastic modulus
Creep
Thermal expansion
Thermal conductivity
Enthalpy
Melting temperature
Radial power distribution
Fission gas release
Theoretical Density:
Plutonia lattice parameter measurements exist to enable a quantification of the urania-
plutonia theoretical density as a function of the plutonia concentration. The fuel theoretical density
(gm/cc) increases with increasing plutonia concentration; the theoretical density of 10 w/o PuO2 -
UO2 is - 0.5% greater than the theoretical density of urania. The effect of this difference on fuel
rod thermal-mechanical performance is that, for the same fuel exposure accumulation and fraction
of theoretical density, a slightly greater number of fissions would occur per unit volume of fuel
2.3-7
thereby leading to slightly greater fuel irradiation swelling and gaseous fission product inventory.
The difference, however, is relatively minor.
Elastic Modulus:
Measurements of the elastic modulus of plutonia indicate that the elastic modulus of urania-
plutonia fuel increases with increasing plutonia concentration; the elastic modulus of 10 w/0 PuO2
- UO2 is ~ 1.5% greater than the elastic modulus of urania. ,This difference.results in a slightly
stiffer pellet, and correspondingly higher stresses for a given elastic strain. This difference is
relatively minor and overshadowed by the difference in fuel material creep behavior describedbelow.
Creep:
Measurements of the temperature-dependent creep behavior of urania-plutonia
compositions indicate a significant increase in the fuel creep rate relative to UO2; the creep rate of
10 w/o PuO2 - UO2 at representative operating temperatures and stresses is - 3 times higher than
that of Urania under the same conditions. This difference indicates a more compliant fuel pellet
and correspondingly lower fuel rod cladding stresses for the same loading conditions. This
difference also indicates a lower pellet-cladding interfacial pressure and correspondingly, lower
pellet-cladding thermal conductance and higher fuel temperatures for the same loading conditions.
Thermal Expansion:
Measurements have been performed to determine the isothermal thermal expansion
coefficient of plutonia. These measurement results indicate decreased fuel thermal expansion with
increasing plutonia concentration; the thermal strain at 2500F for 10 w/o PuO2 - UO2 is -- 95% of
that for urania. The effect of this difference on fuel rod thermal-mechanical performance is a
slightly reduced imposed strain on the fuel rod cladding by the fuel pellet (for the same fuel
temperature condition) and correspondingly reduced cladding stresses under certain loading
conditions (such as a rapid power increase).
Thermal Conductivity:
Measurements of the thermal conductivity of plutonia and urania-plutonia compositions
have been performed by drop calorimetry methods. The measurement results consistently show a
slightly decreasing fuel thermal conductivity with increasing plutonia concentration, the thermal
conductivity of 10 w/o PuO2 - UO2 at typical operating temperatures is ~ 97% of that of urania.
The effect of a lower fuel thermal conductivity is to increase fuel temperatures, and
2.3-8
correspondingly fuel thermal expansion and fission gas release, for the same operating powerlevel.
Enthalpy:
Enthalpy measurements of plutonia and urania-plutonia compositions have been performed
by drop calorimetry methods. The measurement results indicate a slightly higherenthalpy than
urania at lower temperatures and a slightly lower enthalpy at higher fuel temperatures; the enthalpy
of 10 w/0 PuO2- UO2 is ~ 1% higher than that for urania at 816C (1500 F) and ~ 1% lower than
that for urania at 1371C (2500F). This difference is negligible.
Melting Temperature:
Extensive measurements using the thermal arrest technique have been performed by GE to
determine the solidus and liquidus boundaries of urania-plutonia over the entire composition range
..... (0-100% PuO2). The measurement results indicate a decreasing melting temperature (solidus) with
increasing plutonia concentration; the melting temperature of 10 w/0 PuO2 - UO2 is ~ 60C
lower than that for urania. One US licensing constraint is that fuel temperatures not exceed the
melting temperature during normal steady-state operation, including anticipated occurrences.
Therefore, this lower fuel melting temperature will reduce, to a small extent, the operational
capability of the fuel.
Radial Power Distribution:
For urania fuel, the radial power distribution across the fuel pellet is relatively fiat at the
start of irradiation. With continued irradiation, the progressive buildup of plutonium near the fuel
pellet outer surface causes a significant peaking in the radial power distribution near the fuel
surface, with a corresponding depression of the power in the pellet interior. The effect of this
time-varying change in the radial power distribution is to reduce the pellet centerline temperature
for the same power level and pellet surface temperature. For urania-plutonia fuel, the presence of
plutonia near the pellet surface causes an increased peaking (relative to urania) over a greater region
toward the pellet surface even at the start of irradiation. This increasingly surface-peaked radial
power distribution persists throughout lifetime for urania-plutonia fuel. The effect of this
difference is slightly reduced fuel central temperatures, relative to urania fuel, for the same
operating power level and fuel surface temperature.
Fiss;on Gas Release:
The release of gaseous fission products from the fuel to the fuel rod void space produces
two primary effects:
2.3-9
I
(1) The thermal conductivity of the gaseous fission products is approximately an order
of magnitude lower than the helium filler gas introduced during the fabrication of the fuel rod.
Therefore, the release of gaseous fission products from the fuel pellets to the void space reduces
the gas mixture thermal conductivity, reduces the conductance between the fuel pellet and the
cladding, and increases fuel temperatures. The fission gas release mechanism is fuel temperature
dependent, and therefore, this fueltemperature increase produces additional fission gas release,
(2) The release of gaseous fission productsto the fuel rod void space increases the fuel
rod internal pressure. The maximum fuel rod internal pressure is limited by US licensing
constraints. Therefore, excessive fission gas release and fuel rod internal pressure can limit the
operating conditions or mechanical design of the fuel.
The available information indicates that the fission gas release behavior of urania-plutonia
" " fuel is significantly affected by the as-fabricated fuel microstructure. For example, significant open
porosity (fuel pellet pores in direct contact with the pellet surface) increases the rate of fission gas
release. In addition, inhomogeneous microstructure resulting in localized islands of pure PuO2
and/or UO2 also increases the rate of fission gas release. The preferred microstructure includes
standard grain sizes (8 - 10l.tm), low open porosity (<0.5% TD), and 100% homogeneous
(U,Pu)O2 solid solution. With this preferred fuel microstructure, no difference in the fundamental
fission gas release processes and behavior are expected relative to urania fuel.
US fuel licensing criteria can be broadly categorized as either (1) thermal performance
limits (e.g., fuel melting temperature limit), or (2) mechanical performance limits (e.g., fuel rod
cladding stress and strain limits). The urania-plutonia fuel properties and performance assessment
indicates that, relative to urania fuel, fuel thermal performance is somewhat less favorable and fuel
mechanical performance is somewhat more favorable. These differences, are, manageable and can
be accommodated either by the thermal-mechanical design of the fuel rod or specification of
operating constraints. The performance area that can be influenced to minimize performance,
design and licensing differences is fuel pellet fission gas release; application of known preferred
microstructural features will eliminate the potentially significant difference.
Based on the above evaluations, it is concluded that a preliminary license could be granted
for use of MOX fuel without additional testing. Specific issues could be raised by the licensing or
reviewing agency which could be addressed in the LTA program.
2.3-10
References:
1. "Study of Pu Consumption in Advanced Light Water Reactors, Compilation of Phase 1B
Task Reports," GE Nuclear Energy, RFP DE-ACO3-93SF19681, September 15, 1993.
2. Haschke, et. al.,"A Hydrogen Recycle Process for Plutonium Recovery," LA-12086-MS,
Lawrence Livermore National Laboratory, 1991.
2.3-10A
2.3.2 PROGRAM PLAN FOR LEAD FUEL TESTING
The Lead Test Assembly program outlined below was developed by GE during Phase IB of
this study. In developing this program, input was received from all disciplines within GE as well
as from national laboratory experts at LANL, LLNL and WHC. The program envisions obtaining
all the necessary confirmatory data for fuel fabrication within 12 months of program initiation,
verification of individual rod behavior within 36 months and final licensing of MOX core through
calibration/validation of nuclear analysis within 48 months.
These projections are based on pursuing an aggressive schedule, not limited by funding
availability.
A. Objectives"
The objectives ol. is program are:
Fuel Fabrication ValidationFuel Mechanical/Chemical Performance Validation including Fuel Properties Data-Base
Generation if RequiredFuel (Rod) Nuclear Performance Benchmark Data Generation
The objectives under each of these categories are described in more detail below:
Fuel Fabrication Valid_tiQn
• Develop MOX Processes to Complement Pit Processing
As a first step in safeguarding the weapons plutonium, the pit which contains the plutonium
in a weapons has to be removed and its shape destroyed. Although a number of techniques are
available, research in this area has focused on a "chemical processing,' means as a clean, safe and
economical means of pit shape destruction. This process consists of hydriding and dehydriding the
plutonium. In such a process, the pit shape falls apart leaving a powdery plutonium as the final
state. It is therefore useful to consider MOX processing steps which complement this process. If
the fission process is not chosen as the route to plutonium disposition, the Pu powder from this
process would presumably be melted or converted to an oxide and stored in a desired unclassified
shape.
2.3-11
The plutonium obtained from the hydride-dehydride process could be directly converted,
under a controlled oxygen partial pressure, to produce plutonium oxide. This could form the
starting stock for the MOX process with urania and gadolinia powders to be supplied by a
commercial vendor such as _3E. However, the activity of the plutonia powder and the particle size
distribution thereof, may not be compatible with what is required for MOX fuel fabrication.
Therefore, further studies on the milling of the plutonia powder and blending.with the urania have
to be conducted under controlled conditions to develop an acceptablemixed MOX powder for
sintering. In this regard, the minimum acceptable plutonia particle size in a MOX pellet is already
available based on earlier studies. Initial fabrication development will therefore concern itself with
the ability to meet the final fuel specifications and quality requirements in a number of respects
including density, grain size and Pu distribution specifications. The urania required for this
fabrication evaluations could be provided by GE or the laboratory could find its own source for
this supply. Specifications for the MOX fuel pellet will be provided by GE as well as the post-
.......: _ _sintedng ex:amination requirements.
• Demonstrate Fabricability of MOX with Gd
Although an extensive MOX fuel fabrication .data base exists, there is relatively little
experience in the way of fabricating MOX fuel with Gadolinia. GE has extensive experience in
fabricating Urania fuel with Gadolinia burnable poison and will provide the initial sintering
parameters which might be expected to work with MOX fuel. A major parameter of interest is the
oxygen overpressure during sintering to produce acceptable final chemistry. It will be necessary to
inspect and analyze fuel sintered over a range of oxygen overpressures to ascertain the correct
sintering atmosphere to be used in production. It is also proposed that a range of Gd-Pu contents
be examined under this evaluation. A maximum of 10% Gd (by weight fraction) is suggested as
the upper limit for the investigation. It is expected that the Gadolinia powder will be provided by
GE together with the specifications for this powder. Post-sintering examinations will be carded out
by the laboratory based on examination specifications to be provided by GE.
• Effect of Impurities on Processing and Sintering
Americium and other metallic impurities are expected to be present in minor or trace
quantities in the initial plutonium feed. There is a need to reduce their presence as much as
possible. High levels of Americium will tend to increase worker exposure. The processes to reduce
or eliminate them from the initial plutonium feed material are well established but need to integrated
2.3-12
with the overall process flow. It has been suggested that by building the fuel fabrication factory to
allow complete remote automated handling, the level of Americium that can be present could be
raised significantly. This does not however eliminate subsequent problems in handling the fuel
bundle transport to the reactor and receiving inspection where high Americium levds lead to higher
worker exposure. The proposed concept therefore aims at reducing these impurities to a low
enough level so they do not pose any problems downstream. In this regard, some fabrication
development and post-sintering examinations are required. In addition, it is also known that
Americium preferentially evaporates during sintering. Data from the initial runs will beuseful in
acc0tinting for possible worker exposure and designingpreventive systems to trap the evaporatedAmericium.
• Validation of Fuel Pellet Quality
• GE wilt be providing the fuel pellet specifications including those for density, grain size, impurity
limits and fuel chemistry. Post-sintering examinations should be conducted on sufficient pellets to
Validate the overall fabrication process and individual process steps to aid in fine tuning the MOX
Fuel Factory Requirements, Equipment Specifications and Process Parameters. Although sintering
parameters for MOX fuel are well established, a limited number of additional studies should be
conducted to optimize the process, particularly with a view to reducing the waste stream and scrap
recycle fraction which are the major contributors to worker exposure.
Fuel Mechanical/Chemical Perf0rmanc¢ and F0¢I Pro_nertie_:
• Mechanical/Chemical Performance
The objectives here are simply to verify the pin mechanical/chemical performance of the rod
and verify its integrity to goal exposures. Fuel rods will be fabricated and irradiated in research
reactors closely simulating the rod parameters expected in the disposition reactor. Post-irradiation
examinations will include among others: gathering fission gas release data and pin dimensional
(strain) measurements as a function of fabrication/operating variables; checking the migration of
specific fission products through gamma scanning;' fuel length change measurements; fuel-cladding
interface examinations for chemical compatibilit.y. The fabrication variables will include Pu-Gd
fraction, fuel power density and the range of allowable fuel pellet physical specifications.
• Fuel Properties
2.3-13
Properties of MOX fuel are readily available from previous DOE programs including the
Liquid Metal Fast Reactor Development Program. However, limited additional data are needed,
particularly for MOX fuel with Gadolinia poison. Material property correlation models for oxide
fuel containing small fractions of rare earth materials (of which Gd is one) have been developed at
GE, nevertheless, experimental confirmation of these data might be requested in licensing review.
Two Specific properties of major interest are: fuel thermal conductivity, and fuel solids, liquids
temperatures. Pellet Oxidation tests inoxygenated water might also be requested in order to answer
any questions relative tothe performance of a breached pin for the limited duration it remains in the
reactor prior to removal.
Fuel (Rod) No¢lear Pcrfgrmance Benchmark Data
' -The primary objective here is to confirm the nature of Gd bumup as a function of the radius
within an individual rod. Different Pu-Gd compositions could be examined. Rods can be irradiated
individually or in a cluster, in an experimental reactor such as the ATR. GE will provide the test
specifications and post-irradiation examinations and conduct the associated nuclear analysis. A
range of fission densities should be considered in the experiment. The results will provide the
benchmark data for confirmation/validation of nuclear codes and refine the nuclear design.
B. Preliminary Test Plan
Based on the foregoing objectives, the following test plan is proposed:
Fuel Fabrication Validation Tests
a. MOX Fuel will be fabricated by a DOE designated laboratory based on mechanically
mixed process, with Pu oxide from Pu feed stock prepared from weapons grade Pu or comparable
chemistry. The process specifications from the Pu Oxide phase to MOX fuel will be arrived at
jointly between GE and the fuel fabrication laboratory.The final specific_itions for the MOX pellet
will be provided by GE. The final specifications for QA for the fuel will be specified by GE. GE
will provide the initial input for sintering Gd bearing MOX fuel. The fabrication procedures will
also attempt to verify any key elements related to automated production/safeguards implementation
that may be required. For example, this could consist of active interrogation of the Pu content and
inventory logging in on-line computer systems.
2.3-14
b. Range of Fuel Pellet Parameters:
Different Pu-Gd compositions (Gd from 0 to 10% and Pu from 2 to 20%)
Nominal Pellet density: 96.5%
Nominal Grain Size: 10 microns
Specific Pu-Gd combinations to be specified by GE in Phase 2.
c. Post-fabrication tests for:
Optical Metallography (etched and unetched)
Fuel chemistry (O:M Ratio, impurities)
Pu homogeneity
Fuel density, description of porosity (by metallography)
Surface Roughness Post-fabrication examinations to be specified by GE; these tests
will be conducted on various batches to describe any effect of process variables on
final fuel pellet characteristics
d. Fuel properties tests:
Thermal conductivity (by laser flash technique)
Thermal Arrest Studies
(Above tests for various Pu-Gd compositions)
Fuel Creep for selected compositions
e. QA of reference fabrication process pellets
GE to specify QA requirements; laboratory will conduct the QA audit of the fabricated
The results of this series will be documented by the.laboratory with particular emphasis on
process verification for production of Gd bearing and non-Gd bearing MOX fuel.
In-Reactor Rod Tests
The in-reactor rod tests are aimed at meeting the nuclear benchmark data generation and
thermo-mechanical performance confirmation objectives. The rods will be assembled by the
2.3-15
laboratory and shipped to the test site. The test site will be responsible for the in-coming
inspection, thermal-hydraulic design of the test, for obtaining preliminary and final approvals for
the test and for the safety analysis based on any required input from GE and the fabrication site._
The detailed test specifications will be provided by GE. Either individual rods or a cluster of rods
will be irradiated in a research reactor. The power level and rod physical parameters (diameter and
length) will be provided by GE.
As currently envisioned, these tests could be individual capsules or a cluster of capsules.
None of the tests in this series have any active monitoring of either the temperature or for any other
parameter. Indirect confirmation of the temperature could be obtained by incorporating TEDs
(Thermal Expansion Devices) which have been used in the past. To the extent these tests are
entirely uninstrumented and passive, the cost of these tests should be low and an aggressive
schedule could be pursued. Data will be obtained as a function of exposure by removing the
capsules containing the rods at regular intervals, to be specified by GE.
Post-irradiation examination requirements will be specified by GE and will include, as
noted earlier, gamma scans, fuel length changes, rod profilometry, limited fuel pellet
metallography, fission gas collection, and limited microprobe examinations of the fuel-cladding
interface. These examinations will be conducted at a suitable facility to be designated by DOE. GE
will conduct Gd radial profile measurements in its VNC facility from shipment ef samples to be
made from the designated post-irradiation examination facility.
The details of the test matrix will be worked out after meetings between the interfacing
organizations. As presently envisioned, the test matrix will include the following parameters:
Fuel Characteristics:
MOX ROd without Gd for Comparison with Gd with Pu as variable
Pu-Gd Composition (Up to 10% Gd)
Fuel density, grain size within allowable range
Operating Conditions:
Exposure
Power Level
2.3-16
Because these tests are designed to be uninstrumented, fuel behavior as a function of
exposure will be obtained by removing the capsules after specified exposure levels rather than
monitoring pin behavior as a function of time while in-reactor. Individual rods or rod clusters
(depending upon the test cavity size) will be irradiated in a reactor such as ATR.
Assembly Tests
A limited number of full length MOX assembly tests are recommended, primarily to
confirm the performance of full length rods. These tests are not however anticipated to provide
any additional data for licensing. In conducting full assembly tests, it should be recognized that the
MOX fueled assembly will be placed in a sea of urania fueled bundles and the nuclear performance
of the reactor system will be controlled by the urania fuel rather than by the isolated MOX fuel
assemblies. A preliminary assessment has indicated that with the older reactors only partial MOX
• 'core loads are possible. In point of fact, the MOX fuel test assembly has to be designed to
minimize its impact on adjacent urania fueled assemblies and to operate within the limits of the
criteria for which the plant was originally designed. A number of design features in the ABWR, as
pointed out in the Phase IA report, allows it to accept a full core of MOX fuel while this is unlikely
to be the case with older LWRs. In addition to these factors, it should be borne in mind that if the
test were to be conducted in a conmaercial facility, it might be necessary to match the cycle length
of the specific core and an independent specification of the test fuel exposure may not be possible.
For these reasons, MOX fuel assembly tests in a reactor are not considered necessary for
licensability. Rather, it will be the confirmation of individual rod tests and their use as benchmarks
in the nuclear analysis codes that would be used for proving the nuclear and thermo-mechanical
performance of MOX fuel in an LWR.
If full scale MOX assembly tests are requested, two types of LTA assemblies are under
consideration and these will be examined in more detail at a later date. An assembly consisting of
all MOX rods could conceivably be inserted in the outer periphery and data could be obtained on an
all MOX bundle. A second type of assembly will utilize the so called "island" design whereby the
MOX rods are placed in the interior of the assembly and are surrounded by urania rods, so that the
LTA assembly's effect on adjacent fuel assemblies is minimal.
Full assembly tests could be used to study the effect of different Gd/Pu enrichments and
provide data on axial and radial power shaping in the assembly. Data will be obtained on rod
profilometry, length changes and fission product migration by gamma scanning, on a cycle-by-
cycle basis. GE will be providing the test specifications, MOX fuel will be fabricated at a
2.3-17
i
designated laboratory and assembled and shipped to the test site. Where urania rods are needed,
these could be provided by GE.
Facilities for examining full length rods are expected to be available at GE-VNC. A number
of interface issues relative to shipping, receiving and disposal have to be worked out, however,
these can be tackled once it has been determined (by the licenser/reviewer) that such full scale
MOX assembly tests are needed and agreement has been reached for inserting MOX bundles into
an operating BWR.
C. Interfaces
In order to carry out this program, a number of interfaces have to be established and the
activities coordinated. A PDR-LTA Interface Control Board may need to be established with
specific responsibilities and assignments and charged with the conduct of the program. The type of
interface arrangements that are suitable for implementing the lead test program was outlined in the
Phase 1B report.
D. Cost:
Detailed cost estimates can be generated after an initial licensing review to more clearly
define the areas where additional data are deemed necessary. It is anticipated that approximately
50 to 60 capsule tests might be needed. Cost of the program for fuel fabrication development and
for use of the test reactors at the interfacing sites will have to provided by the site chosen for these
activities. As a rough order of magnitude, it is estimated that the rod tests would cost up to $2
million at the te_t site including the post-irradiation examinations. The cost for fuel fabrication
activities will depend upon whether a facility has to be dedicated to this task or whether facilities
(glove box lines) where Pu is being handled is already available. Initial rough cost estimates are
approximately $2 to 2.5 million per year for a period of 3 years. The fuel properties data generation
tests are expected to cost approximately $0.5 million per year for a period of 3 years. The GE
resources needed for this program is estimated at 6 full time engineers and 2 senior professionals
for integrating all the external and internal interfaces.
E. Schedule:
Details of a proposed schedule were presented in the Phase 1B report. At that time it was
assumed that the lead testing program would be initiated in FY94. The previously suggested
schedule has been reevaluated and no changes are suggested for the duration for the various
2.3-18
activities. The following durations for the major milestones had been identified during Phase 1B:
(all dates after program start)
1.Produce initial MOX test fuel: 12 months
2. Complete Fuel Fabrication Development: 24 months
3. Complete Fuel Properties Measurements, if required. 24 months4. Initiate Rod Test irradiations: 12 months
5. Complete Rod Testing and Issue Validation Reports: 48 months
6. Initiate Assembly Irradiations, if required: 30 months
7. Complete Assembly Irradiations and Issue Confirmatory Reports: 72 months
Based on the evaluations presented, it is once again concluded that preliminary licensing for
MOX core utilization could be granted based on already available data and any remaining issues
will be answered with the completion of milestone 5 above or in 48 months after program
initiation. The full scale assembly test data are not needed for licensing but may be useful to refine
"the power shape of down-stream cores.
2.3-19
2.3.3 US INFRASTRUCTURE FOR LEAD FUEL TESTING
The infrastructure for implementing the disposition project, discussed in Section 4 of this
report, also covers several aspects related to the lead testing program. In summary, an extensive
infrastructure that is ready and capable is available to support the LTA program, specifically in the
following areas:
a. To define the Pu interface between the LTA fuel fabrication and pit material (LLNL,
LANL)
b.To identifytheprocessrequirementsandfabricateinitialbatchesofMOX (LANL)
c.To furtherdefinetheprocess,incorporaterequiredautomation/safeguardsverifications
(GF_ANL/LLNL)
d.To shipMOX fuelrodstoresearchreactorsites(thishasbeendemonstratedinother
DOE programs)
e.Testsitestoconductrodtests:Reactorsareavailable(ATR,HIFR,others)
f. Post-Irradiation Examination Facilities: (WHC, INEL, GE-Vallecitos, others)
It is only in conducting full scale MOX assembly tests that a suitable reactor site and the
necessary interfaces have not yet been firmly identified. Inquiries have been made with several
BWR vendors and it is clear that isolated full scale MOX assemblies could be irradiated in existing
reactors. Initial inquiries have also indicated that a similar situation exists for PWRs, in that there
are at least some reactors which would accept MOX fuel assembly tests. A number of interface
responsibilities have to be worked out including those for licensing the test assembly, hardware
procurement, exposure limits, post-irradiation examinations and eventual disposal. However, firm
commitments could not be obtained because of the highly preliminary nature of this project at this
time. Before proceeding further with full scale MOX assembly tests, it will be necessary to obtain
from the licenser/reviewer the specific objectives to full scale MOX assembly tests since partial
core loads of MOX fuel have already been irradiated to significant exposures in both BWRs and
PWRs, as reponed in the final Phase 1A report.
Further discussions in this regard have also indicated that with the potential for private
funding for the reactor system by utility/IPP groups, these groups will take the responsibility of
identifying the reactor to conduct the MOX assembly irradiation tests and working out any interface
responsibilities. An alternative, if the disposition were to be entirely funded by the US
government, would be to use reactors which were built with government funding.
2.3-20
2.3.4 European Infrastructure for MOX Testing
In general terms, the performance of a new fuel type must be proven by irradiation in a
commercial reactor. This is requirednot only to demonstratethat the fuel behaves satisfactorily,
but also to allow the validation of models for the processes which occur during irradiation.
Performance mustbe proven not only for steady-stateoperation but for a varietyof frequent faultconditions. However, since commercialreactor fuel contains no instrumentation,no data can be
produced. Data can be obtained in two ways: from post-irradiation examination (eitherat
poolside or in hot-cells), or by conducting irradiationexperiments on instrumentedfuel in .a test
reactor. Commercial reactor irradiation is also relatively benign, whereas the validation of
licensing methodologies must cover more sever operation. Specialized test reactor experiments,
e.g., PCI failuretests, mustbe mountedto addressthese issues.
'- ......... °--These general,principlesareapplicableto the testing of MOX fuel containinghigh grade, weapons
plutonium. Clearly,however, it would be inappropriateto regardsuch a material as a completely
new fuel type, since it is closely relatedto conventionalMOX fuel with which there is reasonablyextensiveexperience. Nevertheless, claimingthe priorexperience with conventionalMOX fuel as
relevant is only reasonable if, (1) design of the new fuel rods does not depart significantlyfrom
the base of experience with conventionalMOX fuel (2) manufacturingprocesses used to produce
the new fuel have a proven track record for producing conventional MOX fuel (3) it can be
demonstratedthat the differencesin plutoniumisotopic composition between the new fuel and (4)
conventional MOX (as well as any other differences) do not significantly change the fuel's
performance.
The second of these points is very important. It has always been recognizedthat the performance
of MOX fuel is much moremanufacturingroute specific than is the case with standardurania fuel.
Fuel microstructurein general, and plutoniumhomogeneity in particular, are found to be sensitive
to the method of plutonia/urania blending and pellet sintering, and other aspects of the
manufacturingprocess. Achieving high plutonium homogeneity-has long been recognized as one
of the key factors in producinggood MOX fuel. With high-grade,weaponsplutonium, this factorwillbe even moresignificant.
It mustbe assumedthatthe new fuel willbe manufacturedusing processes that have been proven
to produce highly homogeneous MOX fuel, and that the fuel rod and assembly design will not
departappreciablyfromproven experience. Given this, the qualificationprogram can be confined
in scope to studying the differencesbetween the new fuel and conventional MOX fuel. The
expected differences arise from the different isotopic composition. Weapons plutonium will
2.3.4-1
consist of typically 95 % fissile isotopes, compared with perhaps 65 % for material from a
reprocessing plant. The absence of the absorbing Pu-240 will make weapons MOX appear, from
a neutronics standpoint, closer to standard enriched uranium fuel than normal MOX fuel. For a
given lifetime average reactivity, weapons MOX is likely to give higher reactivites at start-of-life
and (because of less Pu-241 build-up) lower reactivities at end-of-life compared with conventional
MOX. This will help to reduce the problem of high late-in.life powers than can lead to high clad
corrosion and fission gas release in conventional MOX. The disadvantage is that rod power
peaking factors will be higher and burnable poison loadings (such as gadolinia doping) would
need to be increased. The higher fissile fraction also leads to the greater sensitivity to plutonium
homogeneity discussed above.
2.3.4-2
2.3.4.1 Single Rod Tests
Should single rod tests be required, either in advance of or in parallel with demonstration
irradiationsin a commercialreactor, then the Halden reactor is available as the ideal vehicle to
performthese tests. This is a smallBWR run by an internationallyfunded group of which both
BNFL and GE aremembers. The group have considerableexperience in runningirradiationtests
on MOX as well as uraniafuel. Theirparticularstrengthis the development of rig designs and in-
pile instrumentationwhich can elucidate the real processes occurringduring irradiation. Several
of their most recent developments have been focused on obtaining data on high burnup, fuel
without the needfor the long dwell times that are normallyrequired.
Comparative irradiations (Urania versus conventional MOX versus weapons MOX) of a small
number of rods could be mounted quickly. With appropriate in-pile instrumentation, data on
......•"_hermal-performance, dimensional stability and fission product release could be generated.Performance data would become available as soon as the test was loaded. Such a test could be
giving important insights into the behavior of the new fuel many years before a commercial
reactorirradiationandPIE programcould cometo fruition.
2.3.4-3
2.3.4.2 Commercial BWR Irradiation
The key step in qualifying a new fuel type is to conduct the irradiation of one or more
demonstration assemblies under prototypic commercial reactor conditions. However, since such
an irradiation does not in itself produce any information with which to draw a conclusion, it must
be followed by a program of post-irradiation examination (PIE).i
It is first necessary to identify a suitable commercial BWR in which a trial irradiation could be
undertaken. There are currently 23 commercial BWRs operating in Western Europe. Their
capability with regard to irradiation of MOX fuel is summarized below:
Reactor Utility MOX Lieense Status
Oikiluoto 1 TVO License not applied forOikiluoto 2 TVO License not applied forSwedenBarseback ! Sydkraf_ License not applied forBarseback 2 Sydkrah License not applied forForsmark I FKA License not applied forForsmark 2 FKA License not applied forForsmark 3 FKA License not applied for
•Oskarshanm ! OKG License not applied forOskarshamn 2 OKG License not applied forOskarshanm 3 OKG License not applied for
Ringhals 1 SSPB License not applied for
Brunsbuttel KKB Submitted application in 1986Gundremmingen B KGV Submitted application in 1989Gundremmingen C KGV Submitted application in 1989lsar 1 Bayernwerk Submitted application in 1989Krummel KKK Submitted application in 1990Philippsburg I Badenwerk License not applied forWurgassen Preussen Elecktra License not applied forSwitzerlandLiebstradt KKL First MOX load in 1998Muhleberg BKW License not applied for[-lollandDodewaard GKN Has current MOX license
Cofrentes HE License not applied forSanta Maria de Garona Nuclenor License not applied for
2.3.4-4
It is clear from the previous summarythat in Finland, Sweden and Spain the utilities have not
taken any steps towards using MOX fuel; approachingthese utilities is thereforeof littlevalue. In
Germanyand Switzerlandsome utilitieshave applied to use MOX fuel, although none are likely
to be ableto load it early enough to be of use in the dispositionstudy, unless the licensing process
were to be accelerated. The Dodewaardreactorin Holland has already operatedwith MOX fuel,
but its smallsize bringsits suitabilityinto question.
Following this initial surveythe utilities at KKB, KKK, KGV, Bayernwek,.KKL and GKN havebeen contacted to establish their views on performing MOX lead test assembly irradiations. The
resultsof this are given below:
1. KKB and KKK (Brunsbutteland Krummel)
Contact:- Mr. Rigerof H.E.W.
•Mr. gieger stated that the Krummelreactorhas many differentfuel types within its core and
they are always interested in testing new fuel designs within the reactor. As such the
nsertion of GE MOX fuel would be an attractiveproposition to them.
However neither Krummelnor Brunsbuttel currentlyhave a MOX fuel operating license. A
public inquiryfor the Krummelwas planned for November 1992 but this was canceled and
to date there has been no furtherprograms. His view was that it is unlikelythat a MOX fuellicense will be obtainedbefore 1998.
He stated that H.E.W. were still interested and he has been requested to supply furtherinformation on the facilities available at the Brunsbuttel and Krummelsites.
2. KGV - (Gundremmingen B&C)
Contact:-Mr. Passig (RWE)
Mr. Passig stated that the license to load MOX fuel into Gundremmingenwas expected in
January 1994. However he did anticipate some problems in loading US DOE MOX fuel
assemblies into German reactors. The main problemfor RWE (and other German utilities)
was that they have to put priority on the utilization of plutonium arising from their
reprocessing contracts. Nevertheless he did state that he was interested in the concept andhis colleague Dr. Dibbert has been contacted with a view to provide furtherinformationon
the Gundremmingensites.
2.3.4-5
3. GKN - (Dodewaard)
Contact:-Mr. Van der Hulst
Mr. Van der Hulst stated that under normal circumstancesDodewaard would be ideally
suited to accept MOX fuel assembliesfor GE. However Dodewaard are experiencing
severe difficultieswith the Dutch licensingauthorities. Thishas resulted in a compromised..
'position being reached_underwhich a temporary-license.has been granted for continued
operationof Dodewaard. This license was restrictiveto the extent.that.GKV were ob!iged
to removeMOX fuel from the reactorandthey will only be allowed to load uraniumfuel inthe nearfuture.
4. BaYernwerk- (ISAR I)
Contact:- Mr. Huber
Mr. Huber stated that although a MOX operating license had been requested for ISAR 1
(BWR) Bayemwerk are not pressingfor its issue as they have decided to load MOX fuelonly into theirPWRreactorat ISAR2.
5. KKL- Liebstadt
Contact:-Mr. J. Afonso/Dr.Patak
Dr. Patak stated that the testing of leadMOX fuel assembliesfor GE would be an attractive
proposition to them. Mr. Afonso has been requested to provide additionaldetails of thefacilitiesavailableat the Liebstadtsite.
2.3.4-6
2.3.4.3 PIE Program
If a suitabledemonstrationirradiationin a commercialreactorcan be accomplished,it would need
to be followed by a PIE program. The aim of the program would be firstlyto confirmthat the
fuel had behavedsatisfactorily,and secondlyto studyindetailthose aspects influencedby isotopiccomposition. If rodscontainingburnablepoisonmaterialsare used then these wil!lalso need to be
includedin the PIE program. A typical programmightstudy perhapssix rods _nd consist of the
following:
Pool-Side Work
- visual examination
- ultrasonicleak testing
- measurementof assemblygrowth
- measurementof rodbowing- selection, removaland decontamination of chosen rods
- ECT measurementof clad oxide layer thickness
- transferof chosen rods to transport container
TransportfromReactor Site to Hot-Cell Facilities- hot-cell work
- fullvisual examinationof each rod
- measurementof rod growth
- gammascanni,_gof each rod
- profilometryof each rod
- puncture, gas analysis, internal pressure/volumemeasurementon each rod
- density measurementson selectedfuel samples
Sectioning at SelectedLocations;
- optical microscopy (showing grain structure, porosity distribution etc)
- alpha autoradiography(showing plutonium homogeneity)
- EPMA/SRF (givingelemental/isotopic distributions,fission gas retention etc)
- chemicalburnup analysis
Disposal of allWastes
2.3.4-7
2.3.4.4 Ramping Program
To assess the PCI resistance of a new fuel type, specialized ramp tests are required. Although
such tests may be judged unnecessary in the case of weapons MOX, if they were to be required, a
suitable test would comprise:
- transport of selected rods to hot cell facilities
- non-destructive and destructive examination of parent rods to include profdometry, gamma •
scanning, puncture testing and g_s analysis
- refabrication of each parent rod into typically three short rodlets_each rodlet to be pressurized
and sealed and neutron radiographed
- ramp testing of each rodlet, consisting of irradiationat a specified conditioning power for a
specified time, followed by ramping at a specified rate to a specified terminal power
...... _--_post-irradiationexamination of selected rodlets to include visual inspection, profilometry
neutron radiography, gamma scanning, internal volume/pressure, gas analysis, cerrnography
and density
- disposal of all wastes
2.3.4-8
2.4.1 FUEL FABRICATION FACILITYSYSTEMS REQUIREMENTS &ARCHITECTURE
Objectives:
A dynamic model of the plutonium inventory flow through the overall Pu dispositioninfrastructure is tobe developed with a focus on the Fuel Fabrication Facility (FFF), itsrelated accountability methods, and its external system interfaces.
The intent of planned model evaluations is to identifyaU system level customer and derivedrequirements which drive the. FFF functional-design and time related performance
" specifications. Then, dependent upon the Pu metal conversion and mixed oxide fabricationprocesses selected, alternate levels of process equipment replication to meet throughputrequirements, and all Special Nuclear Material waste streams with their relatedaccountability methods shall be evaluated.
The overall functional process timing model will be grown to cover the plutoniumdisposition from weapons to a DOE permanent disposal site for the irradiated reactor fuel.'Model details will be of a depth sufficient to address the FFF issues. The functional modelwill allow a consistent trace back to the driving systems level requirements, as well asdefining a consistent set of system interfaces and the allocation of functions to specificcomponents.
Status:
The functional model of the FFF has been defined and executed at a level of detailsufficient to identify and evaluate the pellet/rod storage requirements and the significant,fuel related scrap/waste streams. The RDD model is currently being checked for physicalconsistency and completeness. Once this task is complete, the design and manufacturingstaff will be engaged to assure realistic inputs to the simulation.
The FFF functional structure is based on past industry best practices, including maintainingphysical separation between the gadolinium and non-gadolinium fabrication processes.Hence, the FFF has a minimum of two independent fuel fabrication lines which, based ona GE9 bundle design, would be of equal capacity. The FFF design will be based on meetingthe ABWR reload core requirements for throughput; which can range from 324 to 972 fuelbundles per year, dependent on the reactor strategy selected (see section 1.0). To clarify theexternal interfaces and initiate the discussions with thedesign and.manufacturing staff wehave proposed an initial set of derived FFF requirements as follows:
• all materials required for fuel bundle fabrication shall be received in aform suitable for immediate use,
• all materials shall be made available as required to meet fabrication schedules,
• all process equipment shall be continuously monitored for acceptable performance
2.4.1-1
and, if necessary, adjusted on-line to maintain acceptable performance,
• process equipment lifetimes shall be greater than FFF usage requirements, or theirreplacement impact on fabrication schedules shall be minimized,
• all glovebox filter change operations shall be performed on-line,
• fuel accountability procedures shall be automated using current state-of-the-artequipment and processes,
• ..... fuel _accountability procedures •shall, not-havea.significant impact on fabricationschedules,
• fabrication process formation of fuel scrap material shall be minimized, and shall bebetter or equal to recent LWR fuel fabrication experience,
• fuel scrap wet recovery process shall y/eld a minimum of 99.5% of plutoniumback to the FFF,
• the storage area for completed fuel assemblies shall be sized for a full fresh core,
• FFF generated Waste shall be minimized
The current functional model of the FFF is presented in Figures la through lf, whichfollow the RDD decomposition of the system definition to the pellet fabrication process.The figures are the RDD-100 Behavior Diagrams (BDs), which are consistent with the database elements and the performance attributes. Table 1 provides explanatory notes for theBD' symbology as an aid in interpreting the functional model. The RDD model is beingchecked by simulating the production of a one fourth core reload, or 58 fuel bundles. Theemployed manufacturing strategy was to operate two independent pellet and rod fabricationlines, by campaigning the number of rods of each enrichment type, in ascending Puenrichment. Fuel bundles were _,ssembledas soon as the appropriate mix of rod types wereavailable. The GE9 fuel bundle is composed of 60 rods of the following types:
Rod Type Pu Content Gd Content Number of Rods
1 1.0% -- 4i i
3 1.6% -- 8
5 2.3% -- 8
6 2.8% -- 8
11 2.8% 1.0% 4
12 3.2% 1.0% 8
13 4.2% 1.0% 20
2.4.1-2
Sources of fuel-containing scrap were identified from previous LWR fabrication experienceand are indicated as the following mass fraction of fuel material throughput:
Pellet line operations: pellet pressing -> 0.008pellet grinding -> 0.024
Rod line operations: pellet loading -> 0.006
Glovebox coarse and HEPA filters, as well as vacuum bags were assumed to be consumedat fractional rates based on throughput of fuelpowder, and whenever a new enrichmentcampaign waslnitiated. A spent filterwasassumed to contain about 0.2Kg of fuel materialwhich was 99.9% extracted from the filter body before the filter was assigned to the lowlevel waste category (LLW).
The results of the simulation are shown in Figures 2 through 5. The top level results aresummarized by Figure 2. The curves labeled "generic pellets" and "generic rods" depict thetotal amounts in interim pellet and rod storage areas, summed over all fuel types. Thisparameter shall be used to size the FFF storage areas, and shall be the subject of futuretrade studies. The curves labeled "fuel bearing scrap" and "Total wet scrap" are the timedependent and integral of the fuel bearing scrap sent to the wet recovery process. Figures3 and 4 show the fabrication time lines for the different pellet and rod types, and their"generic" storage requirements.
The LLW curve values are presented in cubic feet, with compaction factors applied to thegenerated waste, which includes filters, bags, rags, Pu shipment cans, rejected fuel tubing,etc. The factors for compaction are currently arbitrary, and need to be revised. The timedependent, fractional life consumption of filters is depicted in Figure 5 for the mox fuelline, it is very similar for the gadolinium line. It is currently based on assuming that theglovebox operations generate 0.1% of the material throughput as spills or airborne dustwhich is collected by the filters and vacuum bags. Vacuum bags are assumed Io accumulateup to 0.25 Kg of fuel material during their useful life; realistic values will be determinedprior to any trade studies.
2.4.1-3
y !I
T. ""_' I Averticlelinerepresentsaprocess.i , ! _me_ downward,sequentially AIoopflowiseitheropenendedm toeachfuncl_n. orcontrolledbyanilera_logic,e _...do.,,
Funotlon In loop Oran'end_' exR, Ii I I ' ",no.. =or; i o_
_, Aparallelsetofprocesseshas tZ _, I • ! independenttimeflowwhichawaitsI ,.:-,I I--._.Z--Icompletionofeachprocessati I I, I itsconfluence.
i,Functions can outputitems which are either
real objects and/orinformation. Inputitems can be used totrigger functions.
jr 0._,)multi-exitfunctionisanOR output
branch,Theexittakenisdependentofthefunction
Table1,AnAidtoRDD=IO0BehaviorDiagramSymbology2.4.1-4
I decomposethisfunction I t
in Fig.lb
Fig.la. FunctionalModelfor:DisposeWeaponsPlutoniumviaReacto_r
newfuel reactorsite power "_make tritium
RequestZBUF 3.3
Output Tritium• Fuel Rx Electrical woduction
Pu02shpmt Fabrication Power (Option)
98
& ZBUFXfer HLWCask
J
decompose this function
in Fig. lc
Fig.lb-FunctionalModelfor:TransitionPu02toHighBurnupFuel
I0 _7
Io , ,_
decompose this tun_ioin Fig. ld
Fig.lc-FunctionalModelfor:FuelFabrication
I_iii::i::!i!!iiii!i!!!!...... :,,.. ,:,,!!!iiiii!i!!'.:,::!!iiI % i I n Itl m,I I_-e I
_iiii!i!ii i :ii_iiiilii!i!iiiiii iiiii::::!!i !ii!i_i_i/ m ox iI n _ I" : I .] il
813' _"
t ) _'---._ e.l .t.e... I
• ...... 18.1
.7 _ ..,,.: " )
_',i'_iii,,',!i__i',ii',ii',_ I_,__._I
\ \ __-..___ _,.,-.I.l.l.e... 11I.II _ 3 "1"1"e_', . _ 1e.I
\ i _,r,.,.,_'r°,,n,.r.w.,1 ----1,,,,,,oxt___.,r..r.__,.L____..___-I _ J I ,_,r _ _:._.'.'._I ,,,,,,,pies II '"_ '_N°x Il """" I
etorage
:1.1.1.¢1... fab.10
r_h,.,,,... ,,,.l
• ._._ .e... 21,1 _._ .1 .e... II1 .e
_,,d po,o, Request mox._..., _ .,.,.,,,,r, ,-...-.,.
Fig,lf- FunctionalModelfor:FabricateMoxPellets2.4.1-:10
I
ABWR Pu Phase Ic by Switick " 30_December 1993 at." 8.'22:46 am
Resource (LL'W pu/u vol)l20__
Item vs. Item View I Re'source (Total wet 's(:;ap)_ 10 -.-I.,500- 400 -- l 16 -4
14-.1450- , i 350- __12 -l4oo-] 300- _ _;lo 4 ....................._i-!ii:-!i!i!iii!iii_!i_i-!i!35o-} .-. " -"'___!_ _ 8 4 _i_':_::_;_'_:::.-=_i:_i
-_.oo--] _i50- -I _i!i:i-!i:i!i!i!i:iii!i_!i_!::_::-!i-!i!i!:i::_-. _'oo- 6 --t _!;!_!C!_!_!_!;!_!_!::!=_::!_ii!_::=--!i!::i_!::i::!_)_!ili!i!i!i!i_i_=,<.5 0 -_ ) " -- •===============================================================i _: =i=:::=::::::::::::::::::::::::::::::::::::::::::::::::::::::::::::I : "=============================================:'::::::: =::::==:::==: 4 ==============================================================================
" =========================================================:: ===::=-:=:=: 3 ================================================================================:_15o_ ......................................................................................................................".................."........":......-,oo_ -,oo_. _i!iii!_i;i!!_iiii!_i;!i_i!_;ii!:_ii_!i_i{!_{!_ii!_i!_iiii!iii!__:_::i_i_i_i_i_i_iii_ii_!]_iii_ii_;ii_i_;iiiii_ii_!_i_i_i;4!ii_!_ii!_!i_i_i_;o _(, _i;ii!!:;i;;;i!;i;i;;;i_i;!;i_i_!i_i:i_i_i_i_i:i_i!_;_;_:_;_i!;_i;i:i:!;i;;_n_.oo ,_o _6oo2ooo2_oo
- 50- IRe_our_e(_02u_e_ca,,_)I50--__ 0 --" -1 160_ -0
0 _ _. & 4 _ 6 0 400 800 1200 1600 2000 2400| 4
iltem vs. Item viewI MOX rod type Resource. (MOX bearing scraPl, I / 120J ===========================================:=::==::===:::=....
_oo_. _,_ _!_:i!_i_!i_!!i!i_i_- _oo ...._:.:::i:i!:_:i!:.:.:_:_....
_i .:.:.v.:._.-:-=-:i:-=-:--:_-=::.:700 -4 22 _ =============.=====:=-:':=:=':=::'===.:::-1:======-========6oo-4 :_o
o_ " 18
:_ $.oo4 ] 14
• x - [ _10
,_ =0"00"4_ !IJi'l-I]_J JJ IIH]]]_JJ ll_JJ .I;_@i_!_!il_i_!i!!ii,; :i..,. _ _ ' ,_our¢_)O_i_i_i!!!_i_i_i_i!i_i!i_ii!i_i!i!!_i_iii_i_i!i_ii_i_i_i_i_i_i_i_i!iii!i_!_!i!i_i_i_ii!i_i_i_i!!!_iiii_i_i!!i_(FreshFuel _)undle)'rO 1600 20-00 24:00
"" 200--1, (5
1 O0 -4 2 60- !=_ . _
..=
0 '*10 11 12 13 1" 0 400 800 1200 1600 2000 2400 50 -J,
MOXG rod type time I40 ..4
-,i.
',30 ..4
-t.20 ..4
J10--4
-,,i.
1500 1700 1900 21 O0 2300 2500
[ time
Fig. 2- General Characteristics of RDD Simulation: Fabricate 1/4 Core Reload
2.4.1-12
2.4.1-13
2.4.2. Plutonium Feed Material Interface
As weapons are retiredfromthe existing nucleararsenal, the warheads are removed and shipped
to a disassemblylocation where the nuclearweapons components or "pits" are separatedfrom the
warheads and placed into storage. The plutonium in these weapons components is removed and
either refabricatedinto new warheads, stored as part of the strategic reserve_or identified as
excess and placed into storage for non-militaryuse. A plutonium-hydride/dehydrideprocess is
being developed and demonstrated by national labortorypersonnel for removing the plutonium
fromthe "pits"in retiredweapons. The four main advantagesassociated with this process are (1)
plutonium can be easily removedfrom the tight fitting"pit" configuration, (2) no additionalwaste
is generated by the processing, (3) the plutonium configurationcan be easily unclassified, and (4)
consolidation into metal ingots provides a smaller, more compact form for storage. In this
process, a "pit" is cut in half and each half is treated separately in a reaction vessel. A small
amount of hydrogen is introduced into the reactor chamber at an elevated temperature. The
hydrogen reacts with the plutonium, forming plutonium hydride which spalls from the "pit"
(volume change during the reaction) and falls into a collection vessel. In the process being
developedbyLawrence Livermore National Labortory(LLNL), hydrogen is added to the reactor
until all of the plutonium has been converted to hydrideand fallen into,the collection vessel. At
that point the hydrideis heated to drive off the hydrogen(dehydrided),the remainingplutonium is
meltedand cast into a metal button or ingot. In the Los Alamos National Labortory (LANL)
approach, uranium hydride is used to provide a small amount of hydrogen for reaction with the
plutonium to form the hydride. The hydride falls into a heated collection vessel where the
hydrogen is driven off and flows upwardto the top of the reactorhydridingmore plutonium. This
cyclic process continues until all of the plutonium is converted to hydride and spalls from the
"pit". The plutoniumin the collection vessel is meltedand cast to forma metal ingot. The reactor
vessels are relativelysmall for the hydride/dehydrideprocess and readilylend themselves to glove
box applications as well as automation.
The demonstration of the systems necessary for the disassembly of retired "pits" is being
completed in the Automated Retirement and Integrated Extraction System (ARIES) program
currentlyunderway at LANL. (Ref. 2.4.2-1) The conceptual design and prototype testing phases
of this activity have been completed and the long lead time equipment items have been ordered.
The need for the storage of the plutonium recovered from the retired"pits" has been identified
and LANL personnel are currently developing the requirementsfor ttfis facility. (Ref. 2.4.2-2)
This facility must be capable of handling the retiredweapons plutonium as well as plutonium
residues fromvarious weapons sites located thought out the country. As a result, four different
2.4.2-1
storage forms("pits", oxide, metal and stabilizedresidues) are being considered for this facility.
Ideally"pit" storage would be eliminatedfromthis facility because theirclassified naturerequires
additional security and control measures. The metal could be stored irrespective of impurity
content and the oxide would be accepted only after calciningso the loss on ignition at 1000 C
would be less than0.5 percent. As in the case of metal,no impuritylimits have been defined for
the Oxideform. The storagerequirementsfor thestabilizedresidueshavenotbeen finalized. -....
Both LLNL and LANL have indicated thereshould be no problemconvening either the hydrideor metal form of plutonium to an acceptable oxide, see Attachment 2.4.2-I for a draft
specification. Each laboratoryhas some historical experiencewith convening metal to oxide, but
additional testing will be requiredto evaluate the morphologyof the resultingpowder. Blending
tests are also suggested to evaluate the plutonium homogenity of the material resulting from a
physicalblendprocess. Although the presenceof galliumas an impuritydoes not appear to be a
problemfrom the neutronics standpoint,the effect of gallium on the physical behavior(thermal
conductivity,fissiongas release, irradiationdamage)of the fuel in a reactormust be evaluated if it
has not been removed during processing. If a problem is identified, technology must be
developed and demonstratedto remove the gallium from either the plutonium hydrideor metal.
Uncertainties concerning the powder morphology and the overall effect of the gallium in the
irradiatedfuel need to be evaluated andsolutions developedshoulda problembe uncovered.
The mixedoxide fabricationprocessbeing considered for the PlutoniumDisposition Study (PDS)
is based on proven glove box technology with automated operation and hands-onmaintenance.
In order to minimizethe worker radiationexposure during fabrication,the americium shall be
removed from the plutonium feed material. The hydride/dehydriderecovery process described
above contains a step for americiumremoval just prior to the casting of plutonium ingots.
Although this approach may be acceptable if the plutonium is fairly old because most of the
plutonium-241has decayed to americium-241,the lowest exposure plutonium dioxide feed would
result if the americiumis removedjust prior to oxide conversion. This scenario which results in
the lowest exposurefeed materialthat minimizesthe plutonium recyclingactivities suggests that
unless the plutonium is to be used as the oxide within a relatively short time, the plutonium
storage form that best fits with the PDS is unpurified plutonium metal. In this scenario, the
plutonium is removed from the "pits" by the hydride/dehydrideprocess, cast into ingots and
placed in storage. The plutoniumidentified for PDS activities is removed from storage, melted
and the americium removed. During these subsequent steps, any impurities could be removed
from the plutoniumby either electrorefiningor vacuumdistillation. The resultingplutoniumcould
2.4.2-2
be converted to oxide by an oxidation technique and shipped to the mixed oxide fabrication
facilityidentifiedfor the PDS mission.
References
2.4.2-1. W.Dworzak, al et., "ARIESConceptualDesign Report,"Los Alamos National .......
Laboratory,June 1, 1993, (NMT-DO:(U)93-041).
2.4.2-2. "PlutoniumStorage FacilityLead LaboratoryInterface Document," Los Alamos
National Laboratory, September, 1993.
2.4.2-3
Attachment 2.4.2-I
DRAFT SPECIFICATION
Pu02 Powder, Ceramic Grade- Dry Process
1.0 SCOPE
1.1 This specification establishes *,herequirements for ceramic grade plutonium dioxide
powder produced from plutonium obtained during retitrement of nuclear weapons.
2.0 APPLICABLE DOCUMEaNTS
2.1 The following publications form a part of this specification to the degree indicated
where applicable:
ASTM B214°1964 "Sieve Analysis of Granulated Metal Powders"
ASTM B329-1961 "Test for Apparent Density of Refractory Metals
and Compounds by the Scott Volumeter"
ASTM B330-1965 "Average Particle Size of Refractory Metals and
Compounds by the Fisher Sub-Sieve Sizer"
When the contents of this specification conflict with any document referenced herein,
the specification takes precedence.
3.0 GENERAL DESCRIFFION
3.1 Pow.der Description
The material to be furnished in accordance with the specification shall be plutonium
dioxide ready for fabrication of mixed (uranium-plutonium) oxide fuel pellets. The
supplier shall furnish all required reports and information defined in this
specification. The supplier shall perform and report all tests required by this
specification.
4.0 REQUIRF_aMENTS
4.1 Functional Criteria,
Purchaser will specify the nominal plutonium isotope compositions. The
requirements of this specification are for material in the form of homogeneous
powder. For this specification, homogeneity is defined as follows: A homogeneous
SpecificationRevision 1/11/94
2.4.2-4
DRAFT SPECIFICATION
Pu02 Powder, Ceramic Grade- Dry Process
lot shallconsist of all materialproducedfor a single shipment in a continuous processsequence; or it shall consist of a uniformblend of material.,producedin a batch-typeprocess; and it shall consist of an amountof material produced with constantprocessparameters. Successful processing of this material is dependentupon the consistencyof characteristics. This specification provides for a wide range of limits for certaincharacteristics. Once the supplier establishes a specific value tor each characteristicwithin the wide range, he shall maintain this value within the specified narrowrange.The limits of the narrowrange selected must fall within the wide range. The narrowlimits shall include both materialand measurementvariations.
4.2 productSpecificationAll plans and procedures for obtaining (1) homogeneity, (2) samples for chemical
analysis, (3) chemical composition, and (4) physical properties shall be submitted toPurchaserfor concurrence.
4.2.1 Powder Composition4.2.1.1 PlutoniumAnldysis
Plutonium minimum 85 w/o of plutonium dioxideIsotopic Content Pu238, 239, 240, 241,242 as specified
4.2.1.2 The total volatile content of the powder shall not exceed 0.05 w/o plutonium dioxide.The total volatile content shall be determined by weight loss on heating a one gram
sample at 1000 + 25 Celsius for four hours in air or by an equivalent method
approved by Purchaser and agreed to by the Supplier.
4.2.1.3 The plutonium shall be in the form of plutonium dioxide produced by oxidation ofplutonium metal at or below 500 Celsius.
4.2.2 Imt)urities
4.2.2.1 Total impurities, shown on Table 1, shall not exceed 2,500 ppm of plutonium.4.2.2.2 The americium content shall not exceed 300 ppm of plutonium.
4.2.2.3 The uranium content shall not exceed 2500 ppm of plutonium.4.2.2.4 The total measured impurities in Paragraph 4.2.2.1 shall be such that:
I;Ci(B=) is equal to or less than 4.0 ppm by weight of natural boron excludingAm-241 where:
SpecificationRevision 11/09/93
2.4.2-5
DRAFT SPECIFICATION
Pu02 Powder, Ceramic Grade- Dry Process
Table 1. TOTAL IMPURITIES
I ,, Illll Ill II
IMPURITY LIMIT PPM OF PuIllll I I I I IllUllll IllII I II III
Aluminum 400Boron 1Cadmium 1Calcium 250Carbon 500Chlorine 25Chromium 150Cobalt 75
Copper 400Fluorine 130Gallium 400Iron 400Le_d 400Maganese 200Magnesium 200Molybdenum 400Nickel 400
Nitrogen 100Silicon 200Silver 25Sodium 400Tin 400Titanium 200Thorium 10Vanadium 400Zirconium 400
ZDyspr0sium_ Gadolinium_ Europium_Samarium 2
SpecificationRevision 11109193
2.4.2-6
DRAFT SPECIFICATION
]Lh,02Powder, Ceramic Grade- Dry Process...w........
Ci ffi Weight fraction in units of parts per million parts of Pu of eachimpurityin Paragraph4.2.2.1, and
(13-)- The naturalboron equivalent of element i from the Table 2 list,which is based on the neutronabsorptioncross sections in BNL-325 S1, and "ResonanceIntegral Data," ANL Newsletter No. 1,assuming a Maxwellian spectrum in the resonance region for atypical thermal reactor. The atomic weights are on the physicalscale. When impuritiesare reportedas less than a stated thresholdof detection, the thresholdvalue shoalbe used for this calculation.
4.2.3 Physical Pro_vertie_4.2.3.1 All material must pass througha 325-mesh U.S. Standardsieve in accordance with
ASTM B214-1964, "Sieve Analysis of GranularMetal Powders".4.2.3.2 Bulk Density
The bulk density shall be no less than 1.0 gm/cc as determined by the ScottVolumeter per ASTM B329-1961, "Test for ApparentDensity of Refractory Metalsand Compounds by the Scott Volumeter". Once the Supplier has established a value
for bulk density, he shall maintain this value within +0.25 gm/cc for the totalamountof materialsupplied underthe purchasecontract.
4.2.3.3 Surfacf Are_
The surface area shall be determined by the Supplier by the B.E.T. method ofanalysis or an equivalent method approved by Purchaser. This requirement forsurface area shall not be the basis for the acceptanceor rejection of material, but isrequestedfor informationonly.
4.2.3.4 Particl_Siz_
The particle size shall be determined by ASTM B330-1965, "Average Particle Size of
Refractory Metals and Compounds by the Fisher Sub-Sieve Sizer" and/or equivalentmethod approved by the Purchaser. This requirement for panicle size shall not be the
basis for the acceptance or rejection of material, but is requested for informationonly.
SpecificationRevision 11/09/93
2.4.2-7
DRAFT SPECIFICATION
Pu02 Powder, Ceramic Grade- Dry Process
..................... , ._ - _
Table 2 BORON EQUIVALENTDATA
IIIIIII II IIIIIII I II
Boron Equivalent Boron EquivalentElement l_m B/ppm of Element Element ppm B/ppm of Element
I IIIIIIIII I I I I I IIIII I III IIII
'Aluminum 1.32x10"4 Magnesium 1.42x 10.4Boron ' i.00' Manganese ,,,i,,' 4101x10"3 .....Cadmium 7.79x 10-! Molybdenum ......1.32x 10-3 ....Calcium ..... 2.87x10 "4 Nickel l i20x10"3
Carbon 5.51X10"6 ...... Nitrogen ......... 2.76xi0 -3 ...... .......Chlorine 1.37x10 '-2 Samarium 0.524Chromium ' 9.35x10 "4 ' Silicon 8124x10"5 'Cobalt 1.25X10"2 ..... Silver ' 2,98X10'2........
'Copper 9.45x10 -4 ...... Sodium 3.37x10 "4Dysprosium 9.7x 10"2 ' Tin _, 2106x10"4
0.434 Tungsten' 7102x10-3Fluorine 3.48x10_ Vanadium 1.49x10 '3 ¢
, , ,,,
Gadolinium 4.191 Zinc 3.71x10 "46allium 4.000x10 "6 Zirconium 2.93X10-:5Iron ' 7.43x10"4 .......
,,, ,,,, , ,,,,
Lead 1.23x10°.......
4.3 QualityAssurance R_uirements4.3.1 Concurrence of Purchaser is required, where indicated in this specification, of the
Supplier's Quality Control Plans and Fabrication Procedures prior to the productionof any material. This concurrence may involve the witnessing of tests and test
equipment at the Supplier's plant. Any significant change in process operations bythe Supplier is to be made known to Purchaser to permit joint performance
evaluation. Such changes may be accidental or planned.
4.3.2 Check Analysis SampleThe Supplier shaU take a sample from each lot of material sufficiently large that the
Purchaserwill receive a 20 gram sample. This sample is to be divided into four parts
for (1) chemical analysis by Purchaser, (2) chemical analysis by Supplier, (3)
SpecificationRevision 11/09/93
2.4.2-8
DRAFT SPECIFICATION
Pu02 Powder, Ceramic Grade- Dry Process
chemical analysis by Referee if necessary, an, (4) archive sample held by Supplier,until shipper receiver differences, if any, are settled. Thesample for Purchasershallbe packaged,-marked "CheckSample"and sentto Purchaserwith the lot shipment.
4.3.3 Referee
In the event the Supplier and Purchaser fail to agree as to the material producedmeeting any attributeof this specification or other requirements of the order, theyshalljointly select a thirdpartyand the method and degree of retesting. The findingof the third party shall be final. The partywhose value is farthestfrom the referee'svalue shall bear the cost of the refereeinvestigation.
5.0 INSPECTION
5.1 The Suppliershall submit to the Purchaser(prior to shipment) a Certificateof Test intriplicate for each homogeneous lot of powder showing that the material conforms tothis specification. The certification shall include the purchase order number,
Purchaser specification designation, the results of the required tests, and a statementcertifying that the material is homogeneous. The test results shall be so numberedthat they can be identified with their related lot of material. The certifications shallinclude the results of the analyses per Paragraph 4.2, including the limit of detection
for each analysis. This section shall include calculated total boron equivalent per theequation in Paragraph4.2.2.4, if applicable.
5.2 Production material is subject to return at the Supplier's expense unless theCertification of Test accompanies the shipment.
6.0 PACKING, MARKING, AND CRITICALITY REQUIREMENTS
6.1 The material shall be packed in a manner which will prevent powder loss andcontamination spread during transit. Shipments must adhere to applicable UnitedState Department of Energy and/or United States Depa/tment of Transportationshipping regulationsand licenses.
6.2 The shipping containers will be decided upon jointly by the Purchaser and theSuppliersuch thatPurchaser'slicense is not violated. The Purchasermust be notifiedand approveall pending shipments.
SpecificationRevision 11/09/93
2.4.2-9
DRAFT SPECIFICATION
Pu02 Powder, Cermnle Grade- Dry Process
6.3 The external and internal surfaces of the outer container and the external surface of
• the inner'container-shall be.as _freefrom contamination as possible, and in no case
shall exceed United State Department of Transportationregulation for smearablecontamination.
6.4 The Supplier shall identify each containeras follows:Outer Container:
(a) Containernumbers, innerand outer, including permitnumbers(b) Purchaseordernumbers
(c) Type of materialand lot number(d) Metal weight of material(e) Shipmentclass (Fissile Class I, II, or III)(f) Radiationunits(g) Purchaser'saddress (to be specified on purchaseorder)(h) Two packing slips mustbe attachedwhich include all shippingpapers including
two courtesy copies of AEC transferdocuments, and "Certificate of Test"analysis
InnerContainer:
(a) Containernumber(b) Type of materialand lot number(c) Gross, tare, net and metal weights
Specifi_ltionRevision 11/09/93
2.4.2-10
2.4.3 MOX Fuel Fabrication Facility Layout
The configuration of manymodernMOX fuel fabricationbuildings being constructed in Europe
and Japanare multi-story structures. As a result, an alternativeMOX fuel fabrication facility
building layout was studied which assumed a multi-story building configuration as opposed to
the original low level two-story building layout presented in the May, 1993, ABWR Pu
Disposition Phase 1A Report. The multi-story buildingarrangementis shown in Figures 2.4.3-
1 through2.4.3-3. The buildingarrangementincludes three floors above grade and one floor
below grade with overall dimensions of 220 feet by 220 feet with a height of 105 feet. This
configuration was compared to the original Phase 1A layout shown in Figures 2.4.3-4 and
2.4.3-5 which consists of two floors above grade with dimensions of approximately 490 feet by
360 feet with a height of 50 feet. These layouts were compared to determine if there was any
significant economic advantageassociated with one layout over the other. Due to the pre-
conceptual nature of the building design, and the lack of definitive site specific geologic,
meteorological, and hydrological data, it was not possible at this time to develop detailed cost
estimatesof the two alternatives. However, some general observationsbased on experience in
the design of similar nuclearstructurescan be made.
The concept of minimizing building costs by minimizing the surface area of external walls due
to nuclear tornado missile and seismic design criteria for a building designed for nuclear
material confinement was believed to be valid for early "conservative" nuclear building
designs. However, development of actual missile effect experimental data and modern seismic
dynamic analyses methods have allowed external wall thicknesses to be optimized resulting in
external walls which are in the range of from one to two feet thick which is often equal or less
than the thickness of interior walls (due to shielding requirements.) For example, a recent
design of a Category 1 plutonium confinement process building to current DOE Order 6430.1A
tornado and seismic design standards resulted in external walls which were one foot thick.
This structure was the main processing building for the Special Isotope Separation (SIS)
Plutonium Laser Isotope Separation Project which was planned for construction at the DOE
Idaho National Engineering Laboratory. The final building configuration was approximately
370 feet long by 310 feet wide with a 50 feet overall height. The external reinforced concrete
walls of this building were nominally one foot thick, which was the same thickness and
construction as most internal walls. This relatively low profile two-story building constructed
at grade was considered optimum. The process material handling steps, the interface between
this building and the laser beam tunnel, the local site soil conditions, and the relatively low
2.4.3-1
seismic acc, lerations for both building and equipment designs due to low building elevations
were the criteria which resulted in the optimized design.
The ABWR MOX fuel fabrication facility process involves transfer of powders, pellets, rods
and bundles as well as analytical samples, scrap and waste material between process/support
system steps. In the low level building layout, most of the process and support systems are
located on the first floor and material transfers are made by cart or horizontal conveyor on the
same floor level. In the ease of the multi-story layout, vertical transfers between floors at
different elevations are required. With the exception of the possible use of gravity feed in the
powder blending and milling processes, these vertical transfers will generally complicate the
material handling operations. The transfer mechanisms and enclosures must provide proper
confinement of material during normal transfers or a dropping accident, assure criticality safety
and allow remote monitoring and communication between floors during transfer. In addition,
these transfer mechanisms must assure that material accountability is maintained and that the
transfer mechanism maintenance can be accomplished within the transfer space provided.
Interfaces with services, utilities and transportation vehicles provided between the ABWR MOX
facility and other facilities on the site are generally located at grade and therefore more
compatible with the low profile building. Other factors which could affect the optimum
building configuration include land availability, local site soil, hydrologic and seismic
conditions as well as site labor rates and craft productivity during construction. Therefore, it is
believed atthis time that the' building configuration can be optimized only when more detailed
design information is developed.
2.4.3-2
_. 220' -0"
I
! ISTAmS --I
zo CHEM/MET LAB
WSR & WASTE ii TREAT_NT
i I
BUNDLE RODSTORAGE POWDER RECEIVING STORAGE
i PU02 INTERIM STORAGE
.1_ PU02 J PU02
VAULT koui/_ 3!" VAULTI ELEVi_
I-LIMIT OFI CATEGORY 1
I STRUCTURE
SHOP INST. NUC SAFETY _JLAUNDRY OFF ICE
io OFF ICES LAB
F ILTERI POWDER MEN' S/WOMEN' S GENERAL STORAGESHOP WASTE RECEIVING CHANGE ROOMS AND SHIPPINGSTDRAGE TRUCK BAY
SECUR II'YOFFICES
..
BASEMENT FIRST FLOOR - "GRADE LEVEL
Figure 2.4.3-1 Mixed Oxide Fuel Fob BuildingBosement ond First Floor Levels
GD203 FAB
D
] I EXHAUSTHEPA FILTERi R_Su
HVAC & FILER ROOMGD203/_X SCAN (POTENTIALLY CONTAMINATED)
X-RAY/NDE[NSPT. _XF_ [_D D_--'_ U
HOT.INT. !--] B B _- I
TEST _ _ _ _ _ -L|MIT .CATEG_Yr-Io oF-! STRUC_RE
.... .... ....... ...............[ W ......MATERIAL TARGET OLD MAINT
CONTR_ / "G" AREA i
_CHANICAL SERVICEEOUIP_NT ROOM
GENERAL OFFICES,_DICAL & CAFETERIA r--1
I I
2ncl FLOOR 3rd FLOOR
Figure 2.4.3-2 Mixed Oxide Fuel Fob BuildingSecond ond Third Floor Levels
, 220'-0 _
GLOVEBOXHVAC & FILTER ROOM EXHAUST
(POTENTIALLY CONTAMINATED) HEPA FILTERROOM5
GD203/MOX SCAN HOT MFGX-RAY/NDE INSPT. MAINT. TEST
OI I
POWDER RECEIVING ROD I
PU02 INTERIM STORAGE STORAGE I GRADE
Puo2
|
SECTION A-A
Figure 2.4.3-3 Mixed Oxide Fuel Fob BuildingMulti-Story Bundlng Section
2.4.3-5
I1 150M ..__1
F ]HOT BUNDLE ASSEMBLY
M,tlNT. AND INSPECTION
CHiEM/M(T LAB _ & WASTETREATMENT
MFCTEST
m
L._ 150M ..._i
F
c-n r--1
(",4
r_ D D D D D D D D D D
n J"-'l n r-'-'_. HVAC & FILTER i_)OM _/_ J
_T,_Lv _T_AT_O_ ._-ICt.ovEooXEXHAUST _1HEPA FILTE _ ROOMS cjr_ jr
V
O3 MECHANICALSERVICEEQUIPMENT ROOM
f
0 10 20; ......... ; ...... _"1
SCALE:-ltTBO
Figure 2.4.3-5 Mixed Ox;de Fuel Fob Bund;ngLow Level Bu_d;ngSecond Floor
2.4.4 FIRST-OF-A-KIND TECHNOLOGIES
An evaluation of the MOX fabricationtechnology was conducted to identify those
aspects of the process or the associatedfacilities which might be construed as first-of-a-
kind technology. MOX fuel has been fabricatedin the U.S. in significant quantities in
the past and MOX plants are now under design or construction in several foreign
countries. An evaluation of the various MOX pellet fabrication processes, includingthe mechanical blend, sol-gel and coprecal processes were presented in the Phase 1A
report. The mechanical blendprocess was selected as the reference process. Many of
the foreign plants, those in United Kingdom and Japan for example, plan to use this
mechanicalblend or dry process for fabricating pellets from mixed oxide powder. Abrief re-evaluationof the alternateprocesses was conduct¢_to confirm the choice of the
mechanical blend process as the reference. This process was chosen as the reference
because of the process simplicity, extensive experience base and the potential to result
in the lowest levels of waste. Unless the initial feed material source is expanded to
include other plutonium materials than those obtained from retirement of nuclear
weapons, there appears to be no reason to change to an aqueous process to remove orreduce impurity levels.
The mechanical blend technology for MOX fuel is well established and no major
processing changes have occurred, significant advances in the areas of automation,
material handling, equipmentdesign, and real time instrumentationcan lower worker
exposure, improve quality and increase throughput. The application of the
advancements in each of these areas, with the possible exception of instrumentation
which is considered a developmentactivity, to MOX fabricationare considered first-of-a-kind type applications.
Since the normal processing operations in the MOX factory must be automated to
satisfy reduced worker exposure limits, automationtechnology must be integrated into
the material handling activities of the factory. The majority of this automation or
robotics equipment is currently available, but integration into the current process
equipmentarrangementshas not been accomplished. Many of the automationconcepts
being utilized by foreign fuel fabricators for processing of plutonium recovered from
spent fuel are directly applicable, but each installation is site specific and the details of
operation must be developed for each application. A good example of this type of
2.4.4-1
integration is the location of mechanical assists to remove or maintain an equipmentitem or modularcomponent.
Although the technology for conversion of the plutonium metal to oxide to provide the
feed material required for the MOX fabrication process in not considered part of the
activities to be evaluatedin this phase of the effort, the facility to produce this material
is considered a first-of-a-kind installation. The required conversion technology hasbeen developed for batch type operations during activities completed in the weapons
community. These processes need to be adaptedto produce the quality and quantity of
plutonium dioxide which meets the requirementsfor fuel fabricationin a MOX facility.A series of process demonstration studies may be required to optimize the conversion
processes to satisfy the requirementsof the plutoniumdisposition activities.
2.4.4-2
2.5 WASTESTREAMCHARACTERIZATIONANDMANAGEMENT
The operation of the plutonium disposition complexwill generate a volumeof
waste each year. The various waste types to be generated tnclude spent nuclear
fuel (or high-level radioactive waste), low-level radioactive waste, transuranic
waste, hazardouswaste, and solid and sanitary wastes. Waste activities are
regulated by a variety of government agencies as listed in Table 2.5.1.
Treatment, storage and disposal of generated wastes are controlled by a variety
of instruments such as licenses, permits, certifications, consent orders, or
other written approvals.
Table 2.5.1. WasteRegulatory Agency
_ Waste:Type { j Regulating AgencY
Spent Nuclear Department of Energy and/or NuclearFuel (High-Level Regulatory CommissionRadioactiveWaste) m,,
Low-Level Departmentof Energy, NuclearRadioactive Waste Regulatory Commissionor StaLe (if......... an agreementstate)
Transuran!c Waste Department of Energy
HazardousWaste Environmental Protection Agencyorcognizant State agency (if RCRA
..... programapproved)
Hazardousand Departmentof Energy and theRadioactive Mixed Environmental Protection AgencyorWaste cognizant State agency (if RCRA
program approved) i
Soltd and Local health agencystandardsSanttary Wastes
One of the primary design goals for the plutonium disposition complex is
minimizationof wastesgeneratedin the variousprocesses.These effortsare
definitelycost effectiveas the cost and treatment/disposaluncertaintiesof
handlingthiswasteincreaseinthefuture.Thissectiondiscussesthetypesand
characteristicsof wastesgeneratedineachactivityoftheplutoniumdisposition
2.5-I
complex. For eachwaste stream, the source of the generated waste is identified
and waste treatment and minimization activities prior to disposal are also
discussed. For clartty, this section begins with defining the various wasteforms dt scussed.
WasteOeflrlltton_;
Hiah-Level Radioactive Waste (HLW)-The highly radtoactive waste material that
results fromthe reprocessingof spentnuclearfuel.
Low-LevelRadioactiveWaste (LLW)- Radioactivewastethat is not high-level
wasteor containslessthan 100nCi/gramTRU concentration.
TransuranicWaste(TRU)- Radioactivewastewitha TRU (alpha-emittingTRUwith
half-livesqreaterthan20years)concentrationgreaterthan100nCi/gramofTRU
isotopesin the wastemass.
HazardousWaste- Wastesdesignatedas hazardousbyEPAregulations(40CFR261).
Hazardousand RadioactiveMixedWaste- Wastescontainingbothradioactiveand
hazardouscomponentsas definedby theAtomicEnergyAct and RCRA.
NonhazardousSolid Waste - Non-regulatedwaste (exceptfor local landfill
requirements)
Sanitar.yWaste - Waste water normallydisposedin a site drainfieldor a
municipalsewersystem.
Spent Nuclear Fuel - Fuel that has been withdrawnfrQm a nuclearreactor
followingirradiation,has undergoneat leastone year'sdecaysincebeingused
as a sourceof energyin a powerreactor,and hasnot beenchemicallyseparated
intoitsconstituentelementsby reprocessing.Spentfuel includesthe special
nuclearmaterial,byproductmaterial,sourcematerial,and other radioactive
materialsassociatedwith fuel assemblies.
2.5-2
2.5.1 MOXFuel Manufacturing Facility
Wastes from the Mixed-Oxide (MOX) Fuel Fabrication Factllty wtll be generated tn
the manufacture of the fuel assemblies and durtng ancillary acttvtt|es. The
prtmary waste types wtll be TRU and LLW, however, small amounts of hazardous,
soltd, and santtary wasteswt11 also be generated. AsdtscussedtnSectton2.4.1
the flOX factllty simulation studtes have not yet reached the potnt where new
waste stream Information ts available. Therefore the results provtded for the
flOX factllty tn thesummary tables were taken from the Phase IA studies.
2.5.2 ABWR
Wastes from the Advanced Boiling Water Reactor (ABWR)will be generated tn the
operation and maintenance of the reactor and during ancillary activities. The
prtmary waste types will be spent fuel and LLW, however, small amounts of
hazardous, solid, and sanitary wastes will also be generated. The following
discussion provides specific information on the ABWRwaste streams. For
simplicity, the discussion wil1 be directed towards waste produced by one ABWR
unit. The total waste volume for the two unit plutonium destruction complex willbe tabulated at the end of the discussion.
2.5.2.1 High-Level Radioactive Waste
The operation of the ABWRwill produce spent nuclear fuel (SNF). Although not
generally considered the same, this report will use the terms high-level
radioactive waste (HLW) and spent nuclear fuel interchangeably. Using strict
definitions, the ABWRproduces only SNF; no HLWis produced as SNFwi11 not be
chemically processed prior to disposal. The disposition of this waste is the
responsibility of the DOE. The amount of SNF waste generated annually is
projected to be 162 bundles (29 MTHM)per reactor, based on a 75% capacity
factor and a 523 day cycle. The ABWRMOXfuel would be irradiated to a burnup
of 38,000 MWd/t which is typical of the SNF generated by light water reactorslicensed in the United States.
2.5-3
Treatment;
Onstte- Spent fuel wtll be stored in the reactor spent fuel pool for a minimum
cool-downperiod prior to further storage on site or shipmentfor offstte storageor disposal. There wtll be no other treatment of this waste at the reactor site,
except for the removal of target rods in the tritium production mode.
'Offstte - No offstte treatment of the wastets_projected at thistime except for
the placementof spent fuel bundles in a repository-required container prior toplacement in the repository (refer to section 4.4.4 for additional information
on repository requirements).
Disposal
.... : Disposal of SNFis the responsibility of the DOEper the Nuclear WastePolicy Act
of 1982, as amended. Current plans are to ship SNFto a national repository for
long term deep geologic storage/disposal.
2.5.2.2 Low-LevelRadioactive Waste
The operation and maintenanceof the ABWRwill produce a quantity of LLWeach
year. The total amountof LLWgenerated annually is projected to be 165 m3. The
disposition of this waste is the responsibility of the DOE. LLWwill either be
treated within the complex and disposed on the same DOE site or will be
transported to another DOEsite for disposal. Dependingupon site specific
acceptance requirements, the waste maybe grouted, compacted,vitrified, and/or
otherwise treated prior to final disposal. LLWwill exist in three physicalforms (solid, liquid, gas) as discussed below.
SolidRadioactiveWaste
Solid radioactivewastesare generatedprimarilyin thevariousreactorfluid
cleanupsystems. Table2.5.2.2,-Iliststhe major contributorsto the waste
2.5-4
volume from "wet" waste processes. After processing, the resulting waste forms
are a solld andwill meet dtsposal crtterta with respect to free l|qutds.
Table 2.5.2.2-1. Wet WasteGeneration - Pretreatment
. m . , _ . _m .... m m r_m]mm m _ _ m
i_::_iiiii_::.,_Vol_;Generated. _iii__SPecitic i-,__.iii_
I
CUWF/D Sludge.... 4.7 7.3 E7
FPCF/C Sludge 1.8 ....... 1.94 E6
CondensateFilter S!udge 4.6 2.40 E5iii i i,rl i
Leg Ftlter Sludge 0.2 1.5 E6i i i J ,1
CondensateDemtnerallzer 18.0 6.7 E4Resin
i ii i
LCNDemineralizer resin 5.0 1.18 E5ii i i
HCWDemtneraltzer resin 2.7 8.4 EO
Concentrated Liquid Waste 27.4 4.67 E3
LTotal 64.4 ......,I
Table 2.5.2.2-2 provides a projection of the volumes(pre-treated) andactivitiesof dry LLg generated by the ABWR.Table 2.5.2.2-3 is a summaryof the treated
waste volumesandactivities that are projected to be shipped for disposal aftertreatment.
Liouid Radioactive Waste
The radioactive waste treatment systemswill generate about 27.4 m3 per year of
concentrated liquid waste. This waste will be solidified in preparation foroffsite shipment and disposal. Therefore, no liquid radioactive waste would
require disposal.
GaseousRadl,oactive Wal_te_s
2.5-5
Theprimary sourceof radioactive gasests generation during the flsston process.
A small amountof noble gases is released tnto the coolant andthen collected by
the offgas treatment system. This systemis basically a holdup that delays the
release of nobel gases a11owtng for a period of radioactive decay prior to
atmospheric release. The flow capacity of the system is 40 m3 per hour. Holdup
times are a minimumof 30 days for Xenonand 40 hours for Krypton. Due to the
relatively short half-life of mostnoble gases, no gaseouswaste is collected fordisposal.
Table 2.5.2.2-2. "Dry" Soltd Haste - Pretreatment
Dry Waste Source VolumeGenerated Totalm3/Year Curiesi i i
i I iiiimmli
Combustible 225 1.6Waste
Compactible 38 0.3Wastei1,11
Other Waste 100 7.0
Total 363 8.9
Table 2.5.2o2-3. Annual ShippedWasteii iiii ' i ii i
Haste Type ShippedWaste Total CuriesVolumem3/Year
i I
Concentrated Waste 4.4 1.3ii ,
Combustible Waste 5.6 1.6iii i
CompressibleWaste ]5 0.3
Resins and Sludges 40 670Other Waste 100 7
i
Total 165 680.2
The reactorfacilityventilationsystemalsocollectscontaminatedparticulates
inthe air. The systemcollectsairbornecontaminantsin variousareasat a rate
dependentuponthe potentialfor and severityof contamination.The collected
2.5-6
airis thenpassedthrougha highefficiencyparticulateair (HEPA)filterwhere
it is cleanedof 99.95%of particulates.Aftera limitedlife,the contaminated
HEPAsare replacedand disposedas solidradioactivewaste.
2.5.2.3 HazardousWaste
' The operation and maintenance of the ABWRwtll generate a small volumeof waste
considered to be hazardous 'by EPA regulations (40 CFR 261). Examples of
hazardouswastes are paints, solvents, lubricants, film developmentsolutions,
laboratory chemicals, andother chemicals that cannot be disposed as solid non-
hazardouswaste. This waste is expected to be transported offstte for treatment
anddisposal by a contractor. This waste is expected to total about 290 m3/year.
2.5.2.4 Non-HazardousSolid Waste
The presenceof the ABWRwork force will result in the generation of a volumeof
non-hazardouswaste. This waste consists of garbage, office trash, packing
material, and other refuse. A disposal contractor will haul the 3,100 m3 per
year of waste to a local landfill for disposal.
2.5.2.5 Sanitary Waste
Sanitary waste is that liquid stream normally disposed in a site drainfteld or
a municipal sewer system. The presence of the ABWRwork force will result in
about 1.6 E6 gallons per year that need disposal.
2.5.3 Tritium Production
Wastesfrom the manufactureof targetrods and from the extractionof tritium
will begeneratedin relativelysmallquantities.No radioactivewasteswillbe
generatedin the manufactureof targetrods. The primarywastetypegenerated
duringextractionis LLW,however,verysmallincrementalamountsof hazardous,
solid,and sanitarywasteswill alsobe generatedduringbothactivities.The
followingdiscussionprovidesspecificinformationon thesewastestreams.
2.5-7
2.5.3.1 Low-Level Radioactive Waste
Tritiumproductionactivitieswouldadd a verysmallincrementalamountof low-
levelradioactivewasteto the complextotal. Thiswastewouldbe generatedat
the reactorsite primarilyduringthe removalof irradiatedtargetrods from
spent fuel bundles. This waste has basically the same_ _ :_,_':_::cs as other
LLWgenerated in the reactor refueling area and the ;_ _ ,._ _rage_pool
area. This waste would consistprimarilyof used protec _ .... r,,,ig andother
disposable items that have comein contact with pool water.
A smallamountof LLW would be generatedfrom the spent targetrods after
tritiumextraction.The spenttargetrodsare expectedto containabout50 Ci
each of residualtritiumand shouldalsocontaina smallquantityof activation
' " productswithintherodmetals. Mostof theactivationproductshaveshorthalf-
lives,but traceamountsof Co-60willcausean externalradiationhazardfor a
relativelyIongperiodof time. Spenttargetrodsshouldbe classifiedas LLW;
probablyas ClassB (10CFR 61.55). However,they couldalsobe consideredas
spent nuclearfuel by one NRC definition(10 CFR 72.3)whicl;would require
geologicrepositorydisposal.At thistime,the mostreasonableandeconomical
disposaloptionwouldbe as I'"
_Thesetwo activitiesare pr_v _edto generateapproximatelyI0 m3/yearof
additionalLLWvolume.Thisvolumeof LLWisapplicableonlyiftheABWR isused
in the tritiumproductionmode and wouldnot be generatedif tritiumis not
produced. The low-levelradioactivewastewill be disposedat a LLW disposal
site. This sitemay be at the samelocationas the complexor the wastecould
be transportedto the nearestDOE LLW disposalsite. Dependingupon site
specific acceptancerequirements,thewastemaybegrouted,compacted,vitrified,
and/orotherwisetreatedpriorto finaldisposal.
2.5.3.2 HazardousWaste
Tritiumproductionactivitieswouldadd an insignificantincrementalamountof
hazardouswasteeachyear.
2.5-8
2.5.3.3 Non-HazardousSolid Waste
Tritium production activities would add an tnsigntficanL incremental amountof
non-hazardouswaste each year.
2.5.3.4 Sanitary Waste
°
Trttium prOduction activities would addan insignificant incremental amountof
sanitary waste each year.
2.5.4 SuM.ary
This section described the characteristics, amounts, treatment, and final
" disposition of the wastes generated in the operation of the plutonium disposition
complex.This information is tabulated in Table 2.5.4-1 and is valid for a HOX
fuel manufacturing facility, two ABWRs,andancillary facilities in the complex.
The information provided indicates that the wastes generated are no greater in
volumethan for other similar projects.
Table 2.5.4-1. ComplexWasteStream Summarym,
i Waste........ .... Principle Quantity Treatment DispositionStream Source Per Year
- iI
MOXFacility ......
Solid Compaction DOE LLWRadioactive 60m3 disposalsiteWaste
i
Low-Level Liquid SolidificationN/AWaste Radioactive - 0
Wastei.,
Gaseous N/A HEPA Atmosphericeffl uent f i l trat i on re l ease
i i i iiii,
Trans- Fuel Included in WIPP Transport touranic Manufacturing LLWtotal certification WIPPWaste Process
2.5-9
;i_ ii!iPrinctple ii i _ Quanttty Treatment _D|sposttton;ii_iiiiiiiiiiStream_iiii_i__:ii!iiii_!Source /i;i !ii__I__!;Per Year i_;_!i! _i ;_
Hazardous Fuel No onsite ContractorWaste Manufacturing lOOm3 treatment disposal per
- Process . RCRA
Solid Non-hazardous 30m3 No onsite Disposal inWaste waste sources treatment local landfill
i i i
Sanitary Sanitary No onstte Disposal perWaste sources 2.1 E6 9al treatment local codes
I I
High-Level Spent Fuel 324 Decay >1 year HLWrepositoryWaste Bundles
58 MTHM
Solid 330 m3 Evaporation DOELLW• Radioactive 1360 Ci Compaction disposal site
Waste
Low-Level Liquid None Recycled toWaste Radioactive discharged Demineral izer condensate
Waste storacje
Gaseous N/A Offgas holdup Atmosphericeffluent HEPA rel ease
filtration
Hazardous Maintenance No onsite ContractorWaste activities 580 m3 treatment disposal per
RCRA
Solid Non-hazardous No onsite Disposal inWaste Waste-Sources 6,200 m3 treatment local landfill
Sanitary Sanitary 3.2 E6 Gal. No onsite Disposal perWaste Sources treatment local codes
m
TrtttumExtraction ..
Low-Level Spent target 10 m3 Compaction DOE LLWWaste rods disposal site
Solid Non-hazardous ~0 No onsite Disposal inWaste waste sources treatment local landfill
, i
Sanitary Sanitary ~0 No onsite Disposal perWaste sources treatment loc._lcodes
i "'"
2.5-]0
3.0 TRITIUM PRODUCTION
In this section, the option to convert the reactor system to produce tritium has been evaluated.
Work during Phase 1C has led to a MOX fueled core design that meets the tritium contract
quantity requirements. This Phase iC core design has a core average Pu enrichment of 5.9%
with an average burnup of 28,000 MWD/MT. All nuclear and thermo-mechanical design
criteriafor normal operation have been met. There are four tritium target rods per assembly
(similar to-1A). The currently designed core assumes that target rods are irradiated for one
cycle and then processed.
The ABWR can produce the contract quantity of tritium using partial core reloads and longer
target exposure time. However, the target rod performance data is currently limited to about one
year exposure. While we believe the target rod is capable of much higher exposures in the
ABWR, the reference design is based on discharging all target rods after one cycle to provide a
conservative basis for initial operation.
Additional evaluations of the worker and public health impacts of tritium production in the
ABWR confirmed that these impacts are minimal and that no plant modifications would be
required. The average dose to a plant worker is 1/100 of the DOE limit and the maximally
exposed offsite person is 1/1000 of the EPA limit.
The key difference between the two designs is the peak clad temperature during the design basis
loss of coolant accident (LOCA). This was the limiting factor in the NPR reference design
whereas the ABWR target rod isn't subject to any significant temperature transients.
The equilibrium core design analysis for tritium production is given in Section 3.1, followed by
an examination of the target rod design and performance in Section 3.2 The support facility
requirements are given in Sections 3.3. The impact of tritium production on ABWR plant
operations is examined in Section 3.4.
3.1 MOX CORE DESIGN FOR TRITIUM PRODUCTION
Phase 1A discussed a urania ABWR fuel bundle and core design for tritium production. This
section reports a MOX tritium production core capable of producing 43 million curies of tritium
per year. The tritium is produced inside specially designed rods containing a lithium-6 target.
3.1-1
Four of these target rods replace fuel rods in each fuel bundle in the core.
The basic concept calls for an in-reactor target residence time of about one year, after which
time the target rods are removed from the fuel bundles and sent to a tritium extraction and
processing facility. The important features of the core design for the reference tritium produc-
tion case are summarizedin the following sections.
..... : The reference tritium production core and fuel design can produce 43 million curies of tritium
per year in a single ABWR power plant. A summary of the important fuel cycle parameters and
tritium production rates is shown in Table 3.1-1.
Table 3.1.1. Parameter Summary, Tritium Case
Numberof Reactors 1Cycle Length, EFPD 320.9
Discharge Exposure, MWd/MT 28164Reload Batch Size 280
Average 239pu Enrichment 5.92%Plant Capacity Factor 87%Tritium Output, curies/yr 43xl 06
3.1.1 Reference Bundle Design
The bundle design for the tritium production option contains four 6Li target rods and 56 PuO2
power producing rods. This bundle design resembles the reactivity characteristics normally
associated with enriched uranium fuel. However the neutron absorption of the target rods nearly
doubles the enrichment required to obtain the fuel discharge exposure.
The bundle axial and radi',d enrichment distribution is given in Figure 3.1-1 along with the
gadolinia distribution. Values of enrichment for plutonium are read in hundreds of a percent.
For instance, the "310" refers to 3.10 w/o PU _'39. For the tritium production case, natural uranium
blankets are not used. This allows a maximum bundle average enrichment to overcome the 6Li
target rod reactivity penalty and also flattens the axial power shape so as to reduce the exposure
peaking in fl_etarget rods. There are a total of five pellet types used in this design, one of which
is a gadolinia bearing pellet.
3.1-2
Figure 3.1-1 Bundle Design for Tritium Productionip..-
iDee e o® @_ 0 0 _(b (D® @@@@@I_ (b
f3_
\2-"
©®_ @@@(- ""II 11
_ j
(9@@ @I
61..i
Urania • 071 071 071 Targel 071
Plutonia > 310 350 400 900
Gadolinia > 05o
Figure 3.1-2 Beginning of Life Assembly Power Distributionfor Tritium Production Assembly
.249 1.212 1.223 0.099i
I E938 0.96; 0.963 1.208
I ._82 1 3.964 1.193,,=. II.=_lm_=
b =23.934 1.227
I 2 3.934 1.227
I . 1 3.963 1.193.,,
' ,3 3.!_58 1.205
0.09911o20811,,193 93 1 ._05 0.099
..... [ [
3.1-4
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The infinite lattice radial power peaking is essential in determining the peak power producing
rod. The distribution of relative power peaking (normalized to unity across the lattice ) is shown
at beginning of life for the 40 percent void case in Figure 3.1-2. The exposure-dependent k.,
are given in Figure 3.1-3 for the uncontrolled lattice. This figure shows the k.. for three void
histories.
3.1.2 Equilibrium Core Design
The equilibrium core design philosophy for the tritium production option was to simulate the
reactivity distribution of a annual refueling equilibrium UO2 core in order to provide simple
operation. A single fuel nuclear design was utilized in an equilibrium batch of 280 bundles.
The detailed core design layout is presented in Figure 3.1-4. The numbers shown in the
beginning-of-equilibrium-cycle core map represent the relative number of cycles since fuel
loading. For instance, the number "1" refers to fresh fuel (loaded this cycle) and the number "4"
refers to bundles which are about to start their fourth cycle.
A single nuclear design of fuel is loaded into the equilibrium cycle. The important fuel bundle
parameters were summarized previously. A control cell core loading strategy that contains 37
control ceils was utilized. Due to the improved hot to cold reactivity swing characteristics of the
ABWR core, it was possible to design the fuel with a sufficient cold shutdown margin and still
maintain sufficient hot excess reactivity. The necessary high fissile content dictated a power
derate for the latter part of the cycle. Figure 3.1-5 shows the power profile through the cycle.
The important parameters of the equilibrium cycle design are summarized in Table 3.1-2.
Examination of the results reveals that all thermal and reactivity requirements are satisfied.
3.1.3 Core Thermal Margins
The critical power ratio and MAPLHGR thermal margin performance.are plotted as a function
of cycle exposure in Figure 3.1-6 and 3.1-7. Operation within the MAPLHGR limit assures the
mechanical integrity of the fuel rods is maintained by limiting their power output in an
appropriate manner throughout their lifetime. The MAPLHGR limits imposed on this cycle with
a relatively low discharge exposure are the same as the fuel licensed for up to 38 GWD/MT.
These results demonstrate ample mmgin to core thermal limits.
3.1-6
Table 3.1-2. Equilibrium Cycle Key Parameter SummaryTritium Case
Cycle Length, EFPD .... 320.9Cycle Energy, GWd 1260_
Cycle Exposure, MWd/MT 7818Core Mass, MT 146.4Reload Enrichment, w/o Pu-239 5.92Reload Batch Size 280Maximum MAPRAT 1.00Maximum CPRRAT 0.82
Maximum LHGR, KW/ft (LHGR limit = 14.4) 13.3MCPR (OLMCRP=l.25) 1.53Minimum Cold Shutdown Margin 1.60Hot Excess Reactivity at BOC 1.00
3.1.4 Reactivity Limit Summary
The reactivity performance of the tritium production option design is summarized in Figure 3.1-
8 and Figure 3.1-9 Due to the improved hot to cold reactivity swing of the ABWR N-lattice,
there is sufficient cold shutdown margin; therefore, there is little or no impact of the core design
from cold shutdown margin considerations.
Figure 3.1-4 Equilibrium Cycle Loading Pattern
{"-'-'2 - Second Cycle
CycleLmded 3 - ThirdCycle4 = Fomh Cycle
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3.1-8
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Figure 3.1-7 Equilibrium Cycle Maximum CPRRAT vs Exposure
EquilibriumCycleMaximumCPRRATvs. Exposure
1.1 I _ .............. ; - _--- ! .........
- _ i = =-1.05 - ............................_......................................................;........................ _..........................i..........................T...................... i.........................: = .......
i i i ii i iiiii -- iiiii ] ..../ i sm
/ : | iI i : ! i
0.95 1-........................................................_.........................._...............................'..........................._.............................._...........................*.............................i j i i i
/ t . ! i ! * !/
0.9 4-............................._................................_..............................._................................_.............................i..............................._............................;................................./ ; i I , i i/ i l _, I ; i i
t" I t _ ! ! t t,. 0.85 ..........................._................................'............................i..........................................................i.............................'..........................'.........................E: / _ _ ! -- I I i !
• __ _ =
i/ i _ ; ! i ' i
o.751..........................!....................................................................... ............................0.7 .L ........................._.........................._....................._.....___ ......
! i * l * i ;
i i ; i i i___u._ ..........................*.............................-:............................_...............................i...........................+............................._.............................._.............................
o.6.. 1 _ ,. i ..........1, .... ._ i . _--_ ......0 1 2 3 4 5 6 7 8
Cycle Exposure {GWDIst)
3.1-11
Figure 3.1-7 Equilibrium Cycle Maximum CPRRAT vs Exposure
EquilibriumCycle Maximum CPRRATvs. Exposure
i i : i
I * i i i i ;
' [ i' | ill i ' ii i i t _ ii _ ii ....i ! : : i " i I
I ; i i i i i ;
v -- ..t_J. _(_J_ ........................... 4 ................................ .+ ............................ i................................ i............................. i............................... . .............................. . ...............................
0.9 ................................_................................._.................................._.................................i................................i................................4...............................i.....................................; ; } t i 1 i
o.85 I .............................._.............................._'............................._.................................'..............................}..............................._'..............................."...................................0.8 i , : - III _ i i 1 i
t ........................... ,_...............................t ..............................i........................ • ..........................i................................4-.................................+..................................
0.75
0.7 ...........................i................................_,..................................i...........
°°'t...............................i..............................................................i.................................i..............................i...............................,i................................i................................0.6
0 1 2 3 4 5 6 7 8
Cycle Exposure (GWD/stl
3.1-11
Figure 3.1-8 Equilibrium Cycle Hot Excess Reactivity
Equilibrium Cycle Hot Excess Reactivity
1.25 "
t,,-,,,. i
m 1 _'"
_._(D
"o
0.75 ........................................................................................................!............................................................................................
{ i ;
t,)n- 0.5
u>¢
L__" 0.25
0 ..............
0 1 2 3 4 5 6 7 8
Cycle Exposure (GWD/st)
3.1-12
Figure 3.1-9 Equilibrium Cycle Minimum Cold Shutdown Margin
EquilibriumCycle Minimum Cold Shutdown Margin
3 i i _
.c== 2m:zC . : : i I
1.5 ....................................................................................................i.........................._.............................................................................................................................................0 : i ;
"¢: i , ,1 ' ' .....ira.
o(J
0.5 .................................................................................................................................._,............................._......................................................................................................
0 ,_ l t
0 1 2 3 4 5 6 7 8
Cycle Exposure{GWD/st)
3.1-13
3.1.5 Core Performance Description
The core performance characteristics as a function of exposure through the cycle are given in
Figure 3.1-10 through Figure 3.1-13. The core maps in these figures show the control blade
patterns in the core expressed in terms of notches (which are three-inch sections of blade)
withdrawn from the top of the core. Those cells which have no numbers represent cells in which
there are no blades inserted. The thermal limits and reactivity mat'gins associated with the given
exposure are noted in the summary included with each figure. As seen from these figures, all
thermal and reactivity margins are met. The resulting core average power and exposure profile
are also given. Since the reactor core design itself provides sufficient margins, it is not
necessary to axial grade the fuel assembly to flatten or accommodate the shifts in power.
However it may be advantageous to flatten the axial power shape in order to reduce the exposure
peaking of the target rods. Table 3.1-3 shows target rod design parameters and performanceresults.
3.1-14
Table 3.1-3. Tritium Core Design Options
Target rod OD (in) 0.483Target rod clad thickness (in) 0.030
Target pellet OD 0.390Target pellet ID 0.2406Li enrichment 50%
Tritium Output 43.2 MCi/yr
6Li n-oc rates (per second per cm rod length)
BOC Average 2.7xl 0t3BOC Peak 5.4x 1013
MOC Average 2.7x10 _3MOC Peak 5.1x10 _3
EOC Average 2.7x 10_3EOC Peak 5.1x1013
3H Inventory (Ci per rod)BOC Average 0BOC Peak 0
MOC Average 6547MOC Peak 7535
EOC Average 12373EOC Peak 14070
End of Life Average 12373
Target Pellet Exposure (GVR)End of Life Average 147End of Life Peak 166
3.1-15
Figure 3.1-9 Equilibrium Cycle Minimum Cold Shutdown Margin
I =1
_ 36i 28 34 28 36 CycleExposure,MWd/st 0
36 24 32i 20 32 ..... 24 36 Cycle Energy, MWD 0Numberof Full PowerDays 0
- - .... CoreAverageVoidFraction .4572
28 i32 20 !30 ......20 _32= 28 "--1 Core Flow_ Mlb/hr I. 151ChannelPealdng 1.2654Maximum34 20 30 14 30 20 134 D CoreAxialPowerPeak ].30]
._d MaximumRAPLHGR .95628 32 20 30 20 32 28 M&gimum CPRRAT .819
• Hot ExcessReactivity, % 1.0536! 24 32J 20 32 24 36 Coid Shutdown Margin 2,84 ,
36 28; 34 28' 36
N N - Numberof 3 inch increments thatthe control bladeis withdrawnfrom fully inserted
26 28
2422 2220 2018 18
-_ 12 -_,.12 ................................................................
10 i _ 108 _ 8
6 i _ 6 i4 .4.,
4 _ i2 i _ 2 :
0 _ 0
0 G4 Q8 12 1.6 2 0 4 8 12 16 23 24 28
CoreAxialAverageRelative CoreAxialAverageExposure,Polar G1/_/st
3.1-16
Figure 3.1-11 Equilibrium Cycle Core Data Summary at 4000 MWD/st
" Cycle Exposure, MWd/st 4000
36 28 36 ,28 36 CycleEnergy,MWD 644660 ,NumberofFullPowerDays 164
, CoreAverage Void Fraction .47436 22 34 24 34 22 36! Core Flow, Mlb/hr 1.151 ..
,,MaximumChannel Peaking ,1129828 34 r26 34 26 34' 28 Core Axial Power Peak 1.40 -
"3 Maximum'RAPLHGR .982
36 24 34 26 34 24 36 . I MaximumCPRKAT .797_..,'i Hot Excess Reactivity, % 0.63 ....
28 34 26 34 26 34 28 Cold ShutdownMargin 1.6 _
36 22 34 24 34 22 36i
36 28 34 28 36
N - Numberof3 inchincrementsthatthecontrolbladeiswithdrawnfromfull)'inserted
2626 i _ i " ' -
24 .......................i................
18 18
10 _ 16
14 7=-_ 12 _..................+.................. -_ 12
1o8 8
6 ....................... 64 4
2 20 0 ! i ! -! ! i,,,
0 Q4 08 12. 1.6 2 0 4 8 12 16 29 24 28
CoreAxialAverageRelalive CoreAxialAverageExposure,Po_er G'V_/st
3.1-17
Figure 3.1-12 Equilibrium Cycle Core Data Summary at 6000 MWD/st
34 36 34 - MWd/st 6000..... Cycle Energy, MWD 966990
- Number of FuUPower Days 24636 34 36: 34 36 "Core Average Void Fraction .480Core Flow, Mlb/hr ' 1.151 .......
34 34 36 32 361 34 34 Maximum Channel Peaking 1.271Core Axial Power Peak 1.40
36 36 j32 32 36 36 _Maximum RAPLHGR .938 -,,
Maximum CPRRAT .777
34 34 36 34 36 34 34 -Hot ExcessReactivity, % 0.10
Cold ShutdownMargin ].6 ....36 34 36 34 136
34 34 34
N - Numberof 3 inch incrementsthatthe controlbladeis withdraw_fromfullyinserted
26 28
24 ................................................................,..................... i
i i _ i : i! ! _ ! ; i
16 _ 16
z_12 ....................i......................_....................................4................... ,.= 12
J
Idlam ! i =
8gJ ................... _................... _ ................... i................ ._...................
.................. i ..................... ! .................. i................. _ ..................
" ............... i..............i...............2 0 4 i 1 i 1 1
0 i ..... 4 l l 0 4 8 12 16 20 24 28
0 Q4 03 1,?. 1.6 2 CoreA,_ialAverageExposure,Core _al Average Rela_ve Po_r G_O/st
3.1-18
Figure 3.1-13 Equilibrium Cycle Core Data Summary at End of Cycle
-- ,,
Cycle Exposure, MWd/st 7818CycleEnergy,MWD" 1259960 "Number of Full Power Days 321CoreAverage Void Fraction .479
f m , ,,,,
CoreFlow, Mlb/hr 1.151Maximum Channel Peaking 1.242
J ..... Core AxialPower Peak 1.61MaximumRAPLHGR .918
.... MaximumCPRRAT .687,,
Hot Excess Reactivity,% 0.0Cold ShutdownMargin 116 _
N- Numberof3 inchincrementsthatthecontrolbladeis withdrawnfromfullyinserted
28 28
24 ..........................!............................._................!.............i...............22 2220 2O18 ..................._...............................,.................._................... 18
-.¢12 '_ 12
10 _ 108 8
4 4
2 2
0 Q4 (38 12 1.6 2 0 4 8 12 16 20 24 28
COreA_ialAverageRelalive CoreAxialAverageExposure,Pover GWD/st
3.1-19
3.2 LITHIUN TARGETRODDESIGNANDPERFORMANCE
This section describes basic design details of the target rod and reports the
results of perf_ormance analyses for one-cycle exposure. The performance
, assessment includes: thermal analysis, pressure/stress analysis, normal
permeation leakage, failed rod leakage, and tritium distribution in and leakage
from the reactor coolant system. A computer program TEP2was used to perform theanalyses. This code combines all the calculational functions and introduces a
new approach to estimating failed rod leakage (see 3.2.7.1). Details of this
computer code are given in Appendix C, "T2P2: A computer Program for
EstimatingTritium Target Performanceand Tritium EnvironmentalSource Terms".
3.2.1 Design Approach
Reviewof the light water reactorlithiumtarget developedby the TritiumTarget
Development Projecti (TTDP) and its performanceduring normal and off-normal
operatingconditions indicatedthat it is readilyadaptablefor applicationin
the ABWR. Thus, the target rod design propesed is identicalto the reference
design developed by the TTDP except for minor dimensionalchanges required to
interfacewith the standardABWR fuel bundledesign and eliminationof the outer
permeationbarrier.
As a result of the extensive informationavailablefrom the TTDP the approach
used to establishthe ABWR tritium productioncapabilitywas:
• Scale the referenceTTDP targetrod design to fit the standardABWR
fuel bundle consistentwith the design criteria and requirements
establishedby the TTDP2.
• Select a Lie enrichment and target rod placement appropriate to
achievegoal tritiumproductionwith minimum impacton the ABWR core
design.
3.2-1
• For the selected core design determine the Li e n,a rate, the target
rod heat generation rate, etc. necessary to assess target rod
performance and the .,_nual tritium production capability.
• Confirm that contract quantitiesof tritium are produced and that
the targetdesignselectedmeets the performancerequirementsduring
normal operations and transient conditions based on test and
analysisresults from the TTDP.
3.2.2 Reference Design
The target rod, as illustrated in Figures 3.2-I and 3.2-2, consists of
cylindrical,annular LiAlO2 pellets surrounded by a nickel-platedZircaloy-4
getter to absorb and retain tritiumduring irradiation,thus maintaininga low
tritium partial pressure in the free gas space. An inner zirconium liner,
located in the central hole of the annular pellets, assists in absorption of
tritium in the getter by chemicallycracking3H20released from the pellets to
tritiumgas. For conveniencein assembly,the pellets,getters,and liners are
packaged into 12.5-inchlong units called "pencils". There are twelve pencils
in the 163-inch long ABWR target rod.
The pencils are contained in a Type 316 stainless steel cladding tube. An
aluminidecoating is appliedto the inner surfaceof the claddingto provide a
barrier to permeationof tritium into the reactorcoolant.
Table 3.2-I summarizes the key target rod design parameters for the ABWR pre-
conceptualdesign and the TTDP referencedesign.
As indicatedin Table 3.2-I the ABWR targetrod is larger in diameter and has an
aluminizedbarriercoatingon the insidesurfaceonly. The ratio of diameterto
wall thickness is essentiallythe same in both designs. The ABWR rod has a
smaller void volume ratio which results in a higher gas pressure for the same
GVR. However, the clad hoop stress at EOL is stillwell below the unirradiated
yield strength.
3.2-2
Figure 3.2-1 ABWR Tritium Target Rod Cutaway
GETTER(Nickel-Plated
UPPER GETTER Zlrcaloy(Nickel-Plated ALUMINIZED
CLADDINGTOP END Zlr©aloy) TABS BOTTOM
PLUG ENOPLUG
|4_
SPRINGGETTER DISK INNER LINER(Nickel-Plated (Zirconium)
GETTER DISKZircaloy) CERAMIC TARGET (Nickel-Plaled
PELLETS (LIAIO2) Zlrcaloy)IN PENCIL
Figure 3.2-2 ABWR Tritium Target Rod
Table 3.2-1. Target Rod Design Parametersii
_ PARAMETER .... ' ABWR_ NPRREFERENCE_!
Outside Dia. (in.) ..... 0.483 0.371
Barrier Coatin_l I.D. Onlj/ I.D. & O.D.
O.D./(2 x Wall) 8.1 8.2
Rod Void Vol./LiAI02 Vol. 0.8 1.3
Average GVRat EOL 79* 83
Average LHGR(kW/ft.) 0.9 0.4
Average Clad Temp. (°F) 544 610i i llll i
Peak LOCAClad Temp. (°F) <600 1,700
* - In-reactor test data to GVR= 116
The key differencebetweenthe two designs is the peak clad temperatureduring
the design basis loss of coolantaccident (LOCA). This was the limitingfactor
in the NPR referencedesign whereas the ABWR target rod isn't subject to any
significanttemperaturetransients.
3.2.3 Target Rod Performance-Normal Operation
Table 3.2-2 summarizesthe key operating parametersfor the average and peak
target rods in the ABWR core during normal operation. The values in this table
were generated by the T2P2 computer program.
The TTDP in-reactortests operated at clad temperatures-50 °F higher than the
ABWR target rods. However,the n-a reactionrates in the ABWR rods are higher
than in the TTDP test rods. The net result is the ABWR rods have lower clad and
gettertemperaturesthanthe in-reactortestsand comparablepellettemperatures.
The end-of-life(EOL)gas-volumeratio (GVR)for the ABWR rods is well withinthe
range of the in-reactortest data.
3.2-5
The EOLhelium pressure and clad stress in the average ABWRrod are comparable
to the WC-1 test (1,900 vs 2,300 psi and 13.6 vs 9.3 kst). The clad stress in
the peak ABWRrod is higher but still well below the unirradiated yield strength
for 316 stainless steel. Further, the yield strength of the cladding will
increase substantially during irradiation providing additional margin.
Table 3.2-2. Target Rod Operating Parametersi i ,, i
PARAMETER AVERAGE ROD PEAK RODi i i
Clad O.D. Temp. (°F) 542 543i
Getter Temp. (°F) 579 588,,
Pellet I.D. Temp. (°F) 679 712,,,,
EOL AverageGVR . 79 101
EOL Helium Pressure(psi) 2,300. 3,000
EOL Clad Hoop Stress (ksi) . 9.3 14.0
EOL (H+T)/Zr in Getter . 0.3 0.38
Cum. Tritium Perm. (Ci) 0.07 0.09
EOL Tritium Inventory(Ci) 11,400 15,600, ,,
Finally,the getter loading at EOL, adjustedfor hydrogen ingressbased on in-
reactortest experience,is well below the design specificationlimit of 0.7 and
comparableto that observed in the WC-I test.
As indicatedin Section3.2.1 the ABWR targetrod was designed to remainwithin
the criteriaestablishedby the TTDP. The above summaryconfirmsthat this goal
was achievedand furtherthat the normaloperatingconditionsfor the ABWR target
rods are within the experiencebase establishedby the TTDP in-reactortests.
3.2.5 Off-Normaland AccidentConditions
The TTDP conductedextensiveanalysis and testingto determinethe responseof
the getter-barriertarget design to light water reactortransientand accident
conditions. While the TTDP analyseswere based on a 1,250MWe pressurizedwater
3.2-6
reactor(PWR),the followingfundamentalconclusionsrelatedto target response
also apply to the ABWR:
• Because of its low heat generation rate the target rod does not
experience significant temperature increases during over power
transients. For example,in PWR rod ejection accidents3 the target
internaltemperaturewas estimatedto increaseonly 30 °F.
• During a loss of coolantaccident(LOCA)where the core is uncovered
the target will be heated by thermalradiationfrom the surrounding
hot fuel rods.
• The getter-barriertarget can readily be designed to withstand a
LOCA, includingcore uncovery,without clad breach or bulgingto an
extent that would effect coolability.
• All other off-normal,transientand accidentconditionsaddressedin
a typical PWR SAR were much less limiting in terms of target
integritythan the design basis LOCA.
Based on the resultsof the TTDP the only transientor accidentconditionsthat
posed a concernto the integrityof the targetcladdingwere those which involved
core uncovery. Howeveras long as the core remainedcoveredthe target rods did
not overheatbecauseof their low heat generationrate. Since theABWRcoredoes
not uncover during the largest LOCA, there is no concern for target integrity
during off-normaland accidentconditions.
3.2.6 Multi-CycleOperation
As demonstrated in Section 3.1 the ABWR can produce the contract quantity of
tritium by dischargingall the target elementseach year. Since the available
in-reactorperformancedata is limitedto _ I yr exposure,the full core annual
discharge was selected as the reference case. This provides a highly
3.2-7
conservativedesign approach that could be implementedwith the first core if
desired.
The one-third replacement,three year exposure mode more fully utilizes the
burnup capabilitiesof the ABWR fuel and would be a much more economicaltritium
production cycle. A preliminaryassessmentof the potentialto achieve three
cycle exposurelevels in the target rod indicatesthat it should be feasible.
Further,generatingthe in-reactorperformancedata necessaryto supportmulti-
cycle target operationcould readilybeaccomplished using lead test assemblies
in the ABWR after startup.
3.2.7 Impactof Tritium Productionon Plant Operationsand Effluents
The effectsof lithiumtargets on the nuclearcharacteristicsof the ABWR core
were discussedin Section3.1. This sectionaddressesthe loss of tritiumfrom
target rods to the coolantsystem,the buildupof tritiumin the coolantsystem,
and the losses of tritium to environmentalpathways from the coolant. The
tritium is assumedto be chemicallycombinedwith oxygenas T20 or HTO molecules
in the coolant. With this assumption,all calculationsof the tritiumbehavior
are straightforward. The assumptionis easily justified , since all hydrogen
present as gas in the reactor coolant system is recombined with oxygen
continuouslyin the recombiner portion of the off-gas treatment system. The
water produced in the recombinationreaction is re-injectedinto the reactor
coolant system. Since the tritium behaves chemically like normal hydrogen
(protium), there is likely to be only a very small fraction of tritium as
elementaltritium in the coolant.
3.2.7.1 Bases for TritiumTarget Releaseto Coolant
The lithium targets loose tritium through the stainless steel cladding by
diffusionrelatedprocesses. This normal escapemechanismis controlledby the
partialpressureof tritium in the targetrod. There is also the possibilityof
a defect (hole) in the target claddingleadingto depressurizationof the rod.
3.2-8
Normal Permeation Losses
The vastmajority (99.99percent)of the tritiumproducedis containedwithin the
LiAl02pellets and the getter. The partialpressureof free tritiumin the gas
space of the target rod during normal operationis only _ 10.4atm. This partial
pressure is determined by the dynamic balance among the process involved in
releaseof tritiumfrom the pellets,its absorptionin the getter and liner and
permeation through the clad. These processeshave been modeled in the TKTARI
code describedin the literature4. This code, and an evolutionaryupgradeused
in this study, combinetheory and validationexperimentsto give a mechanistic
calculationfor the permeationreleaseof the tritium from the targets during
irradiation. For the design and irradiationparametersof the ABWR target the
calculatedtritiumpermeation(T2P2code) is O.07Ciover the 274 day irradiation
period per rod or 244Ci for the 3,488 rods in the core.
Losses from Failed Rods
No cladding failures occurred during the in-reactor testing and no failure
mechanismswere identifiedother than fabricationflaws,externaldamage and gas
pressure inducedstresswhich is readilyaccommodatedin the design. There was
no mechanical interactionor chemical attack among the cladding, getter and
pellets. However,the possibilityof claddingbreachesin the targetrodscannot
be dismissed
While there is no failureexperiencefor getter-barriertargetrods considerable
data exist for the failureof fuel rods in light water reactors (LWRs). The clad
failure rate for LWR fuel is generallyacceptedto be in the range of 1/10,000
rods (ReferenceI). The target rod failuremodel used in the Phase IA studies
assumed a clad failurefrequencybased on LWR fuel elementexperience and that
50% of the tritium inventory in the failed rod would be released. The 50%
release impliesthat all claddingfailuresexist from beginningof life and that
the capacityfor retainingtritiumis determinedby the solubilityof tritium in
the LiAl02. This was consideredto be very conservativeon the basis that the
most likely target rod failureswould occur late in life when almost all the
3.2-9
tritium would be tied up in the zirconium getter and LiAI02 pellets and not
available for release. However, because the tritium permeation from intact
target rods is so low, the release from failed target rods using this model
dominatesthe tritium sourceterm.
To overcome this limitationa more mechanisticand realistictarget rod
failuremodel was developedas summarizedin the attacheddescriptionof the T2P2
code. The overalleffectof the new model is to delay the targetrod failureto
an exposuredeterminedby the Weibulldistributionfunctionand the frequencyof
reactorshutdown-restartcycles. This reducesthe impactof clad failurebecause
only tritiumreleasedfrom the LiAlO2 after clad failureescapesto the coolant.
This revised failuremodel predicts that 130 Ci of tritium are released to the
coolantper refuelingcycle which for the referencecycle is one year.
Total Iritium Losses to the ReactorCoolant System
The total sourceterm per cycle can be summarizedas 244+130= 374Ci/year. Since
both the normal p__rmeationand the releasefrom failed target rods are expected
to be very gradualthis sourcetermhas beenlinerizedas 1.4 Ci/dayduring power
operationbased on 274 full power days per cycle (i.e 374/274 = 1.4).
3.2.7.2 TritiumDistributionin the Coolant and HVAC Systems
The water inventoryintowhich the tritiumcan mix during operation includesin
the reactorvessel,the feed and condensatesystem,the condensatestoragetank,
the rad waste system and the spent fuel pool. This total mass of coolant
(I.538E+07pounds) is partiallyredistributedduring refuelingto include the
dryer-separatorpool. For simplicityof calculationthe tritium is assumed to
be "well-mixed"in this total inventory. The only differences in water loss
during power operationand refuelingperiodsare no steam leakageto the turbine
buildingduring refuelingand the spent fuel pool exposes 300m2 to evaporation
and the dryer separator pool, which is filled during refueling, essentially
doubles this evaporationarea.
3.2-10
Evaporationfrom the pools is assumed to be controlled by the mass transfer
coefficientfor evaporation(K.)and the pool-to-reactorbuilding air humidity
(watervapor concentrations).Using a literaturevalue of 3000cm/hrfor K,B with
the pool at tOO°F,the evaporationlosseswere calculatedfor a reactorbuilding
refueling platformHVAC flow of 35,000cfmwith the flow assumedto be well-mixed
in the buildingspace. The inlet HVAC air was assumedto be 70°Fat 50 percent
relative humidity. The resulting evaporationrates were 16,4001b/dayduring
power operationsand 31,3001b/dayduring refueling.
The ABWR is designedas a closed systemwith no routineliquiddischargesto the
environment.However, based on experienceit is expected that _ 10,000gal/day
(83,0001b/day) of makeup water will be required to compensate for pool
evaporationand miscellaneoussteam leaks. Based on the pool evaporationrate
indicatedabove the 10,000gal/daymakeup rate impliesthat _ 66,600 Ib/day are
lost through steam leakage, primarily in the turbine building,during power
operation. The loss of water from the discharge of solidifiedrad waste is
negligiblein the overall tritiumbalance.
To accountfor the gradualbuildupof tritiumin the coolantsystem,an unsteady-
state mass balanceof tritiumin the systemwas modeled in differentialequation
form and integratedfor severalcycles of power operationand refueling. In the
model it was assumed that the tritium is present as water and no separation
occursduringpool evaporation.The overallloss rate includedradioactivedecay
but was dominated by the water loss terms. Results of the analysis show the
tritiumconcentrationin the primarycoolant increasingduring the power cycle
and decreasingduring the refuelingoutageas a resultof dilutionwith the water
from the spent fuel _nd separatorpools. However,the tritiumconcentrationin
the primarycoolantreachesa repeatinglevel in two to three refuelingcycles.
The effectivehalf life of tritium in the reactor coolant system is only 0.28
years. The average coolantinventoryover each cycle reachesa steady level of
154Ci or 1.0E-O5Ci/Ib. The modelalso provides a tabulationof tritium source
terms for worker exposure and site dose calculations.
3.2-11
3.2.8 Summaryof Airborne Tritium Releases
Tritium losses to the HVAC system are again assumed to be well-mixed in the
building space to determine the building airborne concentrations. These
concentrationsare also time-averagedover both the power and refuelingperiods
of each cycle. Table 3.2-3 summarizesthe pertinentcalculatedresults needed
for the dose calculations.
Table 3.2-3. Summaryof Tritium Concentrations and Source Terms*
;' J i t i i, iiJl ,
Liquid Phase 154 Ci averagecycle inventory_ ..... 1.0E-05Ci/Ib avera_lecycle concentration
Gas Phase Turbine Building (TB)347 Ci lost**/cycle@ power0 Ci lost/cyclerefueling1.5E-07#Ci/cc in TB @ powerO. #Ci/cc in TB refueling
Reactor Building (RB)31 Ci lost/cycle@ power24 Ci lost/cyclerefueling7.8E-08#Ci/cc in RB @ power
.... ].gE-07pCilcc in RB refuelin9
* - Steady operation (>3 cycles)with 274 days of effectivefull powerfollowed by 91 days refueling.
** - "Ci lost" are time-integratedvaluesthat do not includedecay afterexitingthe stack.
3.2.9 References
I. PNL-8142 "Tritium Target DevelopmentProject Executive Summary iopical
Report", W. J. Apley, September1992.
2. WHC-SP-0840"Topical Report" NPLWR TritiumTarget Design", J. W. Weber,
September 1992.
3.2-12
3. RSA-010 "Assessment of 10% Core Performance During Reactivity Insertion
Accidents Through End-of-Life", B. E. Schmidt, et al, Pacific Northwest
Laboratories, October 1991.
4. WHC-SP-0684 "TKTARI: A Computer Code for Predicting Tritium Target Rod
Performance,"D. R. Wilson, 1991.
5. "Handbook of Chemical Property EstimationMethods, "W. J. Lyman, W. F.
Reehl, and D. H. Roseblatt,1992.
3.2-13
3.3 TRITIUM TARGETFABRICATIONANDRECOVERYFACILITY REQUIREMENTS
The tritium target fabrication methods are based upon work conducted under the
Light Water Reactor Tritium Target Development Program (LWRTTDP) (Reference 1).
The tritium recovery (extraction and purification) methods are derived from
requirements established during the LWRTTDP, but are also based on existing
facility capabilities and purification requirements of the Replacement Tritium
Facility (RTF) at the Savannah River Site (SRS).
3.3.1 Target Rod Fabrication
ABWR targetrod fabricationinvolves:1) producingthe LiAl02pellets,cladding,
getter and liner components,2) assemblingthe pellets, liner and getter tubes
into pencils,3) loadingthe pencilsintothe claddingand 4) completingthe rod
final assemblyand end cap welding. A simplifiedprocessflow diagramfor target
rod fabricationis shown in Figure 3.3-I.
The Light-WaterReactorTritium Target DevelopmentProject (LWR TTDP) utilized
commercialvendorsto fabricatenickel-platedgetter tubes,aluminizedstainless
steel claddingassembliesand the LiAlO2 pelletsutilizedin the TTDP in-reactor
tests. While additional work remains to qualify vendors for full scale
production,a high level of confidencewas establishedthat commercialvendors
could be qualifiedto fabricateall of the target rod components. Discussions
with the FabricationTask Managerfor the LWR TTDP supportthis approach. Itwas
also identifiedthat portions of the fabricationprocessand componentdesigns
may be classified. This was consideredin the evaluationof alternativesand
appeared to be no differentthan many other DOE or DoD contractswhich involve
classifiedinformation.Therefore,the referencecasefor fabricationof tritium
target rods is commercialvendor productionof all target rod components.
The pellet fabrication process begins with enriched lithium carbonate and
aluminum oxide which are blendedand spray dried. After drying, the powder is
calcined in a furnace and a dry binder added using a blender. Multicavity
hydraulicpressespress the pelletsand a belt-feedcontinuousfurnace sinters
Dry Binder -_
LithiumCarbonate
Blend and SprayDry _. Calcine _, Blender _ Hydraulic Press
Aluminum ._
Commercial Vendors
Stainless Steel Tubes --7 Getter -----]
Helium Pellet [_ Pellet Inspection,Atmosphere ." Tube Loading ,_ ,. _ ",, _ _ Sampling, and _" Sintering Furnace
Drying _uoassemolv
Target Assembly
!'
Helium Leak Test
Final Target Rod _. Welding _. and Weld _. Store .-'_ Transfer to FuelAssembly Radiograph Assembly Area
Figure 3.3-1. TargetFabrication Process Flow Diagram
the pellets. The pellets are inspected and sampled and centerless ground to
final size. The pellets are then assembled with getter tubes and liners into
pencils and stored until required for loadtng into the cladding tube assemblies.
No development or qualification for large scale production of pellets was
initiated during the LWR TTDP. A stngle batch of lithium aluminate was
synthesized for the test program. Two methods of pellet fabrication were
utilized and both seemedadequate. Isostattc compaction and uniaxial pressing
were both used. A commercial vendor was used to fabricate approximately 630
pellets by tsostatic compaction.
The barrier-coated 316 stainless steel cladding Is fabricated by a pack
aluminizing process in which a mixture of aluminum alloy powder, alumina powder,
and an ammoniumchloride activator is blended in a predetermined mixture. The
prepared pack is loaded into a clad tube which has the lower end cap already
welded in place. A group of packed clad tubes is then placed in a retort and
heated under conditions that produce the desired barrier coating. Eddy current
and air-gaugeinspectiontechniquesare used to assureuniformityof the barrier
coating. It was determinedin the early phases of the LWR TTDP to pursue the
barriercoating developmentutilizinga commercialcoatingcontractor. It was
assumedthat the commercialvendor would impose a productionorientedapproach
during the developmentthat would enhance future scale up to productionscale.
Initialdevelopmentof the barriercoatingprocesswas performedon 48 inchtubes
due to furnace availability. Further development on full length tubes was
initiated. Several coatingparameterexperimentswere conductedand eventually
coatingparameterswere establishedthat resultedin homogeneousmicrostructures
and uniform thicknessalong the full interiorlength of the tubes.
The nickel-platedzircaloy (NPZ) getter tubes are fabricatedusing commercial
electro-platingtechnologyfollowedby vacuumannealing. The NPZ conceptrelies
on the electroplated nickel layer, highly permeable to tritium and highly
resistantto oxidation,to diffusionbond to the Zircaloy,providinga means of
entry for the tritium into the Zircaloy storage media. A commercial tubing
supplier, capable of making acceptable getter tubing by redraw of existing
reactor grade Zircaloy-4 fuel cladding tube stock, was found through the
competitiveprocurementprocessduring the LWR TTDP.
3.3-3
The inner liner is a roll-formed tubular shape of O.O04-inch zirconium metal.
Metal shapes of this geometry are routinely fabricated in industry and a
commercial vendor should be readily available.
The final rod assembly consists of loading subassemblies in air on a work bench.
Loaded rods are inserted manually into a portable evacuation/drying magazine.
Each magazineholds a one-day throughput. The internal atmosphere is replaced
with dry, high purity helium. Each magazine is allowed to soak until the
components are within acceptable limits for moisture and oxygen. The magazine
would then be connected to a welding enclosure and flushed with high purity
helium. The springs and top end caps are inserted into the tubes and the end
caps welded. Whenall the end welds are complete the magazine is uncoupled from
the welding enclosure and the rods removed and placed in the helium leak-test
chamber. Following final inspection,the rods are placed in containers and
transportedto a bundleassemblyareawhere targetrods, fuel rods and associated
hardware are assembledinto the fuel bundle for shipment to the reactor.
With commercialfabricationof the targetcomponents,the onlytarget fabrication
operationsto be performed as part of the target rod assembly are:
- Target rod componentinspectionand storage
- Pencil assembly
- Rod loading
- Helium back fill
- End closure assembly and welding
- Final rod inspectionand packaging.
Based on the design developed for the LWNPR support facilities,approximately
15,000 target rods per year can be fabricated in an industrial-type,single
story, 15x30m building. The ABWR would require only 25 percent of this
throughput. However, the need to handle 4m long rods horizontallyrequires
approximately 10m of working length. Allowing space for component storage,
inspection,helium back fill and welding equipment, a total of approximately
300m2 should be more than adequate for target rod assembly.
3.3-4
The equipmentrequired for target rod assembly includes:
- Two 5m long assemblytables
- Five evacuation/drying magazines
- Helium fill station equippedfor pump-downand back-t'ill
- Helium glove box with two automatedGTAWmachines- Helium leak testing and weld radiography equipment
- Miscellaneous inspection tools and instruments.
The targetrodassemblyoperationis a normalindustrialactivity.Thereareno
specialsafety,licensing,workerhealthand safety,effluentor waste issues
associatedwith this activity.The onlywastesgeneratedthatare potentially
regulatedunderthe ResourceConservationandRecoveryAct (RCRA)arethe spent
radiographyfilm processihgchemicals. Thesewasteswould be collectedand
packagedas requiredfor disposalby a commercialche_'calwastecontractor.
Safeguardsand securityassociatedwith targetfabricationis relatedto the
enrichedlithiumpelletsandpotentially,protectionof anyclassifieddocuments
related to the fabricationprocess. The protectionand accountability
requirementsassociatedwith enrichedlithiumwhich is classifiedas "Other
NuclearMaterial"(Category4, AttractivenessLevelE) underDOE Order5633.3A
(Reference2), arelessstringentthanfissilematerialandaresimilarto those
associatedwithpreciousmetalsin the DOE system.
Finalassemblyand inspectionof the targetrods couldbe handledin several
ways. Thereappearto be threeprudentoptions.PhaseIAdiscussedassemblyof
the componentsin the MOX Fuel FabricationFacility. This is stillvalidand
expansionspaceis beingretainedinthecurrentlayoutof theplant.Additional
optionsconsideredin thisphaseincludeutilizinga commercialfuel fabricator
to performthe targetassemblyor performanceof thistaskat a DOE facility.
Theremay be politicaladvantagesfordecouplingthe tritiumtargetfabrication
fromtheMOX fabrication.Thisis particularlytrueif theMOX Plantwas placed
under internationalinspectionas part of a bilateraldisarmamentagreement.
This couldeasilybe accomplishedutilizinga qualifiedcommercialvendorand
3.3-5
implementing the security measuresrequired to protect classified information.
There are manyfavorable aspects of targe' assemblyutilizing a commercial fuel
fabricator. These fabricatorsare intimatelyfamiliarwith fabricationof
assembliesof similarphysicaldimensionand components.The fabricationwill
requiresimilarweldingand inspectioncurrentlysupportedin existingfuel
fabricationfacilities.QualityprogramsincorporatingNQA-Irequirementswill
be very similar. Physicalsecurityof thematerialswill alsobe similar. If
thedesignof the targetrod is classified,additionalmeasureswillneedto be
implementedatthecommercialfacility.However,the incrementalcostassociated
with classificationshouldbe less if the targetrods are fabricatedat a site
where personnel and physical security measureswere already being imposed.
It is also feasibleto performthe targetrod assemblyat an existingDOE site
in a modifiedor newfacility.Thefactthatthereare no radioactivematerials
involvedimposestherequirementthatthefabricationfacilitiesbe cleanrather
thanexistingfuelfabricationfacilitiesthathavebeenshutdownunlesstheycan
be adequatelydecontaminated.An existingDOE sitewouldhavethe advantageof
in-placeinfrastructurefor safeguardsand security. A decontaminatedfuel
fabricationfacilitywouldalsohavethe advantageof experiencedpersonneland
technologyrequiredfor the target assembly,along with stringentquality
programs.
All informationacquiredsupportsthe use of commercialvendors for the
productionof componentsfor the tritiumtargetrods. Italsoappearsfeasible
to utilizea commercialfuelvendoror DOE fuelfabricationcapabilitiesforthe
assemblyof completetargetrods. It is notedthatsomeadditionaldevelopment
work is requiredpriorto productionscalefabricationof LWR TritiumTarget
Rods. Theseare addressedin Section6.3.
3.3.2 TritiumRecoveryFacilityRequirements
The phaseIA report(Section6.6) (Reference3) describedthe processrequired
to extract tritium from the ABWR target rods and performthe necessary
purificationand isotopicseparationsto obtainthedesiredproductquality.It
also includeda descriptionof the supportsystemsneededto minimizeworker
3.3-6
exposure and environmental releases. Except for the head end extraction
operation, all of these process capabilities are included in or provtded for the
Replacement Tritium Facility (RTF) whtch has just been completed at the SRSand
is currently undergoing operability testing.
It was determined during meetings with SRSpersonnel not to introduce the new
extraction off-gas into Building 232-H because tts useful life cannot be
guaranteed for the life of the PuDisposition project. The 232-H factlity is 36
years old and Pu Disposition introduces a need in about ten years for a life of
about 30 years. The level of upgrades and modifications required to retain the
232-H facility product extraction capabilities would not be cost effective.
Previous studies have been conducted addressing addition of extraction capability
to the RTF. The "Replacement Extraction and Purification Facility (REPF)--
Building 231-H" Functional Performance Requirements was issued February 16, 1990
(Reference 4). This project proposed extensively greater purification
capabilitiesthan will be proposed for the Pu Destructionmodification. It is
noted that some of these other modificationswill still be needed for the RTF to
operatewithout the supportfunctionsnow providedby the 232-H facility. These
are being addressedin anotherrequirementsdocumentfor the ReplacementTritium
PurificationFacility (RTPF) (Reference5). The RTPF is currentlyunfundedand
studiesare being conductedto justifyupgradesfor extendingthe use of Building
232-H to the year 2050 for off-gastreatmentwithoutextraction (resultingin a
minimal source term within the old facility).
If the RTPF were funded to provide replacementpurificationcapabilityfor the
aging 232-H facility,it would be most prudentto add the LWR Target Extraction
Facilityto it (resultingessentiallyin the original REPF but with LWR rather
than HWR extractioncapability). Consideringthe unfundedstatus for the RTPF,
there are three viableoptions: I) a greenfieldfacilityat the reactor site to
extract and purify tritium gas prior to gas shipment to the SRS tritium
facilities,2) the identicalgreenfield facility adjacent to the SRS tritium
facilities for extraction and purification,or 3) a new extraction hot cell
locatedadjacentto the RTF with supplementalpurificationcapabilityinstalled
within the expansion area of the RTF. The cost of options I and 2 would be
3.3-7
similar for any location and for any of the LWRoptions being considered. This
cost will not be specificallyestimatedsince it will be comparablefor all LWR
options. It is anticipatedthat the greenfieldfacilitywould be downsizedfrom
the REPF which was sized to supportRTF and 100% of goal tritium productionin
1989 (capitalcost ,,$200M).
;" Performingthe extractionof the SRS requiresonly a few annual spent fuel cask
shipmentsof irradiatedtargetsand the materialshipped is less attractivefor
theft or diversion than a purified tritium product. See Section 4.4.7 for
discussionsregardingthe transportationoptionsconsidered.
Since the gas purificationand isotopicseparationsystemsplus their associated
supportsystems(blanketgas,purge gas, etc.) representthe majorityof the cost
of a LWR target processingfacility,the recommendedapproach is to add a target
extraction facility (with a shielded hot cell) to the north end of the RTF.
Pretreatmentcapabilitywill also be added in the expansionarea in the north end
of RTF, becausethe existingRTF diffuserand TCAP capabilitiesare specifically
for processing returns without protium. The RTF TCAP currently separates
deuteriumfrom tritium. It is anticipatedthat th_ additionalsupportservices
required for this approachwill tax the existing RTF support systems to their
originaldesign capacitieswhich were sized for future expansion.
Figure 3.3-2 shows the 200-H Area Tritium Facilitiesand the proposed location
of the ExtractionFacilitynorth of the 233-H building. The extractionfacility
is sited adjacent to stairwellS-2 such that a doorway from the grade level
landingof the stairwellcan enter directly into the shieldedoperatingarea so
that separatesecurityaccesscontrolis not requiredfor personnelaccessto and
from the extractionfacility. A new railroad spur is indicatedto providerail
car access for standard spent fuel shipping casks which will be used for
transportof the target rods. Figures3.3-3 and 3.3-4 show the plan view and
elevationof the targetextractionfacilitythatwould be locatedadjacentto the
expansionarea on the north sideof the RTF. As indicated,the buildingconsists
06 a 18x20xgmhigh reinforcedconcretestructureenclosinga 10x10xS.Smhigh hot
cell and a rail car bay for receiptof standardLWR spent fuel casks. The rail
3.3-8
233-IH
Figure 3.3-2 Tritium Facilities
3.3-9
i 20.0m i
10.0m i
/ Railroad SteelFrame4 ft.ReinforcedConcrete
Ful'll_C¢
Steel Wall
AirLock
StorageHolesE
0
06Shidded
M_" MagazineStorage DoorC>
TransferCanUnloading FurnaceMagazineLoading
Figure 3.3-3. Tritium Extraction Cell Plan View
i 20.0 m iI00Ton Crime
HVACSpace
f:: 5TonCrane
C)
Railroad EBay _
Hot Cell
/ ShieldPlugs Gallery
.........GRADE............_,........!
100SteelStorageTubes6"Dia.x 14'Ling
Cask TransferTmmel
Figure 3.3-4. Tritium Extraction Cell -Elevation
car bay ts equippedwtth a 100 ton crane and a below-grade cask transfer tunnel
that provides access to the top loadtng spent fuel casks through the hot cellfloor.
The hot cell ts dtvtded tnto an atr-ce11 and a nitrogen-cell by a metal
partition. The partition ts less than full height wtth a closed top makingthe
furnace portion of the cell a sealed nitrogen-filled box. Thts permits the tn-
cell crane to operate over the nitrogen boxwhich ts equippedwith access ports
for furnace removal. The nitrogen box can also be equippedwtth a wall or
ceiling mounted light duty electro-mechanical manipulator tf needed. The
partition also contains an air lock for transfer of furnace magazinescontainingtarget rods to and from the nitrogen blanketed furnace cell.
The process gas, blanket gas and the argon purge streamswould be connectedto
the corresponding RTFsupport systemsfor processing andtritium recovery. The
extraction furnace system includes valves, gas receiver tanks and vacuumpumps
which would be located in the RTFexpansionarea as shownin Figure 3.3-5 along
with the required pretreatment systemsto removeimpurities from the furnace off-
gas stream prior to introduction into the existing RTF processes. The furnace
off-gas in the receiver tanks will be sampledutilizing a 200 ft capillary tube
to the Hass-Specglovebox in room010. This samplewill be for accountability
along with verification that the extraction is complete. It is anticipated that
the small amountof impurities (i.e. protium and hydrocarbons) introduced into
the RTFfrom the samplebefore pretreatment will be tolerable.
Pretreatment capabilities will be addedto the RTFin the expansionarea at the
north end of the facility. Pretreatment is required to removeprotium from the
product gas. This will require gas separation and isotopic separation. The
process gas from the extraction receiver/hold tank is first pumpedthrough a
heated uraniumbedto decomposethe oxides of tritium andhydrogen. Gasfrom the
decomposeris cooled and pumpedthrough the primary uranium hydride bed to
collect the hydrogen isotopes. The effluent gases (primarily helium and
impurities) are sent through a palladium/silver sacrificial bed to remove
impurities that could damagethe diffuser, then pumpedthrough a multistage
3.3-12
I l_nt Northt I New Pretreatment Area _ ([_
IL 1 _ __,_ New Extraction Hot Cell
0441
Figure 3.3-5. New Pretreatment Area in RTF
diffuser to recover the remaining hydrogen isotopes. The gases that flow through
the diffuser tubes, primarily helium, are collected in one of two tanks which can
also be sampled to determine tritium content. These gases are sent to the RTF
off-gas cleanup system where they are further treated, if necessary, or sent to
stack discharge. The hydrogen isotopes, which diffuse through the tubes, are
collected dtrectly on the secondary urantum hydride bed. Hydrogen isotopes are
released by heating the hydride beds, then pumped to the thermal cycling
absorption process (TCAP) feed tank. The TCAPcycles the gas between two beds
of palladium-coated kteselguhr (Pd/K). One bed is heated to drive off the gases
while the other is cooled to promote gas absorption. The palladium-coated
kteselguhr preferentially retains the tritium and thus separates it from the
other hydrogen isotopes (prottum). Whenthe pressure in the TCAPproduct tanks
reaches the desired level, the gas is assayed and pumpedto the TCAP Product
storage (2-500 ltter) or TCAP Feed storage (2-500 liter) beds in the TCAP
glovebox (T105-300-1) located in Room017 of RTF. The raffinate material is sent
to the RTF off-gas cleanup system for removal of residual tritium and is then
stack discharged.
Room ventilationfor the operatingareas of the new extractionfacilitywill be
provided with the building. It is planned that the RTF stack (and effluent
monitoring system) will be utilized for all gaseous discharges from the
extractionfacility. The nitrogenblanketedfurnacecell will be connectedto
the existing stripper system in RTF. The air portion of the hot cell will be
HEPA filtered (to remove potential particulate materials introduced during
storage or transport of the irradiatedtargets) prior to discharge to the RTF
stack.
The air cell includes storage holes for 125 percent of the annual reactor
dischargeof targetrods, an unloadingstationfor the transfercans used to ship
target rods from the complexto SRS and a furnacemagazineloadingstation. The
nitrogencell includestwo vacuum furnaces,their associatedsupportequipment
and the transfermechanism for loadingand unloadingthe furnaces.
3.3-14
The spent target rod after extraction Is expected to retain approximately -50
curies of tritium in the LtAI02 pellets and have a gammadose rate in the 100
R/hr range at one foot. Interim storage for spent target rods at SR$or other
provisions (e.g. return to the reactor site) wtll probably be required if the
spent target rods are disposed of as core componentstn the federal spent fuel
repository. Althoughdtsposal of the spenttarget Podsin the federal repository
ts permissible (Reference 6), this maynot be as cost effective as dtsposal asa low-level waste at the SavannahRiver Stte (see Section 4.4.7 for further
discussion).
3.3.3 Tritium Cost Considerations
TaraetFabrication
TargetfabricationcostsinthePhaseIAreportwerebasedon estimatesdeveloped
duringthe LWR TTDP. No new issueshave been identifiedwhichwouldwarrant
changingthis estimate. Target assemblyat a vendor locationshould be
comparableto the estimatedcostto assemblethem in the MOX Plant.
TritIum Recoyery
Tritiumrecoveryhasseveraloptions.Acommon optionforall LWRfacilitiesis
a greenfieldextractionand purificationfacilitythat shouldbe in the $200M
capitalcostrangesimilarto theREPFprojectthatwas developedby SRS. This
greenfieldfacilitycouldbe locatedat the reactorcomplexor at SRS nearRTF.
One differencebetweenthe two locationswould be transportof the tritium
productversusirradiatedtargetrodsto the SRS.
Our referencecase is the additionof an extractionhot cellnextto the RTFat
SRS as proposedand costedin the PhaseIA report. An additionalmodification
to the RTF is requiredin our PhaseIC referencecasethatwas not includedin
the PhaseIA reportestimates.This is the "PretreatmentFacility"locatedin
the expansionareaof the RTF alongthe northend of the building. This will
includetwogloveboxesthatcontainuraniumhydridebeds,a diffuserand a TCAP
3.3-15
wtth capacity to process 38 mtllJon curies of tritium per year. Based on
discussions with SRS personnel, the estimate for this modification to the
existing RTF ts in the $15-2SM range. This was not tncluded in the ortgtnal
Phase 1A estimate.
3.3.4 References
1. PNL-8142 "Tritium Target Development Project Executive Summary Toptcal
Report", W. J. Apley. September 1992.
i
2. DOE Order 5633.3A "Control and Accountability of Nuclear Materials",
Department of Energy, February 12, 1993.
3. NEDO-32292, "Stuay of Pu Consumption in Advanced Light Water Reactors,
Evaluation of GE Advanced Boiling Water Reactor Plants", GE Nuclear
Energy,May 13, 1993.
4. WSRC-RP-gO-147"FunctionalPerformanceRequirementsfor CurrentAppraisal
of Cost, Project S-3312, Replacement Extraction and Purification
Facility(REPF)(U)Building231-H", H. W. Harmon,January 30, 1990.
5. WSRC-TR-91-442 "Functional Performance Requirements for Replacement
Tritium PurificationFacility(RTPF)(U) Building 231-H, ProjectS-4568",
H. W. Harmon, July 15, 1991.
6. IOCFR Part 961 "Standard Contract for Disposal of Spent Nuclear Fuel
and/or High-level RadioactiveWaste", Department of Energy, January I,
1992.
3.3-16
3.4 ABWR-PDR PLANT OPERATION FORTRITIUM PRODUCTION
3.4.1 Introductioni
During phase la of this project (Reference i), the option ofconvert the reactor system to produce tritium was evaluated.In was shown that tritium production goals can be met by theuse of four lithium aluminate target rods per assembly in astandard UO 2 fueled ABWR core. Rod design, fabrication andproduction considerations were described along with anestimate of the impact on plant operations and effluents.Support facilities associated with tritium production alsowere described. The results of these investigations clearlyshowed that goal quantities of tritum can be produced withthe ABWR.
The current evaluations in Phase ic address the use of a
mixed oxide (MOX) reactor core for the purpose of disposingof weapons plutonium while also inserting Lithium targetrods for the production of tritium. This section discussesthe impact on plant operations associated with the tritiumproduction mission.
3.4.2 Basis for Review
Two goals, to be satisfied concurrently, have beenestablished by DOE. I) the disposal of I00 MT of Plutoniumpreviously used in the US weapons program, within 25 yearsand 2) the production of Tritium on an "as-needed" basis tosupport long term DOD needs.
The planned operation consists of two distinct operatingcycles: a plutonium disposal cycle in which MOX bundles areburned in approximately 18 month operating cyclesconsistent with the power production and disposalobjectives; and a tritium production cycle in which Tritiumtargets are exposed in a MOX fueled core for about 1 year.Steam produced is used to produce electricity which is sold
to a local utility for transmission. Estimates of keyexposure parameters are summarized in Table 3_4-i.
During the Plutonium Disposal Cycle, about 232 MOX bundlesare removed to the reactor pool (and fresh fuel is loaded)approximately every 18 months. The removed fuel is allowed
to cool in the reactor pool for about 1 year before beingtransported to a high level waste depository or a temporarystorage facility.
3.4-1
When operation of a unit is redirected for tritiumproduction, a full core offload and reload is required.Partially spent MOX fuel is stored in the fuel storage poolwhile a new core with targets is inspected, channneled and
loaded into the core. To provide room in the spent fuelpool, transport of previously discharged MOX fuel may berequired in preparation for the Tritium Cycle.
Following the Tritium Cycle, the entire core is offloaded tothe spent fuel pool and the previously operating MOX core isreloaded into the core to continue the plutonium disposalCycle. Tritium targets are removed from the exposed bundlesin the reactor pool and transported to an offsite extractionfacility. The residual MOX bundles (without the targets)are reconstituted for use in a future cycle.
3.4.3 Preparation for Tritium Production
3.4.3.1 Fuel Inspection and Handling
Section 3.2.3 of reference 2 discussed the potential needfor underwater inspection of fresh MOX assemblies. Althoughfresh MOX fuel may be mildly radioactive, no specialhandling is anticipated. To avoid design modification toaccomodate underwater inspection in the new fuel vault,fresh MOX fuel will be inspected and channeled in the samemanner as current fuel designs. If occupational exposurebecomes a concern with this approach, underwater inspectionin the spent fuel pool may be persued as an option.
Addition of the target rods to the MOX fuel bundles is notexpected to require special handling beyond that used forthe recovery of the exposed targets. Therefore nomodifications or other considerations are needed.
3.4.3.2 Fuel Handling
Initiation of the tritium production cycle requires a fullcore offload of the previously operating core and storageuntil the tritium production cycle has been completed.
The need for full core offloading places an added burden onthe fuel storage pool in the ABWR-PDR design. The currentdesign provides for storage of about 270% core loading(Reference 3) or about 5 cycles of operation withoutaccomodating the Tritium production mission. To provide fora full MOX core offload and Targeted core onload, less spacewill be available for spent MOX storage. The current ABWR-PDR design will accomodate about 2 cycles of spent MOX.Therefore, in preparation for Tritium production, it is
likely that offsite shipment of at least 3 cyc__ _ of spentMOX to the waste disposal facility will be required. It isassumed that such facilities are available and that shipmentcan be accomplished within the six month preparation period.
No impact of the target rods is expected on the overall fuelloading or refueling process since the targets are containedin the fuel assemblies. However, due to the need for fullcore offload of the tritiated core and reload with the
previously operating MOX core, care must be exercised toassure that fuel assemblies are well tracked and locations
verified. Enhanced tracking of fuel bundles and theirexposures, with and without targets, has a higher importanceto meet the concurrent goals of disposal and tritiumproduction. Equipment such as LASERTKAC may be beneficialfor this ABWR-PDR use. Such equipment is not currrentlyincluded in the ABWR certified design, but could beconsidered for future development.
3.4.3.3 Fuel Pool Cooling
The ABWR fuel pool cooling and cleanup system (FPCC) isdesigned to accomodate heat removal from 35% an offloaded
core 21 days after shutdown and the spent fuel from 4previously offloaded cycles (Reference 3). If higher heatloads exist, the RHR system may be aligned to providecooling assistance.
Following the initiation of a Tritium production cycle, thefull offloade MOX core will require additional cooling.Therefore, during the some portion of the Tritium cycle, oneloop of the RHR system will be aligned in Fuel Pool assistmode. No plant modifications are needed to accomplish this.Operability of this RHR loop in other operating modes (suchas Low Pressure Co£e Flooding or Suppression Pool Cooling)will rely upon manual realignment, as required.
3.4.3.4 Other Design Considerations
The MOX core design results in a higher fast flux and lowerthermal flux from the MOX core in comparison with uraniumfueled cores. Initiation of tritium production decreasesthe effect. Because of these differences, certain systemsare potentially affected.
Neutron Monitoring/Calibration, Process computer heatbalances, SLC boron requirements and the radiationmonitoring systems were reviewed to determine if other
design considerations would be affected by initiation of thetritum production cycle. No significant differences were
3.4-3
identified which would either require a design modificationor increase operating/maintenance costs.
3.4.4 Operation during Tritium Production Cycle
3.4.4.1 Operating Limits and Power/flow map impact
Section 2.8.1.2 of reference 1 indicated that a higherminimum recirculation flow (35% Speed / 45% Flow) is neededwith the MOX core to avoid the instability region.Analysis of changes resulting from initiation of tritiumproduction have not been conducted, but slight changeJ arenot expected to result in operational difficulties.Specific procedural changes reflecting changes in theoperating limits, rod patterns and the power/flow map willbe needed and appropriate operator training provided priorto plant restart for the tritium production cycle.
3.4.4.2 Occupational Exposure and Routine Offsite Releases
Table 8.2 of reference 1 estimated the average annualexposures from the ABWR at about I00 Manrem (not includingoperation of the fuel fabrication facility). Tritiumleakage was not explicitly included. Initiation of thetritium production cycle would be expected to increase thedose recieved at power and in the turbine building due topotential for required breathing apparatus in some areas andthe resultant inefficiency associated with the maintenancework. In addition refueling activity may be slightly lessefficient due to airborne tritium above the refueling pool.If the exposures recieved in these activities are increasedby 10% due to the inefficiencies, the total annual exposurewould be increased by about 2 man-rem. Therefore, asignificant increase is not expected due to initiation ofthe tritium production cycle.
Additional discussion of the routine offsite releases from
steam leaks and evolution from the refueling pool isdiscussed in Section 3.2.
3.4.4.3 Abnormal Occurances
Target Rod Failure Impact
Evaluation of tritium rod performance (Section 3.2)indicates that the internal rod pressure due to helium gasis appoximatley 2300 psig following one exposure cycle of273.75 days. The stress from such pressure on the targetrod, however is a small fraction (approximately 10%) of theyield stress for the rod under operating conditions. It can
3.4-4
therefore be concluded that failure is not likely. However,if failure were to occur, the impact of such sudden failurewould be minor since the accumulated helium has no
significant impact on the operation of the reactor or plantsystems. On the other hand, the sudden failure of a targetrod could have an adverse impact on plant performance if itleads to a significant release of tritium comparable withhydogen water chemistry (HWC) upsets or resin intrusionevents.
HWC operation injects approximately 2 ppm of hydrogencontinuously in order to suppress the radiolysis occuring inthe reactor core. For the ABWR with full feedwater flow of
17.1 x 106 Ib/hr, about 34 ib/hr of hydrogen would beinjected into the core. Since, based on the evaluations inSection 3.2, the amount of tritium available for release
from a single rod very small (about 6x10 -9 grams), it isclear that failure of a single rod would lead toinsignificant amount of hydrogen gas.
Decay heat Implications
Figure 2.7-19 of reference 1 shows the expected Decay heatfrom an equilibrium MOX core. Because exposure from thetritiated core would be on a fresh core, decay heat levelsfollowing any accident or transient would be significantlylower and, because of the nuclear differences, the MOX core
would have a slightly lower decay heat than a comparablesize UO 2 core. Because of these considerations, shutdowncooling and other shutdown safety issues are less severe inthe tritiated core and during the plutonium disposal cycle.
3.4.5 Termination of the Tritium Production Cycle
Fuel Handling Issues
No modifications are considered to be necessary for thehandling of the tritiated core.
Target Rod Removal
Tooling and procedures to remove exposed target rods fromexposed fuel assemblies within the ABWR-PDR pool need to bedeveloped and tested.
Fuel Reconstitution
Following the tritium production cycle, the remaining MOXbundles exposed along with the lithium targets will have
significant life remaining for additional burnup. Tooling,
3.4-5
procedures and methods to permit reconstitution of thesebundles for additional exposure have not been developed andtested at this time. However, no impediments to developmentof such methods have been identified.
3.4.6 References
I. "Study of Pu Consumption in Advanced light WaterReactors", GE Nuclear Energy, NEDO-32292, May 13, 1993.
2. "Study of Pu Consumption in Advanced light WaterReactors, Compilation of Phase Ib Reports", GE NuclearEnergy, NEDO-32293, September 15, 1993.
3. "Safety Analysis Report, Advanced Boiling WaterReactor", 23A6100 Rev i, Amendment 31.
Table 3.4-1
Key Parameter Study
...... Tri titan Plutonium
Production DisposalCycle Cycle
Number of Reactors 1 6
Cycle length (EFPD) 273.75 392.2 .....MOX Bundles 872 232
Dis charged/cyc Ie
Plant Capacity Factor 75% .... 75%Discharge Exposure N/A 37081 ......(MWd/MT)
100MT Plutonium Disposal N/A i9.3Time (years)Tritium Production ii,595 N/A(Ci/rod) - 3488 rods
t
4.0 INFRASTRUCTURE AND DEPLOYMENT
4.1 MOX FABRICATION INFRASTRUCTURE
4.1.1 JAPANESE FACILITIES
OVERALL SUMMARY:
1. A plant to reprocess LWR fuel, privately funded by Utilities (75%) and
Industry (25%), is being located at Rokkasho in northeast Japan.
2. A privately funded MOX plant that will utilize the reprocessed Pu, to serve
both BWR and PWR reactors, is still in the planning stage.
During preliminary meetings of the industry/utility group it was decided that
while the plant will be highly automated, it will be built to allow non-remote
maintenance if required.
3. Reprocessing of Japanese LWR fuel is expected to be carried out in Europe
until about 2002, after which both reprocessing and MOX fuel fabrication will be
shifted to Japan.
The utilization of Plutonium in thermal reactors has not yet approached a
practical stage in Japan. At present Japan has only a single small-scale MOX fuel
fabrication facility for LWR's. This facility is under the administration of the Power
Reactor and Nuclear Fuel Development Corporation (PNC) and is located at the
Tokai Mura site north east of Japan. A new large scale MOX fuel fabrication facility
will be required before full scale utilization of plutonium in thermal reactors can
progress.
Over the past year, the nuclear community in Japan has been studying the
requirements for a MOX fuel faciity to to realize the full commercialization of
Plutonium in LWR's. The MOX fuel facility has been targeted for start-up in the early
years after the turn of the century. The plant capacity is estimated to be around
100 MT MOX per year. The location of the facility has yet to be decided but it is
anticipated that the plant would be located near the reprocessing site in Rokkasho
Mura in northern Japan.
Early in December a visit was made to Japan to obtain more detailed
information on the status of the Japanese MOX program and in particular the MOX
fuel fabrication facility study. The details of this visit are covered in the next section.
4.1 .I-I
Visit with Hitachi, December 1, 1993:
Japanese MQX Pv0grarn"
The Japanese program, in the MOX area was described by Dr. M. Oguma. Hitachi is
involved in a number of areas of the Japanese MOX program, including MOX fuel
designs for the reactor, MOX factory design and licensing, MOX shipping container
designs and as part-owner of the reprocessing facility. A facility to reprocess LWR
spent fuel has been approved and is being built at Rokkasho in northeast Japan. This
facility, Japan Nuclear Fufels Ltd (JNFL), is a consortium with 75% funding from
utilities and 25% funding by the industry. A MOX factory, to serve both BWRs and
PWRs, is still in the planning stage. At the time of this writing, this factory was also
to be privately funded, however, there were indications that there was some
consideration given to government ownership of this factory.
The reprocessing factory was expected to ship a master blend of 50% UO2- 50%
PuO2 to the MOX factory.
The proposed schedule for Pu utilization is shown schematically in Figure 1. Until the
year 2002, LWR fuel will be reprocessed in Europe, in a number installations
(including British and French plants in the future). Beyond this date, it is expected
that the entire LWR spent fuel load would be reprocessed in Japan.
All facilities will be subject to IAEA inspection standards.
Question and Answer Session:
The following answers were given to specific questions during the Q&A/DiscussionSession:
Process:
The MOX factory will employ a standard mechanically blended MOX pellet
fabrication, similar to the refrerence process identified by GE for the disposition study.
The input will be the 50-50 master blend and the output will be full (BWR or PWR or
Fast reactor) bundles.
4.1 .I-2
4j_ ~ 70 MT/yr
Mox JapanTonnage European _ .f Reprocessing
1 I I I I I1993 95 97 99 01 03 05 07 09
Fig 1 Schematic of Proposed Japanese Plan for LWRMOX (reprocessed)Fuel Uti:ization
4.1 .I-3
Throughput:
The peak MOX factory throughput is pegged at 100 MT of MOX per year. The
average capacity of the MOX plant will be 70MT of MOX per year which should
satisfy the LWR recycling need.
MOX Fuel Fabrication Experience, Fuel Additives:
Only PNC has practical MOX fuel fabrication experience. The Hitachi/Toshiba
designs are based on input from PNC and from visiting European installations. They
have no experience with poison additives to MOX fuel other than Dysprosium used in
small quantities for power shaping in the MOX fuel bundles irradiated in Tsruga 1 unit
in the 1980s. The poison additives proposed for the U.S. Pu disposition project are
used for reactivity control rather than power shaping.
Level of Automation:
The Japanese had visited MOX facilities in Europe and thought the
Belgonuclaire plant had the maximum throughput although they were not the most
highly automated. Hitachi and Toshiba with JNF are the MOX factory designers and
stated that even though they started with a very high level of remote operation, they
have since adopted a strategy that will allow "non-remote" maintenance.
Transportation:
Transportation from Europe, in the form of completed assemblies for LWRs or
for the fast reactor, has already been designed. It is expected that the MOX factory
will be co-located with the reprocessing factory, however, this decision has not been
made. If they are not co-located, they do not see any problem in shipping the master
blend MOX powder as this is "routinely" done in Europe. Hitachi and Toshiba are
presently designing the transportation casks for shipment of completed MOX bundles
from the MOX factory to the reactor sites. In most instances, they expect this
transport to be carried out over water routes rather than inland routes. Casks for sea
transportation are already available and those for road transport are under design.
4.1.I-4
These designs might be applicable to a US program.
Safeguards:
Hitachi admitted that although they are in the process of preliminary design for
the MOX factory, they have not completely incorporated Safeguards requirements into
the design process. All the plants will be subject to IAEA inspection and standards.
Hitachi allowed that only PNC had real experience in material accountability and
safeguards and that any new technology for implementing safeguards would have tocome from PNC.
Licensing of the MOX factory:
Government (MITI) is the process of establishing the licensing criteria. It is
likely to be modeled after European standards and could possibly be used as a basis
for U.S. certification.
Licensing of the Reactor plan for MOX:
Hitachi is doing MOX fuel designs. Only partial MOX core loads (up to a third
of the core) are envisioned. They do not see any problems nor the need for lead tests.
The nuclear parameters were verified with MOX fuel test assemblies in the Tsuruga 1
reactor in 1986. These assemblies were discharged in 1990 with 25000 MWD/MT.
The MOX assemblies are not located at peak power positions and therefore can attain
the same exposure as the urania bundles over a longer period of time. The bundles
will be full MOX designs and not "island" MOX designs.
It was Hitachi's view that for Pu fractions up to 5%, no additional data was
needed. Beyond about 10%, Hitchi felt that there were sufficient changes to the fuel
properties (e.g., solidus/liquidus temperature)and that more detailed assessmentswould be needed.
4.1.1-5
Waste:
The waste from the MOX factory will be mainly a result of the scrap during
fabrication. It is to be divided into "clean" scrap and "dirty" scrap. By definition,
"clean" scrap will be sent to an appropriate entry point in the fabrication process to be
recycled and blended into the process flow. "Dirty" scrap which cannot be.blended
into the process flow and which quite possibly requires solutioning with resulting
aqueous streams, will be returned to the reprocessing facility. Thus, they do not feel
there will be any real waste stream from the MOX factory itself and all the waste will
be handled as part of the waste stream in the reprocessing facility.
Hitachi did not provide any cost estimates. The schedule called for about 8
years to complete the MOX factory from project initiation to initial bundle production.
Of this, 3 years were for design and licensing, 3 years for construction and 2 years
were reserved for test runs. They anticipate only 1 shift operation while in production.
Visit with Toshiba, December 3, 1993:
The role of Toshiba is parallel to Hitachi in the Japanese MOX program. A visit was
arranged with Toshiba for two reasons. First, in Japan, it would not be considered
proper to visit only one of these two industry participants, even if all the relevant
information would have already been obtained from the first meeting. Second, it
served to confirm the details obtained at the first meeting.
Japanese MOX Program:
The Japanese MOX program was described by Toshiba, including their role, as
a fuel MOX designer, fuel factory designer, and transportation cask designer. They,
like Hitachi, are investors in the reprocessing facility and expect to invest in the MOX
factory. They clarified that some of the Pu from reprocessing will be used by PNC for
the fast reactor. During 1991-92 time period, concepts for the MOX plants were
examined and they expected a MOX plant to produce about 70 MT of MOX per year
on the average, 45 MT for BWRs and 25 MT for PWRs. The MOX factory design is
still in the planning stage. Most of the work is being carried out by Toshiba, Hitachi
and JNF. They have completed a layout, equipment requirements and process flowsheets.
4.1 .I-6
All the questions asked of Hitachi was also asked of Toshiba. Toshiba
confirmed Hitachi's answers in each of the areas. Specific areas where Toshiba was
able to provide some additional information are noted below:
Hot-Cell vs. Glove Box Type Lines:
To accommodate both BWR and PWR bundles, it would be necessary to •
change the enrichment frequently. GE pointed out that this could lead to considerable
down-time in cleaning out the line and complying with material accountability. Better
availability could be obtained with a combination of glove-box line type with a hot-cell
type set up where the entire contents of the hot-cell could be swapped with a standby
unit, to minimize downtime and enhance accountability. They (Toshiba) have not
considered these areas in detail.
Cost and Schedule:
Toshiba indicated the cost of the MOX plant to be in the neighborhood of US$1
billion. It was noted that this was considerably higher than the 90% finished Siemens
plant which was reported to have cost DM900 million. It was also noted that by using
existing facilities, principally structures which meet required licensing standards, the
cost could be lower. The large discrepancy between the German plant and the
Japanese estimate was left unresolved. No O & M estimates have been made.
Visit to JNF, December 3, 1993:
JNF is a major urania fuel fabricator in Japan. JNF is a principal investor in the
reprocessing facility and is expected to play a major role in designing, building and
operating the MOX factory.
The meeting was attended by M. Petski and T. Shigeto from GE. Principal
contacts at JNF were K. Murota, Chief Engineer, K Kumoro, Senior Staff Member and
T. Ishikawa, Section Manager, Business Dept. T. Ishikawa was identified as the
principal member of the Japan MOX planning team from JNF. Ishikawa also had spent
I0 years on assignment with the IAEA in Vienna.
JNF is located in Kurihama, approximately one hour south of Tokyo, on the
entrance to the Tokyo Bay. JNF is a joint venture operation of the General Electric Co
(40%), Hitachi (30%) and Toshiba (30%). The venture was first established in 1967
4.1 .I-7
and fuel fabrication started in 1970 under a technology agreement with GE. JNF
produces fuel to designs provided by the three companies (shareholders).
The plant capacity is 850 MT U02 per year. Present fuel production is 570 MT
U02, equivalent to 3000 bundles per year.
The operations are 75% automated from powder receiving through bundle
assembly. There are two fabrication facilities on site. Both are licensed to handle 5%
enriched fuel. The main plant has a capacity of 580 MT. The sub plant has a capacity of
270MT and is used for fabrication of the Gadolinia fuel.
JNF fabricates its own spacers, tie plates and other small components. UO2
powder is procured from GE in Wilmington NC, France, England and Sumitomo in
Japan. Zircoloy tubing is supplied through GE Wilmington, Sumitomo and Kobe in
Japan.
JNF has assigned several people to work on the Japan MOX planning team.
This team meets on a weekly basis to integrate plant and equipment design inputs.
Preliminary plant and equipment designs have been proposed and are under further
team study and refinement. Preliminary cost numbers for the facility are in the one
billion U.S. dollar range. There seems to be some feeling that this number may be off
because of the uncertainty of the cost of safeguards. The PNC Tokai facility is
considered by the IAEA as the model for safeguards standards. The team feels that
integrating these practices into a manufacturing facility would greatly increase cost
projections. These safeguard differences still need to be resolved. No one could
identify a completion date for these studies but an early 1994 completion date was
estimated.
MOX LEAD USE ASSEMBLY TESTING:
Toshiba presently has the lead in the fabrication of fuel assemblies for
insertion into a Toshiba designed reactor. Plans call for loading of MOX bundles
equivilant to 1/4 of core size. The MOX fuel for this program will be fabricated at BN
in Belgium and the Bundles assembled at FBFC in France and shipped to Japan. The
target reactor for the fuel assemblies is Reactor 3 at the First Fukushima site. JNF
will handle the administration of the fabrication contract, qualification of Belgium and
French facilities, and will supply all UO2 rods and bundle hardware. This activity
will take place over the next 3 to 5 years. JNF has also been directly involved with
Hitachi and Toshiba in MOX studies at BN in Belgium, COGEMA in France and
BNFL in England. From the manufacturing perspective, JNF has, over the past
4.1 .I-8
several years, participated with Hitach, Toshiba and others in MOX bundle transport
container design and fabrication, and equipment design related to automated rod
inspection ,bundle assembly and welding equipment for MOX fuel manufacture.
4.1 .I-9
4.1.2 BNFL Experience and Facilities for MOX Fuel Fabrication
4.1.2.1 Experience Base
A. Fuel Cycle Capabilities
BNFL and its predecessor, the AEA, have been designing and manufacturing fuel for nuclear
reactors, enriching uranium, transporting fuel, reprocessing spent fuel and managing the waste
products for over 40 years. BNFL was created out of the former Production Group of the
AEA in 1971 and became a public limited company in 1984. In 1990, BNFL Inc. was
formed as a wholly owned US subsidiary of BNF.
In the UK, the company's principal business is the provision of the complete cycle of nuclear
fuel services as the supplier to the UK nuclear utilities. In addition, the company has
extensive experience in the provision of fuel for many types of reactor systems including
Magnox and AGR, water reactors (PWR, BWR and SGHWR) and fast reactors (PFR). This
includes uranium metal fuel for the Magnox reactors and ceramic oxide fuel for the other
reactor types, including MOX for the fast reactors.
In the US, BNFL Inc. has grown rapidly with a number of design and engineering projects,
largely in the DOE waste management area, but also with a number of commercial clients. In
addition, BNFL Inc. has performed project management for the transport of nuclear materials
from US sites to the UK for BNF.
B. History of Involvement in MOX Fuel
BNFL and its predecessor, the AEA, firstbegan manufacturingMOX fuels in the early 1960s
when about 3 tonnes HM was produced for a wide variety of reactor systems including PWR,
BWR, and gas- cooled reactors. BNFL and AEA recognized the importance of fuel
homogeneity and indeed, all the fuel performed well in reactor, thus demonstrating the
feasibility of using plutonium fuels in LWRs.
From 1970 until 1988, BNFL and AEA produced over 18 tonnes HM of fast reactor MOX
fuel containing plutonium at enrichments of up to 33 %. A very close working relationship
with fuel cycle development has been maintained between the fast reactor fuel fabrication workin BNFL and in AEA.
In addition to the large scale manufacturing experience gainedto support the Fast Reactor
4.1.2-1
Program, AEA at Windscale also provided a service for the manufacture of experimental fuel
for the project. The experimental fuel fabrication facilities at Windscale were operated for
almost 20 years and during that time were used for fabricating a variety of different uranium-
plutonium containing fuels (oxides and carbides), a range of pellet sizes, various designs of
fuel pins, different cladding materials and irradiation rigs. Following manufacture, fuel was
transported from the Windscale/Sellafield site to the reactor facility. In addition, BNFL has
had experience of transporting plutonium in nitrate and oxide form to customers
internationally.
C. Plans to Provide Thermal MOX Fuel Capability
As a major reprocessor BNFL is committed to the effective utilization of its customers
reprocessing products. The provision of a MOX fuel supply capability is an essential part of
that strategy.
In March 1990, BNFL and AEA established a formal collaboration agreement in the Thermal
MOX Fuels business area. As a demonstration of commitment to providing a secure and
reliable thermal MOX fuel service to its customers, BNFL in collaboration with AEA Fuel
Services invested in the design and construction of a small-scale MOX fuel production facility
(MDF) which is approaching completion at Sellafield. The experience gained in the
experimental fuel fabrication facilities at Windscale is directly relevant to the technology and
production of modern thermal reactor MOX fuels and will be fully utilized in the fabrication of
MOX fuel in the MDF. In particular, the plant incorporates the BNFL short binderless route
process which produces a highly homogenous MOX powder fuel for fuel pellet production.
This facility will be capable of producing up to 8 tonnes HM per year as PWR MOX fuelassemblies from the end of 1993.
BNFL is further developing its interests in the MOX fuel market through the construction of a
commercial scale plant in the UK known as the Sellafield MOX Plant (SMP). The BNFL
board has recently approved the next stages in the design and construction of SMP which is to
be operational by the end of 1997. The plant will incorporate expertise from MDF and other
fuel fabrication facilities. The plant will be designed to be highly reliable with low
maintenance requirements and low in-line stocks supplemented by off-line secure storage
facilities. The operation, containment, shielding, and maintenance of equipment will be
designed to achieve stringent BNFL targets for radiation exposure levels. The plant capacity
will be around 120t HM/year based on market assessments for the requirements of MOX fuels
towards the end of the 1990s. To achieve the tight program time scales required to ensure the
4.1.2-2
completion of SMP by end 1997, BNFL has drawn together a multi-discipline Task Force
within their offices at Risley, Warrington. This Task Force currently comprises a substantial
Team with expertise drawn from a number of recently completed major projects at Sellafield,
(including input from MDF commissioning experience), and fuel fabrication expertise from
Springfields. At this time the project team is more than 200 strong and is expected to reach
approximately 300 by early 1994. The task force includes project-management, all main
design groups, safety assessors and project services. In addition specialist input from, for
example, shielding and criticality assessors is being extensively used. The SMP process is
now fully established and is supported by an extensive development program. Building layouts
are complete with design progressing for commencement of construction and procurement
programs early in 1994. The first issue of the associated safety case has recently been
submitted to the regulators.
It is the expertisewithin this task force linked with BNFL Inc.'s US personnel which has been
used in the compilation of this report.
4.1.2.2 Description of the Sellafield MOX Plant
To support initiatives for MOX fuel supply, BNFL is currently embarking on the construction of
a large scale Mixed Oxide Fuel Fabrication plant on the Sellafield Site. This will give increased
capacity beyond that available from the MOX Demonstration Facility (MDF) and enable return of
plutonia to customers as usable MOX thermal reactor fuel. This plant will be called the Sellafield
MOX Plant (SMP).
The purpose of the plant will be to receive feed powders (plutonia and urania), convert them to
MOX pellets and subsequently manufacture fuel rods and assemblies. The plant is being designed
to manufacture a nominal 120 tonne heavy metal per year of MOX fuel assemblies. SMP will be
capable of producing a wide range of products in the form of PWR and BWR rods and
assemblies.
The manufacture of fuel pellets will be achieved using a process developed by BNFL known as
the short binderless route. Unlike conventional binderless granulation techniques where the urania
and plutonia powders are ball-milled for a considerable period to give the required properties, the
short binderless route utilizes a high energy attrition mill. This permits short cycle times of about
40 minutes compared with 4-8 hours for a conventional mill but still ensures that 100% of the
material is blended. Upon production of a homogeneous powder mixture, the conventional pre-
compaction and granulation has been replaced by a new process which utilizes a spheroidiser.
4.1.2-3
This agglomerates the powder in the presence of small quantities of die lubricant to give free-
flowing granules for the press feed.
The short binderless route offers a number of advantages:
a) Homogeneous pellet structure.
b) Good dissolution properties.
e) Good fission gas retention
d) Small hold-up in plant.
e) Fully contained process
f) Simple process which is easy to maintain and operate.
The urania feed material will be produced by the BNFL owned Integrated Dry Route (IDR)
process at BNFL's Springfields site. Over 14,000 tonnes of urania powder have been made by
this process which has been licensed in the US and France. The mechanical, physical and nuclear
properties of IDR material control to a large extent the properties of the MOX fuel when mixed
with typically 5-8% of plutonia. Following pellet production, the fuel manufacturing route
follows conventional practice. The majority of processes within SMP are fully automated to
reduce the dose uptake to the workforce.
The plant will incorporate campaign operations for producing pellets of a single enrichment for
loading into fuel cans. Campaign sizes will vary and could range from 0.5 to 10 tonnes. The
process plant is being designed so as to minimize loss of throughput due to enrichment and
campaign changes.
4.1.2.3 Process Description
The overall SMP process flow sheet is given in Figure 4.2.1-1. To support the
effective management of the design and subsequent installation of Mechanical,
Electrical and Instrumentation equipment, SMP has been divided into ten plant
areas.. Areas 100-600 correspond to the main stages in the process, and 700-
900 cover the service functions. Area 000 covers the general equipment within
the building.
4.1.2-4
These plant areas, as currently defined, are as follows:
Area 000-General Building Equipment
Area 100 -Powder Receipt
Area 200 -Powder Processing
Area 300-PelletingArea 400-Rod Fabrication
Area 500-Fuel Assembly
Area 600 -Sampling/Effluents and Contaminated ResiduesArea 700-Ventilation
Area 800 -Control, Electrical and Instrumentation (CE&I) SystemsArea 900-Services
Powder Receipt - Planl; ArQa 100
The following powder feeds are supplied to the SMP plant:
Plutonia (THORP, MAGNOX COGEMA)
Urania granules.
Urania granules combined with Zinc stearate lubricant.
CONPOR, this is a proprietary material used to control the porosity of the
sintered pellets.
Zinc stearate which is required as a lubricant in the spheroidisers.
The above powders are supplied in containers and are placed in an appropriate
storage on receipt. From storage, each powder container is transferred into a
glove box, weighed and then opened. The powders are transferred,
pneumatically, to a feed hopper when required for powder production.
Powder Processing - Plant Area 200
Powder production comprises 3 major stages. These are:
a) Milling
b) Blending
c) Spheridising
4.1.2-5
I'
The powders (plutonia and urania) are weighed, in the correct proportions, and
transferred into the feed hopper. From here they are fed, under gravity into an
attrition mill to enable the milling of the powder to take place.
The CONPOR powder is added to the process to control the porosity of the
sintered pellets. Zinc stearate powder is added to the homogenization process to
act as a lubricant in the attrition mill, spheridiser and blender.
The milled batch is fed, under gravity, to a blender. Three mill batches are
required for each blender operation. From the blender the blended material is fed,
via a screw conveyor, to a further attrition mill and then to a spheroidiser whe,_ it
is granulated. The granulated material is fed via a screw feeder to the press
hopper.
The common powder feed system supplies two parallel process towers from
milling to pellet pressing inclusive.
Pelletiing - Plant Area 300
Pellet Pressing
Granulated material is fed from the press feed hopper into the press where it is
pressed. Pellets produced are inspected, loaded onto boats and transferred to the
sintering furnace.
Sintering Furnace
The pellet boats are fed into one of four sintering furnaces via gas locks. The
boats are pushed through the furnace at a constant rate. Within the furnace the
boats pass through a series of zones. These are:
a) reducing zone
b) sintering zone
c) cooling zone
The sintered pellets are removed from the furnace still on the boats, via a gas lock
and are transferred to the grinding process glove box.
4.1.2-6
Pellet Grinding
The pellets are unloaded from the boats, passed through one or two grinders
where they are ground to size. The pellets are then passed through a series of
brushes where residual dust particles are removed. The pellets are checked for
size, inspected for surface defects and then loaded onto trays. Pellets may be
removed during inspection for sample analysis if required. Sampling systems will
be employed with use being made of the existing THORP sample transport
systems. The pellet trays are conveyed to a sintered pellet store.
Rod Fabrication - Plant Are_ 400
Sintered Pellet Store and Stacking Systems
The trays are weighed into storage and placed on shelves awaiting results of
analysis. The trays are weighed out of storage. The pellets are removed from
these trays where they are formed into stacks on stack trays containing thecorrect enrichments for the fuel rods.
Rod Filling, Welding and Inspection
At one of the two rod fill and weld stations, empty rods are purged with helium
prior to the pellets being inserted into the rods from the stacking system. The
filled rods are moved to a welding station where a cap is TIG welded onto the end.
The fuel rod is pressurized via a drogue hole which is subsequently welded. The
fuel rod is then passed through a series of inspection stages:
a) leak test
b) mass spectrometer
c) X-Ray inspe,:tion
d) rod scanner
e) identification/contamination inspection
f) geometry inspection
g surface texture inspection
h) straightness inspection
4.1,2-7
Rod Transport
Finishedfuel rods are insertedinto magazinesprior to manufactureof the final fuel assembly.
Fuel rods are received from the rod inspectionstages, identified,aligned in the correct positionand inserted into a magazine,positioned at the insertion machine. The orientation within the
loaded magazinereflects the requiredorientationof the fuel pin arrayin the final fuel assembly.Magazines arestoredin a dedicatedmagazinestore.
FueJAssembly-Plant Area 500
AssemblyProductionand Inspection
The loaded magazines are transported to one of the two fuel assembly lines. Skeletons are
constructed from components on a bench. (Skeletons may also be supplied to SMP pre-constructed.) The magazines areasalignedwiththe skeletonandthe fuel rods are loadedinto theskeleton.
Whenthe fuel rods havebeen installed into the skeleton,bottom andtop nozzles are fitted where
appropriate. Thecompleted fuel assemblyis inspectedinitiallyfor:
a) geometry
b) grid location
c) rod to rod gaps
The assembly is subsequently cleaned, if required.
The final inspection of the fuel assemblycomprises of:
a) cleanliness
b) damage scratches
c) grid and spring condition
d) weight check
The inspected fuel assembly is stored until it is required for despatch in a suitable transportcontainer.
4.1.2-8
PCM Waste Handling- PlantArea600
During normal SMP operations, solid wastes will arise principally in the form of plutonium
contaminated material _CM). These arising can be split into two main streams:
a) process feed waste (mainlyempty plutonia cans)
b) maintenancewaste
The process has been designedto minimizeas far as practical the generation of secondaryPCM.
The PCM wastes are exported inside drums to the site waste treatmentcomplex via the PCM
marshallingroom withinthe THORPcomplex.
Ventilation - PlantArea700
The main functionsof the ventilationsystem are:
a) To assist in providingactivitycontainment.
b) To provide a satisfactoryworkingenvironmentfor personneland equipment.
c) Remove airborneactivity fromdischargeairto ensureemissionsareacceptable.
The overallventilationsystem will comprise a numberof separatesystems designed as a cascade
to ensureairflows from areas of lower contaminationclassificationto areas of potentially highercontamination.
The system for contaminationclassificationis broadlyas follows:
a) C4/C5 Zones: Normallyunmannedareasincludingglove boxes.
b) C3 Zones: Occasionallymanned areassuch as process cells containing gloveboxes and maintenance areas.
c) C2 Zones: Normally mannedareas such as access corridors,change rooms,
operating working areas.
d) C1Zones: Inactive areas includingoffices and change rooms.
The main components of the ventilation system are:
a) C5 Glove box Extract. The glove box system will be at the greatest depression and air
will flow into the glove box from the surrounding C3 cell areas. Each glove box extract will
be filtered and connected to a combined C5 extract duct. This will be filtered and monitored
4.1.2-9
prior to dischargevia the THORPStack.
b) C3 Cell Ventilation,Air will cascade into the C3 cell areasfrom the C2 operatingareas.C3 air extractedfrom the cells (i.e. that which does not cascade into the Glove box C5
System)will be combined in a C3 extract duct. This will be filteredand monitoredprior todischargevia the THORPstack.
c) C2 Operating Area Ventilation.SMP will have a dedicated C2 air supply and extract
systemwhichwill providefiltrationand monitoringof extractedairpriorto dischargeattheTHORProof level.
d) C1 AreaVentilation. C1 areaswill be servedby commercialtypeventilation system.
..ControlandInstrumentation-PlantArea 800
Location of Control
In line with the overallphilosophy to minimize operator dose uptake the majority of the process
plant and services will be controlled, monitored and surveyed remotely from a central control
room. This will be located in the SMP process building. Some fuel assemblyoperations will be
controlledlocally using control stations dedicatedto each operation. The main control functionscarriedout fromthe SMP control roomwillbe:
a) plant productioncontrol
b) fissilematerial accounting
c) monitoring and display of process parametersand alarms
The plant environmentalmonitoring systems (e.g. activity-in-air,ventilation and service supplies)will be controlledand monitoredfromwithinthe mainTHORPcentral control room.
4,1.2-10
Typeof Control
IntegratedAutomation System (IAS) consisting of ProgrammableElectronic Systems (PES) will
be used to carryout the operationalmonitoringandcontrol function of the process. The IAS will
conform to relevant BNFL and InternationalStandardswith respect to integrity levels and willprovidethe following control features:
a) Startup, normalrunningandshutdownof operations
b) Sequence control ofbatch operations
c) Stop and/orreset of individualoperations
d) Authorization of releaseof plant equipment to maintenancemode
The IAS does not containanycomponentsof the engineeredsafety protectionsystems which are
necessary to support the safety case. These are providedin accordance with the requirementsof
the appropriatecompany standardsand utilizefully independentcomponents from the IAS. The
status of systems defined as protection systems as well as the status of plant parameters and
alarmsdesignated as safety relatedare displayedin the SMP controlroom.
4.1.2-11
MassFlowrateAppmxlmately600 kg/day
i I MILL' 50kg i_ MILL0
!
150kg I BLENDER i BLENDER
50kg ' MILL t MILL ORfI!
,,i
, f• 50kg ti PRESS PRESS _Samplest .
I " jlf ....._- 1 _' I _
lt .....I_....i' .... I, I
!GRINDER I GRINDER _.._day
; TRAY(PELLET)STORE
fROD LOADING ROD LOADING
I
L
Figure 4.1.2:1' SMP'Process Flow Diagram
4.1.2-12
4.1.3 COMPARISON OF U.S. AND FOREIGN (UK) MOXFUEL FABRICATION FACILITY REGULATORY
REQUIREMENTS
Although DOE Orders and NRC regulations for handling and processing Special Nuclear
Materials (SNM) exist, as pointed out in Phase 1B studies, there has been no regulatory activity
since mid 1970s that specifically concerns the requirements for constructing and operating a MOX
fuel fabrication facility. In this section, a comparison has been made of the existing U.S.
regulations with the those applied in UK for the Sellafield Mixed Oxide fabrication Plant (SMP), to
identify the key differences between the applicable requirements and their impact on cost, schedule
and development requirements.
The following comparison is based on construettng the current design of SM.Pwithin the U.S. Thecomparisonhas been conducted b)' reviewing the SIvIPdesign criteria or current BblFL standardsagairt,_tthe equivalent US Standards and regulations. Where key differences have b_n highlighted the ',ff'feetofth,_c on cost. program and development requirements has been identified. Inorder to simpli_,"the reviewthe comparison was carried out against the following meas:
a) Safety and Environmental Standardsb) Safeguards Requirementsc) Security Provisionsd) Regulatory Approvals
To assist the reader information has also been provided on the following:
a) Program tim,'tales for the construction of SMP within the UK,b) lnl'rastructurc requirements currently available on BNFL's Sellafield _ite which will be utilized by
the SMP.
c) Current identified development rcquiremet_ for SMP.
4.1.3-1
4.1.3-1 SAFETY & ENVIRONMENTAL
This safety and environmental comparison was carriedout by reviewing the SMP radiological safetyrequirementsand criteria(see Appendix.23)against the comparableUS regulatory requirements. It isrecognized that within the US each plant will have internaloperationallimits. These are more stringentthanthe regulationsin orderto avoidapproachingthe overallregulatorylimits. However. in the absenceof any information on exact location of the plant a review against the operational criteria was notpossible.
I, COMPARISON
The following section is a comparisonof the criteria applicable to SMP (as described in Appendix A)against the comparableUS regulations.
I.A. Occupational Limits
The principle dose target for BNFL Plants is that dose uptake to individuals should not exceed 15mSvy"*with the group average < 5mSvyj. The US Regulatory requirement, translated into a quarterly limitof 12.5 mSv is less restrictive than the current BNFL design target in terms of occupational exposure.
BNFL's _ practice is to operate plants with monthly dose uptake "limits" of .3mSv (whichcompares favorably with the US Regulatory limit).
The Engineering Design Principles which are used by BNFL to ensure compliance with the ALARPconcept should also be adequate to demonstrate the US ALARA concept which is largely similar.
I.B. Accident Conditions
For large accidents the US Regulations limit the public dose at the site boundary to 250roSy. The USRegulators would expect such accidents to have a frequency of occurrence < 10_y_. The BNFLapproach to accidents arising from internally initiated events with an equivalent consequence would beto limit the event frequencyto < 10_y_. The BNFL requirement is therefore more restrictive in thisrespect.
The BNFL criteriaforaccidentsaffectingthepublicalsolimitthefrequencyofsmalleraccidents(<250mSvatsiteboundary),e.g.,effectivedosesbetween0.OlmSvandImSv arelimitedto0.Oly* foraerialdischarges.Similarlimitationsareplacedoneventsleadingtoaccidentalliquideffluentdischarges.TheBNFL requirementis,inthisrespectmore restrictive,byrequiringtheminimizationofaccidentalaerial/liquiddischargesfromotherthanlargeaccidents.TheBNFL designcould,therefore,includemoreprotectivesystemsthanwouldtheUS equivalent,assumingthattherearenofurtherUS Regulations.
The inclusionofa timeaverageddosecriterionintheBNFL designrequirementsistosimplifythedemonstrationthatthesummedfatalrisktoa member ofthepublicfroma siteis< 10C'y_. No
equivalentmortalityrateisgivenfora US site.
The BNFL designrequirementsareclearlymore restrictivethantheUS Regulationsforcontrolofaccidentaldoseuptaketooperators.The annualprobabilityofexceedanceoftheUK StatutorylimitisallocatedatargetofI03 whileeventsleadingtodoses> ISvareallocatedanannualprobabilityofI0_.There is no equivalent in the US Regulation. In addition to these requirements, smaller accidents whichlead to airborne contamination/abnormal dose rates sufficient to cause evacuation are restricted to
..
4.1.3-2
< I0_:'j. The US Regulations stipulate that workforce doses from accidents on a plant will beaccommodated within occupational limits (50mSvy_) for other than large accidents(where the workforceare considered 'disposable'), In BNFL terms, any accident which would reasonablybe expected to occurwithin the lifetime of a plant (say once in 50 years) would be included within the predictive dose uptakeassessment (occupational dose) for the plant, taking into account the potential consequenceand predictedfrequency. However application of the BNFL Engineering Design Safety Principles which wherepracticableare applied to all new plants (althoughnot mandatory) will ensurethat the consequences fromaccidents occurringwith this frequency would be very low.
BNFL plants are also designed to ensure that the potential for criticality events to occur is lessthan 10_ _. The building fabric will be such that the 100 mSv contour is within the building itself.
I.C. External Event
Seismic protection on BNFL plants is determined accordingto the design basis earthquake which, in thecase of SMP is the event with a return period of 10_years (ie once in 10,000 yrs). In the event of a DBEoccurring, seismic qualification of plant and equipment will be designed to ensure that the maximumpossible dose to an off-site person will be less than 5mSv. Plant which do not have the potential to resultin an unmitigated consequence of > 5mSv but > lmSv will be designed to withstand a seismic event witha probability of exceedance of 10_. All other seismic protection is made against ALARP considerations.This is considered equivalent to a design against the US "Safe Shutdown Earthquake" although it shouldbe determined how the UK and US 104yreturn period seismic events compare.
An operating basis earthquake (O.B.E.) is defined as that which would reasonably be expected to occurin the lifetime of the plant (once in 50 years for SMP). No safety related plant, system or structurewould be impaired by the repeated occurrence of ground motions at the OBE level. The plant should beshut down safely and brought back on-line when it is shown to be safe to do so. The US Regulationsuse a 3x103yt event to define an OBE. This is more restrictive at face value but comparison should bedrawn against the UK equivalent.
I.D. Discharges
BNFL plants are designed to minimize effluents (aerial/liquid/solid) in accordance with the ALARPprinciple. Additionally the design must be demonstrated to meet the plant allocation of the overall siteaerial and liquid effluent discharge authorizations.
I.E. Summary'
The BNFL system of plant design differs from the US approach of design-evaluate-fix in that safetyassurance is demonstrated against all stages of the design by use of the appropriate standards and safetyevaluation at conception, definition, build, commissioning and operate stages. The final design isvalidated against the Company criteria.
The additional BNFL criteria particularly relating to low consequence events may incrPxtsethe extent ofsafety protection systems required for the plant.
4.1.3-3
I.F. Cost Effects
In general the equipment requirement associated with safety and environmental protection would besimilar for both the US and UK locations. Certain differences may occur in the engineering of therequirementswhich could lead to cost differences. These areas are listed below.
Control System. Within the UK the regulatory bodies will not allow credit .to betaken for the use ofProgrammableElectronic Systems (PES) in the design of protectivesafety systems. This has resultedina dedicated hardwiredprotectivecontrol system being incorporatedinto SMP which is additionalto thenormaloperationcontrol system. This hardwiredsystemcomprises approximately250 circuitswhich inthe mainare duplicatedwithinthe normal operationalcontrolsystem. If the US regulatorswere to allowthe use of PES in the design of protective systems then cost reductions could be made in this area. Inadditioncertain BNFL criteria relating to lower consequenceevents necessitates the need for additionalprotective equipment. As these are not requiredby the US regulations cost savings could also be madehere also.
Extreme Weather/Seismic. The design criteria for extreme weather and seismic events are based oninformation relatingto the areasurroundingthe SellafieldSite. These may differ significantlyfrom areaswithin the US (either more or less severe). As such cost effects could also vary as these criteriaare usedin the design of the overall building structureand also the internalplant and equipment. In the absenceof any details on plant location within the US no furtherinformationcan be provided.
4.1.3-4
4.1.3-2 , SAFEGUARDS
BNFL has worked with the safeguards regulatory authorities (IAEA and Euratom) for many years.forming a good workingrelationshipanda full understandingof the objectives and requirementsof eachorganization. BNFL's experience in safeguardsdates back to the mid-seventies and the Company hasbeen a leading participantin major internationalsafeguardsprojectsandvariousspecialized advisory andconsultants' groups [e.g. SAGSI (Standing Advisory Group on SafeguardsImplementation),LASCAR(Large Scale Reprocessingproject)and is representedon the Steering Groupof ESARDA (the EuropeanSafeguards Researchand Development Association)].
The Company has since 1977. operated a policy of including safeguards in the design criteria of newplants. BNFL has many years experience of designing plants to take account of these safeguards andmaterials accountancy obligations. This experience has been brought to bear on the design provisionsfor materials a_':countancyand safeguards provisions in the Sellafield MOX plant.
SMP has been designed to satisfy the IAEA criteria which state that it must be possible to detect aprotracted loss or gain of 8 kg plutonium over oae years operation (in process areas) or an abrupt lossof 8 kg plutonium within one month. These criteria apply to bulk materials. In the case of discrete items.the inspectorates must be able to detect any item loss or gain.
The safeguardsapproach for SMP has been designed to be of the highest standardand to meet all national(UK) and international (Euratom/IAEA) requirements. SMP may be designated for inspection by theIAEA. The IAEA approach to automated MOX fabrication plants favors the use of Near Real TimeMaterials Accountancy(NRTMA), in-line Non-DestructiveAssay (NDA), and advanced containmentandsurveillance employing surveillance, safeguardsseals, and weighing. The currentdesign provisions thatare being incorporatedinto SMP will ensure that the plant satisfies all the requirementsfor internationalsafeguards.
Cost Effects. Due to the close similarity of the US and UK requirements for safeguards no costimplications are envisaged in this area.
4.1.3-5
4.1.3-3 SECURITY
In the UK minimum standards are laid down for the physical protection of nuclear material which fullymeet the recommendations of the InternationalAtomic Energy Agency (IAEA). These are published asINFCIRC/225/Rev.3 under the title "The Physical Protection of Nuclear Material" and the mandatoryrequirementsof the "Convention on the Physical Protection of Nuclear Material" (INFCIRC/274/Rev. 1)which the United Kingdom has ratified.
In accordance with IAEA recommendations, nuclear materials are categorized according to theirsensitivity which depends upon the type and quantity of material see Table re.L3-1 The category intowhich material is placed is the criterion for the minimum standardof physical protection it is accorded.
As SMP contains Category 1 quantities of nuclear material the physical security provisions are designedaccordingly. The SMP security provisions are detailed within a Security specification. This has beendeveloped in conjunction with the BNFL security advisors to ensure the design of SMP complies withthe appropriate regulations. This document is of a confidential nature and cannot be reproduced here.However, listed below are the type of provisions provided for plants at the Sellafield site. (Note SMPfalls into the inner area criteria_.
Overall, protection is achieved at sites which contain nuclear material by a perimeter fence. This ispatrolled by the United Kingdom Atomic Energy Authority Constabulary (UKAEAC), with access to thesite controlled by a pass system for personnel and a gate or barrier arrangement for vehicles.
To implement a "defense in depth" principle and to concentrate defensive measures where they are mosteffective, three types of areas have been defined for sites holding nuclear material. Progressively throughthese three areas more stringent steps are taken in terms of physical protection measures and in therestriction of access to those who really require it and whose trustworthiness has been established. Thethree types of protected areas may be defined as:
a) Inner Area - created where Category I quantities of nuclear material as defined in INFCIRC225Rev.2 are used or stored and where therefore the consequences of theft or dispersal of the materialare such that extremely rigorous restrictions are required.
b) Intermediate Area - created where Category II and some Category III quantities of material are usedor stored and where although the consequences of theft or likelihood of dispersal of the material arenot so serious as for Category I quantities, rigorous restrictions are nevertheless required.
c) Outer Area - all other parts of the site not specifically identified as Inner or Intermediate area.Protected by the site perimeter and control of entry system.
In addition, Vital Areas are identified and established. These are areas identified by safety specialists ascontaining equipment systems or devices which are, alone or in combination, vul,erable to sabotage, theeffects of which would be sufficient to cause a radiological hazard to the public. Access to Vital Areasis limited and controlled.
The prime physical protection features of Inner Area protection include:
a) a controlled access point with a unique identification system.
4.1.3-6
b) a physical barrier around the building, patrols by armed police within radio communication to thecentral control post, intruder detection systems both external and internal, CCTV, and effectivelighting.
c) nuclear material monitoring equipment.
d) specific standards of construction for stores or rooms where material is held.
The prime physical protection features of Intermediate Areas include:
a) a fence or structure of a building delineating the area so that entry and exit are effectivelycontrolled.
b) a checking system at entry points to allow entry only to authorized persons.c) the provision of an intruder detection system and appropriate area lighting, supported by police
patrols and response arrangements.
Using the above measures as minimum standards the physical protection system for each site is designedspecifically for that site taking into account such factors as facility design, geographical location, and theform of the material being handled.
Many of the safety features incorporated in the design of plants, make sabotage difficult. The nature of
plants and processes adopted in the nuclear fuel cycle mean that most sensitive nuclear materials are ina form that makes them highly unattractive and the fact that they need to be heavily contained alsominimizes the consequences of sabotage.
Comparison of the above with US requirements is difficult due to the confidential nature of the detailedinformation. As the US and the UK standards are both based on the IAEA regulations the view is that
the provisions currently incorporated within SMP would be similar to those required by the NRC if theplant were to be constructed within the U.S.
On this basis it is felt that the security provisions would have minimal effect on cost and program if aplant similar to SMP were to be constructed within the US.
Cost Effects. Due to the close similarity of the US and UK regulations on security no cost implicationsare envisaged in this area.
4.1.3-7
TABLE 4.3.1-1
CATEGORIZATION OF NUCLEAR MATERIAL
.i
I 11 Ill(c)ii m
1. Plutonium (a) Unirradiated(b) 2kg or more Less than 2kg but 500g or less butmore than 500g more than 15g
2. Uranium-235 Unirradiated(b)
uranium enriched to 5 kg or more Less than 5kg but lkg or less but120%235U or more more than lkg more than 15kg
uranium enriched to 10kg or more Less than 10kg10% 235U but less than but more than20% lkg
uranium enriched above 10kg or morenatural, but less than10% 235U
i i ii
3. Uranium-233 Unirradiated (b) 2kg or more Less than 2kg but 500g or less butmore than 500g more than 15g
"41Irradiated Depleted ornatural uranium,thorium, or low-enriched fuel(less than 10%fissile content)
!(d)(e), ,,
(a) All plutonium except that with isotopic concentration exceeding 80% in plutonium-238.
(b) Material not irradiated in a reactor or material irradiated in a reactor but with a radiation level equalto or less than 100 rads/hour at one meter unshielded.
(c) Quantities not falling in Category III and natural uranium should be protected in accordance withprudent management practice.
(d) Although this level of protection is recommended, it would be open to States, upon evaluation ofthe specific circumstances, to assign a different category of physical protection.
(e) Other fuel which by virtue of its original fissile material content is classified as Category I and IIbefore irradiation may be reduced one category level while the radiation level from the fuel exceeds100 rads/hour at one meter unshielded.
it,
4.1.3-8
4.1.3-4 REGULATORY REQUIREMENTS
This section outlines the regulatotVprocesses requiredfor the constructionof nuclear facilities in the UKand the US.
Table 4.4-1 lists areas of regulationand the agencies within the US and UK who have responsibilities ofregulation in each area. All UK regulatory authoritiesare statutory bodies just as in the US. None are
private concerns and they regulate both civilian and other government agencies. In the followingdiscussion it is assumed that the readeris familiarwith US regulatory institutions and practices.
Just as in the US, the British regulatorsare broadly split into those concerned with safety and thoseconcerned with impact on the environment.
I. SAFETY
I.A. Nuclear Installations lnspectorate (Nil)
For safety, the Nil, in an exact parallel to the US NRC, has jurisdiction over all nuclear safety issuesconcerned with the operation of a nuclear facility from initial design and site selection todecommissioning. It is the Nil which grants and monitors compliance with the site licenses for allnuclear sites.
The responsibilities of the Nil. as incorporatedin a site license, include, but are not limited to:
Arrangements for emergency situations (eg. a majorrelease of radioactivity)Consignment of nuclear matter (no materialcan be c¢,nsignedto any place other than a relevant sitewithout Nil permission)Recording and investigating site incidentsAppointment of Duly Authorized PersonsNuclear Safety CommitteesPlants - plans, designs and specificationsOperating rules and Operational Safety AssessmentsSafety mechanismsPeriodic shutdowns
Decontamination and decommissioning
Before any major tasks can be undertaken it is necessary to involve the Inspectorate and no work onfacilities can commence without written authority.
The Nil has the authority to close down the operations of a nuclear site or facility should the operatorbe 'out-of-compliance', or to prevent the restart of a facility after shut-down or modification if they arenot satisfied with the work undertaken either in quality or completeness. Furthermore, the Nil has theauthority to conduct regular inspections or complete audits of site operations for compliance with licenserequirements. In all these respects the Nil is entirely similar to the US NRC.
However, in the US, since the US NRC has no jurisdiction over the US DOE nuclear facilities, or USDOD facilities, the regulatory picture is different for these facilities and contaminated sites. Thisdifference of safety standards between civilian and government sites does not exist in the UK.
4.1.3-9
I.B. Her Majesty's lnspectorate of Factories
While the Nil is responsible for nuclear safety, compliance with conventional industrial safety legalregulations rests with Her Majesty's lnspectorate of Factories although in certain instances the Nil willtake on the conventional safety role on behalf of the Inspectorate as a matterof convenience.
Her Majesty's Inspectorate is responsible for industrialcompliance for safetywithconventional industrialequipment in a directly equivalent manner to the Occupational Safety and Health Administration (OSHA).Likewise the Inspectorate has the responsibility for monitoring operator compliance in the routine testingof safety equipment and for the investigation of major safety incidents, industrial inquiries, etc.
Like the Nil, Her Majesty's lnspectorateof Factories has powers of enforcementincludingthe authorityto close down operations, or place prohibition notices on plants or equipment if they believe there is athreat to safety. This extends to prevention of equipment start-up if the agency is not satisfied with thework undertaken.
II. ENVIRONMENTAL IMPACT
II.A. Her Majesty's Inspectorate of Pollution (ItMIP)Ministry of Agriculture, Fisheries and Food (MAFF)
One significant difference between the US and British regulatory, process is that in the US differentagencies (the NRC and the EPA) are responsible, respectively, for civilian radioactive, and for non-radioactive (Hazardous) emission and wastes, whereas in Britain the same two regulatory agencies (HMIPand MAFF) working together have responsibility for both typgs of material. In both countries localauthorities also have an involvement.
Despite its overall federal responsibility for radioaaive emissions, the US NRC has no jurisdiction overthe US DOE nuclear facilities, so that the environmental compliance picture is different for these facilitiesand contaminated sites from similar civilian nuclear facilities, just as it is different for safety issues. TheEPA now has provided an element of commonality by providing environmental emission standards forUS DOE compliance and at the same time assisting the US NRC in its emission standards. Furthermore,in the US, the States have delegated responsibility for certain pollution controls and regulations. Thiscompetition between agencies does not exist in the UK.
Radioactive discharges are controlled, in the UK, under legislation administered by Her Majesty'sInspectorate of Pollution (HMIP) and the Ministry of Agriculture, Fisheries, and Food (MAFF). Theseregulators are jointly to limit gaseous or liquid discharges of both nuclear and non-nuclear hazardousmaterials through implementation of discharge authorizations which specify maximum authorized limitsand other conditions and requirements to be met by the operator. For example, this could be theestablishment of a minimum specified environmental monitoring program to assess the impact ofdischarges on the environment or on 'critical groups'. The limits, conditions and requirements arereviewed and reauthorized on a regular basis of one to three years. Breaches of the authorizations arean offense under the law and can result in prosecutions and substantial fines imposed on the operator.
III. USE AND DEVELOPMENT OF LAND
In addition to regulating nuclear and conventional safety, and the impact of operations on theenvironment, the UK also regulates the use and development of land. This is achieved through the
4.1.3-10
applicationofplanninglawsfortheconstructionofnuclearand non-nuclearfacilitiesina complexprocess.
Initially, this control is exercised through the Local Authority. When construction work is major, e.g.,for a nuclear facility, the planning approval sought of a Local Authority is generally referred directly tothe Department of the Environment who will either decide on the application or refer it to a 'PublicEnquiry'. The Public Enquiry allows the case for, and the objections against, the trajor constructionactivity to be heard. Such enquiries are chaired by a Government appointed inspector, who makes thefinal recommendation to the Secretary of State. In a number of cases such enquiries have lasted up totwo years (eg. for a nuclear power plant at Sizewell). The Public Enquiry process has its US equivalentin the Licensing Board hearings during the construction permit process for nuclear plants.
IV. STEPS IN THE REGULATORY PROCESS
Within the US the process of building a nuclear facility is different if the facility is a civilian one subjectto US NRC regulations than if it is a US DOE facility subject to DOE Orders.
On the civilian side the process requires three principal steps:
Approval of a site supported by an Environmental Assessment (approval being signified by thepublication of a US NRC Environmental Impact Statement)Approval of construction supported by a Preliminary Safety Analysis Report by NRC
. Approval of operation supported by a Final Safety Analysis Report by NRC
In practice the latter approval steps may be granted only in part as the NRC regulators become assuredof the safety of the facility and the hurdles of public hearing are overcome.
On the US DOE side the process is similar but different since DOE itself regulates its own safe operation.The process also requires three steps:
Based on a selected site DOE provides an Environmental Impact Statement for the facility whichin recent years has to comply with EPA regulatory emission control standards.
Construction is supported by a Preliminary Safety Analysis Report approved by a different sectionof US DOE, and
Operation supported by a Final Safety Analysis Report again supported by a regulatory section ofthe US DOE.
Because some nuclear facilities are of relatively low hazard (especially those handling low andintermediate waste) a Hazard Classification is performed at the PSAR state tOassist in a graded approachin which some facilities would be subject to full scrutiny and others would not. In a crude sense this haselements of a risk approach to regulation. The US NRC is not involved in these DOE approval steps.
Within the UK the process of building a nuclear facility is subject to one more formal step than in theUS. The steps are:
Approval of land-use through application to the local authority which would be referred to theDepartment of the Environment. This is supported by an Environmental Report containing anassessment of potential hazards to both the population and the environment, as well as the aesthetics
4.1.3-11
of the building, effect on local economy etc. equivalent to the Environmental Impact Assessment,It is at this stage that a possible Public Enquirs,'may be held.
Approval of construction supported by a Precommencement SafeD' Report (PCSR) by the NuclearInstallations lnspectorate (Nil). Risk-based methods are employed from the conceptual designonwards.
Approval of inactive and active commissioning supported by the Pre-Commissioning Safety Reportby Nil.
Approval of operation supported b.vthe Pre-OperationalSafeD' Repor_or the Plant Safety Case(PSC), equivalent to the FSAR. by Nil.
The third of these steps is additional to the formal civilian process in the US although in practice itgenerally occurs as final approval may be granted in stages through commissioning and low powertesting. Since Nil is an independent regulator)' authorit,v the formal process of regulation for newgovernment nuclear facilities is ver), parallel but a little stricter in the UK.
V. CONCLUSION
Whilst the regulator)' bodies which exist in the US and UK are different the overall requirements aresimilar. As such no cost or schedule implications are envisaged due to the regulator approval processes.
4.1.3-12
TABLE 4.4-1
REGULATORY BODIES IN THE US AND UK
Area of Regulation .... [ US Regulatory Authority [.... UK Regulatory AuthoritySafety ' ' ' .....
Licensing of nuclear sites' and " 'Nuclear Regulatory Commission Nuclear installationsfacilities (NRC) Inspectorate (Nil)
Agreement State Authorities Her Majesty's Inspectorate ofFactories
. ii
Environmental Impact
Control of emissions and Nuclear Regulatory Commission Her Majesty's Inspectorate ofreleases of radioactive pollutants (NRC) Pollution (HMIP)
Agreement State Authorities Ministry of Agriculture, "Fisheries, and Food (MAFF)
National Rivers Authority
Control of emissions and Environmental Protection Her Majesty's Inspectorate ofreleases of non-radioactive Agency (EPA) Pollution (HMIP)(hazardous) pollutants
Ministry of Agriculture,Fisheries, and Food (MAFF)
State Authorities Local Authorities
National Rivers Authority
Use of Land
Approval of construction and Nuclear Regulatory Commission Department of the Environmentclosures of nuclear and non- (NRC) for nuclear facilitiesnuclear sites
Local and State Authorities forothers
4.1.3-13
4.1.3-5 SMP OUTLINE PROJECT SCHEDULE
i : i
1992 _ 1993 _ 1994 1995 i 1996 1997- .......................................... _ ..........
Q3 Q41Q1 t212 I213 Q41Q1 Q2 Q3 Q4 Q1 Q2 Q3 Q41Q1 Q2 Q3 I_}Q1 Q2 Q3 Q4i :
........ • .......................... -4 -..4-- ...........
OPERA_
: (SEE NO'I1E)ADVANCED
FUNDSMAIN B&C"CONST _ _TIONV V"P Vc=,,_._ V_
AGREE TO COMMENCE HANIX)VER FORCONSTRUCTION V V AGREE TOCOMMENCE _7.E_..._,.[_.,,o. co=,.ss_,_.
i 15,StlEPCBR ISSUEI_C
"_ E.qIIMATETO\/ _l \/ _7 ISSIJEI:ISC.41,_IVEV Vm AND APPROVAl S
i,,,,.=J_
[ IF51(;NMECH i PRFOR MAJOREQUIP
CONTRACTS: CONTROLSYS DELIVERAB.ES
_,._,_.V V )V V_,_,_ V V ,.,_=,,--_,._,_PROCUREMENT .......... _. i
CCUaENOE.et.t[_,=:;_ _ m_ i :CONSTRUCllON STRUCI'UREV VtlEAlrHEIIlrlGNT ) :"
V MElUINST/I .LkTION
IIISTAI [ A'IION ........................
_ FCmACTWEt-7PREHANDOVER TESTING _ VAND COMMISSIONING
........................
N.Io SIHne hmlassembles _1 Im available I_iOeIo Ihisdale airing from Ildive colnmissiomingaclktili_.
ii
i .
4.1.3-6 INFRASTRUC_ REQUIREMENTS
SMP is located within BNFL's Sellafield site at West Cumbria. As such it utilizes many of the facilitiesavailable within the adjacentTHORP complex and on the Sellafield site generally. Listed below are themain areas where existing infrastructure/facilitiesare utilized.
!. USE OF EXISTING BUILDINGS/STRUCTURES
Due to the close proximity of SMP and the THORP complex, it has been possible to utilize certain areasof THORP to house SMP equipment. These areas include:-
a) THORP West Annex - This area is approximately1600m2and is used to house part of the Heatingand Ventilation equipment, namely the CI supply and extract fans and the C2 extract fans.
b) THORP STACK - The THORP stack is being used to house the SMP C3 and C5 discharge flues.c) General Storage Areas - Several of the general store areas within THORP will be used by SMP
as a location for equipment spares, consumablesetc.
II. USE OF EXISTING PIPED SERVICES
The following services are available to SMP from the existing site facilities:
• LP Steam• Compressed Air• Breathingair• Argon• Nitrogen• Hydrogen• Demineralized Water• Domestic Water• Condensate Disposal• C2/C3 Effluent Disposal• Electricity
III. USE OF SITE FACILITIES
The following facilities are used by SMP
a) THORP Central Control Room - The THORP control room is used to control and monitor theSMP plant environmental systems (eg. activity in air. ventilation etc). In addition this area will actas the incident control center for SMP.
b) Maintenance Facilities - SMP utilizes the workshops and decontamination/maintenance facilitiesavailable within the THORP complex.
c> Sample .Mmlysis - SMP utilizes the existing site laboratories to analyze powder and pellet samples.The samples are automatically transferred via the existing THORP sample transfer system.
d> Chaagerooms/Access Control - SMP personnel enter via the THORP complex. As such theyutilize the existing THORP changerooms and security control systems.
4.1.3-15
e) Waste Treatment Facilities - SMP utilizes the comprehensive waste treatment facilities availablewithinthe Sellafield site. These facilities cover PCM, intermediate, and low.level waste packagingand storage.
The following general site facilities are used by SMP:
a) Site Police Forceb) Site Fire Brigadec) Site MedicalCenter
L
4.1.3-16
4.1.3- 7" DEVELOPMENT REQUIREMENTS
The SMP process for manufacture of MOX fuel is based on that used within the existing MOXDemonstrationFacility (MDF). As such development of the basic process is now complete. However.further development work is underway to optimize SMP design and this can be split into 2 main areas:
• Development work to supportplant performance.• Developmentwork to supportequipment design.
1. PLANT PERFORMANCE
Although developmentwork to prove the main process is complete " ' + • " ._program is nowin place to provide plantoperatingdata. This work will involve varyir.. _'. _,:, s_parameters in orderto optimize the plantperformanceand establish acceptable tolerances f_c proc_uctquality.
II. ENGINEERING DESIGN DEVELOPMENT
This work will take the form of detail studies to prove engineering concepts prior to detaildesign/manufactureof equipment. Typical areas were this type of work is being carriedout are:
• Special radiometric instrumentation• Rod and Pellet inspection• Powder transfer• Remote maintenance
4.1.3-17
4.2 ADAPTING COMMERCIAL MOX FUEL
FABRICATION EXPERIENCE
A Hazard and Operability Study (HAZOP I)_ study was held to discuss the implications of
processing plutonia derived from weapons or 'A' grade plutonium using a commercial MOXfabricationprocess such as that currentlybeing developed for the Sellafield MOX Plant (SMP).
The SMP process is based on the conversion of civil plutonia powder (arising from the
reprocessingof fuel from MAGNOX and Thermal Oxide Nuclear.Power Stations) toMixed
Oxide (MOX) Fuel elements. Figure4.1.2.-1 indicates the main features of the proposed SMP
processand this was takenas the processbasis for the study.
HAZOP is a rigorous systematic process used principally to identify potential hazards and
assess the safety of plant designs.
4.2-1
4.2.1 Objectives of Study
The objectives of the meeting wereto:
a) Identify the key assumptionswhich neededto be madefor thestudy. The assumptions
relatedto the form of 'A'grade plutonium,its isotopic compositionetc.
b) Identify the key issues (in terms of safety, operabilityand process)which would need to
•be addressedin adaptinga commercialMOX fuel fabricationprocess to handle 'A' -
gradeplutonia.
4.2-2
4.2.2 Assumptions
The following assumptions have been made with regardto the form and isotopic composition of
the 'A'gradeplutonium:
a) 'A'grade plutonium contains 5 w/o Pu-240. The sensitivity of the study to plutonia,with Pu-240 content of < 5 w/o would also be considered.
b) The process Willreceive plutoniapowder. The conversion of weapons plutonium.to the
oxide form will occur in a facility providedelsewhere.
c) The moisturecontent of the plutoniapowderis 1.5 w/o.
d) The plutonia powder is 5 years aged. If the age is increased the dose rates would
increase due to Am in growth and the heatoutputwould increaseslightly.
e) The HAZOP I study is based on the assumption that civil grade plutonia is replacedin
terms of throughputwith plutonia,derivedfrom 'A'gradeplutonium
A comparison of these assumptions with the current SMP Reference Case is given in Table4.2.2-1.
Table 4.2.2-1. Comparisonof SMP Reference Fuel with Assumed 'A'GradeFuel
SMPREFERENCE ASSUMED'A' GRADEPuO,. PuO:
Bum Up 45 Gwd/teU 500 Mwd/teU
Aging 5 years 5 years(plus5 year cooling priortoreprocessing)
Pu-240 Content 10 W/o 5 W/oi
Moisture 1.5 W/o 1.5 W/o
4.2.3 HAZOP Study
The HAZOP I study was undertaken using the key word listing given in Table 4.2.3-1 The basis
of the study was the proposed SMP process flow diagram shown m Figure 4.1.2-1.
Table 4.2.3-2 is a record of the discussions from the meeting for each of the key words listed inTable 4.2.3-1.
A single action was generated against Nuclear Technology to determine the isotopic composition
of typical 'A' grade plutonium. This action (Action 1.1) and its response are included in theHAZOP minutes in Table 4.2.3-3.
4.2-t
Table 4.2.3-1. Key Word List for the HAZOP I
EXTERNAL DOSE SHIELDINGINTERNAL DOSE LOSS OF CONTAINMENTVENTILATION EFFLUENTSCRITICALITY FIREEXPLOSION/DETONATION IMPACT DAMAGEMIXING OF FEEDS SEISMICEXTREME WEATHER LOSS OF SERVICESINSTRUMENTATION/INTERLOCKS MAINTAINABILITYTOXICITY CORROSIONDECOMMISSIONING DOMINOHUMAN FACTORS OTHERS
4.2-5
Table 4.2.3-2. HAZOP I Record Sheet
Meeting No: 1Date: 6/12/93
Keyword Discussion Action/Recommendation
ExternalDose/ The anticipateddose rates from the 'A' grade PuO2 Action 1.1 - On D.Shielding will be lower than those being emitted from PuO: Winstanley (Nuclear
derived from civil reactors. The response to HAZOP Technology. Determineaction 1.1 (see attached)provides a comparison with the isotopic composition ofthe expected gamma and neutrondose rates between weapons ('A') gradecivil and A grade PuO2. It is seen for A grade PuO, plutoniumthat will bedose rates are lower, handledwithin the
proposed MOX productionConsequently this reduction in external dose rates facility.may enable the MOX fabrication in the proposedplant to involve more manual intervention.Additionally. the required shielding may be able to be
reduced.|
Internal It is anticipated that 'A' grad PuO: will be lessDose/Loss of radiotoxic than that used in the SMP process;.Containment However. the PuO, handled will never the less be
extremely radiotoxic and similar precautions willneed to be undertaken to ensure that its primarycontainment is maintained at all times.
Ventilation The requirements as indicated above are for all PuO2bearing material to be maintained in primaE/containment. In some instances, containment wouldbe provided by the various ventilation systems (e.g.,C5 extraction on the glovebox system and the C3/C2for secondary containment. These ventilation systemswill need to be provided even though theradiotoxicity of 'A' grade PuO: is less than that forcivil grade PuO:.
The decay heat load of 'A' grade PuO, will be lessthan that from civil grade PuO:. This may reduce theneed for additional cooling or the requirement forcooling to be provided by the ventilation sy,,;tem.
Effluents No additional liquid/solid effluents due to h;mdling of'A' grade PuO2 rather than civil grade PuO: wasidentified.
4.2-6
Table 4.2.3-2. HAZOP I Record Sheet (Continued)
Meeting No: 1Date: 6/12/93
- i....
Keyword Discussion Action/Recommendationi i |li i i ii iHi
Criticality It was noted by Nuclear Technology that thereduction of the Pu240 content of the 'A' gradematerial would result in an increase of the reactivityof the vessels. Nuclear Technology indicated that areduction of the Pu240 content from 10w/o to 5w/oreduces the safe mass bv 5 % and it is concluded thatto maintain the same level of safety, the proposedprocess vessels should be reduced by a similarfraction, though the actual reductio needs to beconfirmed.
This may have an effect on the plant capacity, thoughthe actual magnitude is unclear. However, should thecapacity be reduced, then this may be offset by howthe plant is operated, controlled, or by additionallines.
Fire/ As previousiy indicated (under ventilation) the heat....Overheating generation capacity of 'A' grade PuO2 is less than
that for civil grade PuO,. and consequently theoverheating hazard potential should be reduced.
iii i ii
Explosion/ No additional explosion or detonation hazards wereDetonation identified due to the processing of 'A' grade PuO,
rather than civil grade PuO:., ,1
Impact Damage No additional impact scenarios were identified due tothe processing of 'A' grade PuO_ rather than civilPuO:.
Mixing of Feeds The potential option that th'e plant may also processcivilian PuO: was identified. Should this occur thenincreased dose rates compared with 'A' grade PuO2can be expected. Consequently, should civilian gradePuO: be handled in addition to 'A' grade PuO: theneither additional shielding and/or lower celloccupancies would be required to maintain wholebody dose levels to acceptable levels.
Seismic No additional seismic requirements have beenidentified due to the processing of 'A' gradeplutonium rather than civil plutonium. The existingprovisions would need to be maintained.
Extreme Weather No additional design requirements have beenidentified due to the processing of 'A' gradeplutonium rather than civil plutonium. The existingprovisions would need to be maintained.
4.2-7
Table 4.2.3-2. HAZOP I Record Sheet (Continued)
Meeting No: 1Date: 6/12/93
Keyword Discussion Action/Recommendation
Loss of Services iThe lower decay heat from 'A' grade PuO,.rather .......than civil grade PuO: may reduce the need to provide!cooling following a loss of services.
!instrumentation/ The reductionin the Pu241 (and hence gamma rays) .....Interlocks may make it more diffict 't to detect PuO2in process
vessels. This may make the provision of safetyprotection systems, and product instrumentationmoredifficult to engineer.
Maintainability No additional'maintainability problems due to theprocessing of 'A' grade PuO2 rather than civil PuO:were identified. It was noted that the reduction in thedose rates from the fissile material should reducedose uptake from background contamination levelsduring maintenance.
Toxicity No additional"toxicity hazards due to the processing .........of 'A' grade PuO., rather than civil PuO2 wereidentified. Although radiotoxicity of 'A' grade PuO:is less than civil PuO, the containment requirementswould be the same.
i i
Corrosion No additional corrosion 'hazards due to the processingof 'A' grade PuO: rather than civil PuO: wereidentified.
Decommissioning No additional'decommissioning hazards due to the .........processing of 'A' grade PuO: rather than civilPuO: were identified. If the facility processed solely
'A' grade PuO: then dose rates duringdecommissioning would be less.
Domino It was noted that weapons grade material would not ....be in the form of PuO2 powder required for feed tocommercial MOX plants and that conversion to theoxide state (ie. PuO,.) will need to occur within anupstream plant. The HAZOP study does not considerthis conversion plant. However, it is noted that thecapacity of the conversion plant should match thecapacity of the MOX facility, i.e., the conversionplant should not impose a restriction on theoperability of the MOX plant.
Human Factors No additional human factor issues were raised as the "'
result of handling PuO., arising from 'A' gradeplutonium rather than civilian PuO,.
Others
Table 4.2.3-3. HAZOP Action Responses
ACTION: Determine the isotopic composition of weapons grade ('A' grade) PuO: that will be handledwithin the proposed MOX production facility.It is considered that the facility will handle PuO: that is approximately 5 years aged.
REPLY: The ratios shown below give a trend only and should not be used for specific calculationswithout verification that it is appropriate to do so.
'A' Grade Plutonium usually has a low Pu240 content (--5w/o).Nuclear Technology have in existence a MAGNOX FISPIN run for 500 Mwd/te irradiatedfuel. 5 yr aged. (PDEC 94) which has the following isotopic composition:
Pu240 4.3 w/o Pu236 9.52E-9 w/oPu241 O.19 w/o Pu238 3.8 E-3 w/oPu242 2.95E-3 w/o Pu239 95.5 w/o
A comparison between this composition and the SMP Shielding Design Basis (SDB) based onan arbitrary but appropriate system (i.e.. plutonium surrounded by 6ram steel) gives thefollowing ratios (A Grade/SDB) for 5 yr aged fuel.
RATIOS
3' n Total
Unshielded 0.05 0.11 0.05Shielded 0.11 0.11 0.11(6ram Steel)
The above ratios show that for these set of circumstances 'A' Grade plutonium give ---5 % ofthe unshielded dose rates and 10% of the shielded dose rates assumed from the reference case.
NB: Shielded dose rates are dominated by neutron dose rates. If the Pu240 content increasesthen the neutron dose rates will increase.
4.2-9
4.2.4 Main Findings
The following is a summaryof the mainissues raisedduringthe study:
a) Because the 'A'gradeplutonia contains fewergammaand neutronemitting isotopes (see
response to Action 1.1, Table4.2.3-3) the need for the fully automated processing could be
relaxed which may have particularbenefits for fuel assembly operations. This would not be
the case if the facilityhadto piocess bothcivil and 'A'gradeplutonium....
b) Though the radiotoxicityof'A' gradeplutonium is.marginallyless than civil plutoniumit
is consideredthat the containment integrity of any plutonium beating materialwill need to
be the same to that for a civil plutonia MOX plant.
c) The radiometricdecay heat arisingfrom plutonia derivedfrom 'A'grade plutoniumwillbe less than that from civilgrade material. Consequently this may reducethe requirementsto
providecooling of any plutoniumbeating materialholdup within the process.
d) The Pu-240 content of plutonia derived from 'A'grade plutoniumwill be less than the
corresponding Pu-240 content of civiliangrade material.This may require that the capacity
of any vessel thatmay hold pure plutoniapowder or MOX be reduced to ensure criticalitysafetyunder normaland faultconditions.Typicallyfor plutoniawitha Pu-240 contentof 10
w/o, a reductioninPu-240 contentto 5 w/o reducesthe safemass by S w/o.
The effects on the plant capacity because of this would need to be considered,though it is
....anticipatedthat any reductionin individualvessel capacitycould be off-set by for exampleincreasingthe numberof process lines etc.
e) It was noted that the lower Pu-241 content of 'A' grade plutonia will result in lower
external dose rates. As it is anticipatedthat some instrumentbased protection systems
within the plantmay, in some instances rely on the gamma rays being emitted from the
plutonia(eg. criticalityinstruments)thenthe availabilityof these systems could be affected.
f) The capacity of any upstream facility which may be used to convert weapons grade
plutonium to plutonia suitable for the commercial process should be such that it will not
restrict the operation of the MOX plant. In addition the upstream plutonium conversion
facility could incorporate a capability to blend the weapons grade plutonia with other
material (eg. Civil plutonia or urania) such that the criticality reactivity of the material was
no worse than civil plutonia. This would avoid increasing the difficulty in making a criticalitysafew case.
4.2-10
4.3 PLUTONIUM DISPOSITION COMPLEX INFRASTRUCTUREIN THE UNITED STATES
4.3.1 Introduction and Summary
Itwas concludedin Phase 1A that the infrastructureappearedto be establishedfor deploymentof
an ABWR Pu Disposition Complex in the United States. The brief surveysof each of the DOE
sites conducted underPhase 1Csupportthis conclusion.
TheDepartmentof Energy(DOE)has the sites andcapabilities,and theflexibilitywith these sites
and capabilities,to deploy anelectricpowerproducing,full MOX-fueledABWR PuDisposition
Complex. Forstudy comparisonpurposes,thereferencecase for deploymentof the Plutonium
Disposition Complex in the UnitedStates is a new "Greenfield," in which the facilities are
constructedand located all togetheron a hypotheticalsite at Kenosha,Wisconsin. Thepurposeof
the infrastructureportionof this studywas to determinethe extent to which the "complex"couldutilize the existing capabilitiesat one or moreof theexistingDOE andcommercialsites.
The studyincludedvisits anda collectionof dataforthe following sites:
• IdahoNationalEngineeringLaboratory(INEL)• Ne_,adaTest Site (NTS)• OakRidge Reservation (ORR)• PantexPlant• Savannah River Site (SRS)• HartfordSite• LawrenceLivermoreNational Laboratory (LLNL)• Los Alamos National Laboratory (LANL)
The study team confm'nedfrominterviewsand facility tours with DOE siteoffice and site
contractorrepresentativesthat sites, facilities, resourcesand capabilities alreadyexist at DOE and
commercialsites which can be used at a cost and schedule saving.
The fast five of these sites arealso being evaluatedfor operationalelementsby the nuclear
weapons complexreconfigurationprogram. This section covers these sites, and summarizesthe
capabilitiesfor conducting the Pu dispositionfunctions,and for managingthe interfacesbetween
the Pudisposition and the nuclearweaponscomplexreconfigurationfunctions,at each site. A brief
visit was made to each site. The introductionby GEfor each visit includedFigures 1-2plus a
descriptionof the ABWR plant. The discussionsthen coveredas many of the topics shownin
Figure2 in as muchdetailas allowedby the timeavailable.
4.3-1
In order to provide some consistency between the evaluations for the weapons complex and Pu
disposition programs, the degree of readiness at the site for each of the capabilities required for Pu
disposition was classified as being at one of following three levels, in a manner similar to that used
by the nuclear weapons complex reconfiguration program:
• EXISTING (E) - .Ca_.bility exists at the site and can be applied to Pu disposition withonly minor refurbishing required.
• UI_RADE (U) - Capability exists at the site and can be applied to Pu disposition withupgrading required, such as renovation of an existing facility and installing some newequipment.
• GREENFIELD (G) New capability required at the site.
The latter three sites, Hanford, LLNL, and LANL, although not considered operational site
candidates, have facilities or technical capabilities which could be supporting functions for Pu
disposition, and arc also covered in this section. The mission of the HartfordSite, although it
contains facilities applicable to Pu disposition, is believed to be primarily to implement
environmental restorations. LLNL and LANL were not considered for production-type activities
such as MOX fuel fabrication or reactor operation. Visits to these sites were made during and prior
to Phase IC. The results reported here include for completeness some of the conclusions discussed
in the earlier Phase IB report.
It is clear that considelable cost effective, installed capability is available within the DOE
community now for meeting the Pu disposition needs in the near term with an electric power
producing, full MOX-fueled ABWR plant. These capabilities can be implemented in the short term
(8-10 years) with effort ranging from minor refurbishing to upgrading of existing facilities, with
only a few requirements, such as the reactor, being Greenfield efforts at all sites. It is anticipated
that a minimum cost deployment will be to locate the entire Pu Disposition Complex at one site.
SRS, ORR, and INEL already have in place significant applicable elements. It is also possible to
take advantage of unique capabilities which exist at individual sites and create a distributed
"complex," with some additional cost for transportationbetween sites.
Section 4.3.2 provides a definition of the requirements for the Pu Disposition Complex,
and Sections 4.3.3-10 provide the results of the evaluations.
4.3-2
4.3.2 Pu Disposition Complex Site Requirements
This sectiondescribesthecapabilitiesandfacilitiesneeded bya site in orderto beconsideredfor
locationof some or allof theABWR Pudispositionfunctions,andthe capabilitieswhich the site
musthave for maintainingcriticalinterfaceswithotherprogran_(such asthe NuclearWeapons
ComplexReconfigurationProgram).
OVERALL SITE QUALIFICATION
A completelyGreenfieldsite for theABWRPu Disposition Complexis the baselinefor
comparativepurposes. Theactualsite selectedwill probablynot be a Greenfield,given that several
attractivesitesalreadyexist withinthe DOE infrastructure.Thecandidatesite mustbe sufficiently
largeand have thephysicalresourcesand technicalexperiencein nuclearprogramstoprovide the
space, utilities, and stafffor the operation,safety,security,safeguardsand maintenance of the Pu
disposition functions.
Pu FEED MATERIAL INTERFACE
The study was directed to assume that the weaponsprogramwill provide PuO2feed material for
Pu disposition in accordancewith the specificationsfor fabricationof MOX fuel for ABWR fuel
pins. The site must have the capability to receive this PuO2. A site with the capability for Pu
processing, and which therefore could also be the producer of the PuO2, presents the advantage of
minimizing the complexity and cost for PuO2 transportationfromthe interim storagesite to the Pu
Disposition Complex.
MOX FUEL FABRICATION
The Pu Disposition Complex includes a MOX fuel fabricationfacility whichreceivesand stores
PuO2and UO2, blends the PuO2and UO2, sintersthe MOX pellets, loads thefuel pins,
assembles the fuel assemblies, stores thecompletedassemblies, andpackages the assembliesfor
transportto thereactor. The site musthave the capabilityforPu processingwhich can be applied
to MOX fuel fabrication.A site with the capabilityfor both MOX fabricationandreactoroperation
presentsthe advantageof minimizingthecomplexity and cost for transportationof freshMOX fuelassemblies.
4.3-3
TRITIUM TARGET FABRICATION INTERFACE
It is assumed that the weapons program will provide the tritium targetrods for tritiumproduction in
accordance with the specifications for including the tritium targetrods in the MOX fuel assemblies
for the ABWR. The fabrication of the target components and the completed targetrods can be
accomplished by commercial vendors. The site must have the capability to receive the rods. A site
with the capability for integrating theprocurement of the components of the targetrods produced
by commercial vendors, and then completing fabrication of the rods onsite, presents the advantageof minimizing the complexity and cost of tritium target fabrication.
REACTOR PLANT SITING
The Pu Disposition Complex being proposed by GE includes a 1350 MWe ABWR which is
essentially identical to the ABWRs being built now in Japan. The candidate site must have the
capability for locating this reactor on or near the site. A site with the capability for locating both the
MOX fuel fabrication facility and the ABWR plant on or near the site presents the advantage of
mimmizing the complexity and cost of transportation of the fresh MOX fuel assemblies.
POWER TO THE GRID
The Pu Disposition Complex will sell the electric power generated by the ABWR plant to an
Independent Power Producer (IPP)/Utility customer. The candidate site must have the capability
for access to a commercial grid for transmission of the power offsite. A site in an area with a need
for electric power presents the advantage of maximizing revenue from sale of the power.
SPENT FUEL DISPOSAL INTERFACE
The study was directed to provide ten years of storage at the reactor site for spent fuel. This is
provided by the in-reactor spent fuel pool of the GE ABWR design. It is assumed that the spent
MOX fuel assemblies will be turned over to the U.S. Nuclear Waste Disposal Program after the
period in the in-reactor spent fuel pool. It is anticipated that temporary storage of the fuel
assemblies discharged from the in-reactor pool will be required pending direction for disposal.
The candidate site must have the capability for handling spent MOX fuel assemblies for transport to
this temporary storage location. A site with the capability for additional temporary spent fuel
storage on or near the site presents the advantage of minimizing the complexity and cost of
transportation of the spent MOX fuel assemblies to some other temporary storage location pending
f'mal disposal.
4.3-4
WASTE MANAGEMENT
The Pu Disposition Complex will utilize state-of-the-artmethods andprocedures to minimize the
amount of waste requiring disposal. The candidate site must have the capability for handling high
levei, transuranic, low level, hazardous, and mixed wastes, and for packaging the waste for
shipment to approved facilities for disposal, and for managing the interface with the US Nuclear
Waste Disposal Program. A site with temporary or permanent waste disposal facilities on or near
the site presents the advantage of minimizing the complexity and cost of transportation of thewastes.
TRITIUM RECOVERY INTERFACE
It is assumed that the irradiated tritium target rods will be turned over to the weapons program after
discharge from the reactor and disassembly from the MOX fuel assemblies. The candidate site
must have the capability to coordinate the transport of the rods to the tritium recovery facilities. A
site with the capability for tritium processing, which therefore could conduct the tritium recovery
operations, presents the advantage of minimizing the complexity of shipping irradiated tritium
target rods and transportation costs.
SAFEGUARDS AND SECURITY
The Pu Disposition Complex will require safeguards and security for the storage, handling,
processing, and transport of special nuclear materials (SNM) from the point of receipt of the PuO2
from the weapons program through insertion of the MOX fuel assemblies into the reactor core and
start of irradiation. After irradiation, the fuel assemblies will require safeguards and security
equivalent to that for spent fuel in commercial reactors. The candidate site must have the cat-ability
to provide for the accountability of the SNM and for protection against SNM diversion.
SAFETY AND ENVIRONMENTAl, APPROVAL
The Pu Disposition Complex will require safety approval by DOE and/or NRC, environmental
approval by the EPA, and federal, state, and local pemaitting agencies. The ABWR plant will soon
be certified by the NRC, and GE as the reactor manufacturer will carry the main burden of the final
safety approval process for the plant owner (i.e. either the government or an
IPP/utility). The MOX fuel fabrication plant is expected to be owned by DOE, with safety
approval by either DOE or NRC. The plant owner is responsible for obtaining the environmental
approval, and for obtaining the additional federal, state, and local permits required for operation of
the complex. The main support to the owner will be supplied by the reactor manufacturer and
4.3-5
architect/engineer, for the source terms and plant description, and by the site, for the environmental
conditions and the impact of operation of the complex on this environment. The candidate site must
have the capability to provide effective support to the approval process, including analyses,
documents, and participation in reviews as required. A web characterized and documented site
will present the advantage of cost effectiveness and low risk for safety and environmental
approval.
TRANSPORTATION
The Pu Disposition Complex will require access to the site and transport for SNM and wastes both
on-site and off-site. The candidate site must have the capability for receiving heavy equipment
during construction, for unloading and inspection of incoming SNM shipments, for transport of
SNM and wastes on-site between facilities and between protected areas, for packaging SNM and
wastes for off-site shipment, and for strict accountability of the SNM.
SUPPORTING SITE ASSETS
The Pu Disposition Complex will require support in addition to the areas noted above. The
candidate site must have a management organization experienced in large nuclear projects, a
technical and production oriented work force, adequate utilities such as water and power,
technology capabilities, strict quality assurance requirements, and a community attitude that is
favorable to Pu processing and reactor programs.
4.3.3 Capabilities at the Idaho National Engineering Laboratory
The Idaho National Engineering Laboratory (INEL) was visited on 15 December 1993. A full day
of discussions was held with DOE site office and site contractor management and staff to obtain an
overview of the capabilities at INEL applicable to a Pu Disposition Complex. The summary
presented in Table 1 is the result of these discussions, followup telephone calls, and review of
documents describing the INEL facilities. These results should be considered preliminary pending
thorough review with the INEL staff.
4.3.4 Capabilities at the Nevada Test Site
The Nevada Test Site (NTS) discussions were held with DOE management and staff at the DOE
Nevada Operations Office in Las Vegas on the morning of 8 December 1993. The summary of the
results is presented in Table 2, based on these discussions and review of documents describing the
4.3.6
NTS facilities. These results should be considered preliminary pending thorough review _,itt_ the
NTS staff.
4.3.5 Capabilities at the Oak Ridge Reservation
The Oak Ridge Reservation (ORR) was visited on 11 November 1993. The full day of discussions
with site contractor management and staff at the Y12 facility included a brief driving tour of the
Y12, K25, and X10 (ORNL) areas, the CRBR site, and the barge docking capability on the Clinch
River. The summary of the results is presented in Table 3, based on these discussions, foUowup
telephone calls, and review of documents describing the ORR facilities. These results should be
considered preliminary pending thorough review with the ORR staff.
4.3.6 Capabilities at the Pantex Plant
The Pantex Plant discussions were held with DOE management and staff at the DOE Albuquerque
Operations Office and the DOE Amarillo Area Office on the morning and afternoon, respectively,
of 9 December 1993. The summary of the results is presented in Table 4, based on
these discussions. These results should be considered preliminary pending thorough review with
the Pantex staff.
4.3.7 Capabilities at the Savannah River Plant
The Savannah River Site (SRS) was visited on 8-10 November 1993. The meeting was sponsored
by the DOE Savannah River Site Office, which also coordinated the participation of the DOE
Southeastern Power Administration (SEPA) in the meeting. Discussions were held with
management and staff of the site contractor, and representatives of SEPA and the Southern
Company (an IPP/Utility with interests throughout the southeast and beyond). The ,.,i_k included
several tours of potential facilities for Pu disposition. The results of these discussions, an earlier
visit (August 1993) devoted to MOX fuel fabrication capability, followup telephone calls, and
review of documents describing the SRS facilities, are presented in Table 5. These results should
be considered preliminary pending thorough review with the SRS staff.
4.3.8 Capabilities at the Hanford Site
The Hanford Site was initially visited during Phase 1B to evaluate the Fuels Materials and
Engineering Facility (FMEF) as a building which might accommodate the MOX fabrication
activities for the ABWR Pu Disposition Complex. The Secure Automated Fabrication (SAF)
4.3-7
facilities which were designed for the fabrication of Fast Flux Test Facility (FFTF) fuel are of
particular interest. Since the SAF line was designed for fast reactor fuel, the lower plutonium
concentration and larger pellets associated with an ABWR design preclude the use of most of file
equipment installed in the facility. Despite these different types of ceramic the process utilized in
the SAF line is similar to those being considered for the ABWR Pu Complex and the safety and
accountability procedure should be nearly identical. Some Pu handling technology is available
from weapons work completed in previous years which may be applicable to Pu disposition. The
information developed during site restoration activities will be directly applicable to the waste
handling associated with MOX fabrication and reactor operation. The plutonium processing
experience and capability at Hanford could provide technology support to the Pu Disposition
program, and the FMEF could be upgraded for ABWR MOX fuel fabrication.
4.3.9 Capabilities at the Lawrence Livermore National Laboratory
The Lawrence Livermore National Laboratory was initially visited during Phase 1B activities to
review the processing technology for converting plutonium metal to the oxide utilizing the
hydride/dehydride process which is utilized to remove plutonium from a retired weapon. The
activities completed by the laboratory have resulted in the design and demonstration of a workable
process. A great deal of effort has been devoted to the robotics systems necessary for remote
cjpcrationof the hydride/dehydride processes. Much of the technology will be applicable to remote
processing associated MOX fuel fabrication. Several visits were made to Lawrence Livermore
National Laboratory during Phase 1C activities to furtherreview plutonium conversion technology.
In addition, the processes being developed to handle the scrap and waste steams generated during
plutonium hydride/dehydride activities were reviewed. The Plutonium processing experience and
capabilities at LLNL could provide technology support to the Pu Disposition Program.
4.3.10 Capabilities at Los Aiamos National Laboratory
The Los Alamos National Laboratory (LANL) was initially visited during Phase 1B activities to
assess progress on the technology development activities for Pu-to-Pu02 conversion, and the
potential for fabrication of mixed oxide/gadolinia (MOX/GAD) test assemblies to confirm the
burnout rate of gadolinia in an ABWR. In addition to the burnout rate information, MOX/GAD
processing capabilities could be demonstrated and final fuel fabrication facility design information
developed. "lhe technology utilized at this site to handle plutonium containing scrap and waste
streams, and accountability and safeguards, were evaluated for applicability to an ABWR complex.
A brief review of the Automated Retirement and Integrated Extraction System (ARIES) conceptual
4.3-8
design information was provided because much of the technology to be utilized in the retirement of
plutonium weapons components is directly applicable to the ABWR Pu Disposition Complex. A
second visit was made on October 21, 1993, to review the storage requirements for plutonium
which had been removed from the weapons. The processing technology available for the
conversion of weapons plutonium to plutonium dioxide was discussed and necessary development
activities defined. LANL personnel have begun the development of a fiber optic technique for
analysis of plutonium content, plutonium isotopic, and impurities through the use of laser
technology. Since all of the electronics are located outside of the glove box areas, the technology
is particularly attractive for use in a MOX facility. Generally, these analysis techniques are orders
of magnitude more rapid than standard analytical chemistry techniques. The Plutonium processing
experience and capabilities at LANL could provide technology support to the Pu Disposition
program.
4.3-9
t
Plutonium Dispositon Study-Phase 1CI
DOESTATEMENTOFWORK
GENERALELECTRIC- Advanced Boiling Water Reactor
!
Task 4: Investigate deployment strategies.
Assess alternate deployment strategies Including useof existing DOE and oommeralal facilities, Integrationof fuel fabrication and reactor complex facilities,complex location, transportation and logisticrequirements for the plutonium feed material, wastestream products, and spent fuel.
FIGURE 1 GE STATEMENT OF WORK
J
@ THIS MEETINGII I
I ii
• Provide information on ABWR Plant and Progress of PlutoniumDisposition Study
• Obtain Information on existing Infrastructure for evaluation of feasibility,cost and schedule for deployment of Plutonium Disposition complex,inc,uding tritium production, and possible Integration with weapons
= reconfiguration programI
Pu Feed Material (_Jnterface) Waste ManagementMOX Fuel Fabrication Tritium Recovery (Interlace)Tritium Target Fab (InteHace) Safeguards and SecurityReactor Plant Siting Safety Approval
Power to the Grid (Intedace) Environmental Approval
Spent Fuel Disposal (Interface) Transportation
Supporting Site Assets
FIGURE 2 INFRASTRUCTURE EVALUATION OBJECTIVES
DRAFT 15Jan94 file IDEPLOY.
**************************** DRAFT TO BE REVIEWED WITH INEL ********************************
Table 1 PLUTONIUM DISPOSITION COMPLEX DEPLOYMENT CAPABILITIES AT THE IDAHO NATIONAL ENGINEERING LABORATORY
G = Greenfield. New Capabilities Required.
U = Upgrade of Existing Capabilities Possible.
E = Existing or Planned Capabilities Meet Requirements.
CAPABILITY REQUIREMENT CURRENT CAPABILITY AT INEL
Overall Site Qualification E - Currently operating. Has served for "45 years as DOE site for part of
Nuclear Materials Production Complex, reactor R&D, and fuel processing.
INEL is still performing these missions on a reduced scale.
- Occupies "890 square miles -29 miles west of Idaho Falls, ID
- Eleven Technology Areas: Test Area North (TAN), Test Reactor Area (TRA),
Central Facilities Area (CFA), Radioactive Waste Nanagement Complex (RWMC)
Auxiliary Reactor Area (ARA), Power Burst Facility/Power Excursion
Reactor Test (PBF/SPERT), Idaho Chemical Processing Plant (ICPP), and
Argonne National Laboratory West (ANL-W), Idaho ReseaLch Center (IRC),
Idaho Supercomputer Center (ISC), INEL Engineering
_j PuO2 Feed Material Interface
' (IF w/Nuclear Weapons Complex Reconfig Program)
- Pu Receving and Storage E - Management and staff experienced with SNM handling including Pu
(within weapons program) - SNM vault in Fuel Processing Restoration (FPR) facility could receivePu metal or oxide
- Pu-to-PuO2 Conversion E - Management and staff experienced with Pu processing. Pu metal core
(within weapons program) fabricated and irradiated in Materials Test Reactor.
- Hot cells in FPR facility can be used for Pu-to-PuO2 conversion
- PuO2 Feed Interface Management E - Experience managing SNM is applicable to management of interface
with supplier of ?uO2 feed material whether or not Pu-to-PuO2conversion function is located at INEL
MOX Fuel Fabrication
- MOX Pellet and Pin Fabrication U - Experience with Pu processing is applicable to extension of capability
to MOX blending, sintering, and pin loading. [TBR]
- Fuel Processing Restoration (FPR) facility was completed in 1992, is
uncontaminated, and can be modified to house the NOX lines.
New equipment for powder blending, pellet sintering, and fuel
pin loading and closure is needed.
- FPR is six stories tall, 160,000 square feet floor space,
state-of-the-art radiation protection, contamination control,
decontamination, natural phenomena protection, remote handling,
50/5 ton bridge crane, DOE approved SNM vault
- ABWR MOX Assembly Fabrication U - FPR can be modified for MOX Assembly fabrication
- Fresh MOX Assembly Storage U - FPR can be modified for fresh MOX Assembly storage
Table I (Con't) PLUTONIUM DISPOSITION COMPLEX DEPLOYMENT CAPABILITIES AT THE IDAHO NATIONAL ENGINEERING LABORATORY
CAPABILITY REQUIREMENT CURRENT CAPABILITY AT INEL
Tritium Target Fabrication Inter_ace
(IF w/Nuclear Weapons Reconfig P_ ram
- ABWR Tritium Target Fabrication U -- Could be located in FPR together with or separate from MOX lab.
(within weapons program) New equipment for pin loading and closure needed.
- Tritium Target Fab Interface Management E - Experience in managing procurement from commercial vendors, and in
managing design and irradiation of NP-MHTGR tritium target development
program and irradiation of LWR tritium targets, is applicable
to management of interface with supplier of tritiue targets whether
or not ABWR tritium targets are fabricated on-site
Reactor Plant Site Qualification
- Reactor Site E - ABWR plant could be located on available open, unencumbered,
contamination-free 10,240 acre site (more if needed) completelywithin INEL boundaries. This site was selected for NPR.
- Commercial site could be made available outside protected zones
similar to WPPS plants on Hanford reservation
- Site Qualification E - Site characterization is completo at NPR site, where Pu Disposition
Complex would be located
- Reactor Plant G - ABWR plant will be a new facilityI
t_ - Reactor Plant Operation E - Management and staff experienced in reactor operation
- 52 special purpose reactors designed, construct*d and operated.
Currently 12 operable reactors at INEL.
Power to the Grid Interface
(IF w/IPP/Utility)
- Proximity to Commercial Grid G - New grid needed and in progress. Power transmission corridor from
mid-Idaho to Las Vegas area initiated by Idaho Power Company to handle
increase in power transmission from Canada, and originally expected
from NPR, currently in Environmental Impact Statement process with
Record of Decision expected in near future. Note that this corridor
will be completed regardless of any INEL activities.
- Reactor-to-Commercial Grid Transmission G - New transmission lines needed. Idaho Power will pr-vide transmission
services for power produced at INEL as part of electic power sales
agreement.
- Electric Power Sales U - Idaho Power proposed to lead formation of regional utility consortium
to construct and operate power conversion side of NPR and/or to
distribute electric power generated by NPR. Agreement on purchase
of power generated by Pu Disposition Complex by Idaho Power could
negotiated contingent upon pricing, availability, and reliability
of offered power.
- IPP/Utility Interface Management E - Experience managing large electric power distribution system for INEL
and past working relationship with Idaho Power is applicable to
managing interface with IPP/Utility for sale and transmission of
power generated by Pu Disposition Project
Table I (Con't) PLUTONIUM DISPOSITION COMPLEX DEPLOYMENT CAPABILITIES AT THE IDAHO NATIONAL ENGINEERING LABORATORY
CAPABILITY REQUIREMENT CURRENT CAPABILITY AT INEL
Spent Fuel Storage Interface
(IF w/Nuclear Waste Disposal Program)
- Spent Fuel Storage E - Environmental Impact Statement (EIS) in progress for storage of
(within waste disposal program) commercidl spent fuel (DOE Programmatic Spent Nuclear Fuel Management
and INEL Environmental Restoration and Waste Management Programs EIS).Schedule is draft Jun94.
- Currently available pools include Idaho Chemical Processing Plant (ICPP}and naval reactor facilities
- Construction of additional spent fuel storage facilities planned
- Manages development of storage casks. Conducted Dry Rod Consolidation
Technology (DRCT) project. Conducted testing and analysis of metal
casks for dry storage.
- Spent Fuel Transport E - Management and staff experienced in spent fuel transport, including LWR
(within waste disposal program) - Manages development of transportation casks. Cask Systems Development
Program supports acquisition of prototype casks for spent fuel
transport from reactors to repository.
- Manages transport of spent fuel, including LWR, to Hot Fuel Examination
Facility (HFEF) at INEL
- Manages transport of spent LWR fuel assemblies from reactors as part
of Spent Fuel Storage Cask Testing Program and other programs!
- Spent Fuel Interface Management E - Experience managing spent fuel operations is applicable to managementof interface with Nuclea_ Waste Disposal Program whether or not spent
fuel is stored or transported h_ INEL
Waste Management
- High Level Waste (HLW) E - Management and staff have recent experience in designing, constructing,
operating, and maintaining facilities to handle, process, and store HLW
- Waste treatment capabilities/facilities include handling, processing,
calcining, and storage
- Currently stored at Idaho Chemical Processing Plant (ICPP)
- New Waste Management Center is in planning process
- Transuranic Waste (TRU) E - Management and staff have recent experience in designing, constructing,
operating, and maintaining facilities to handle, process, and store TRU
- Retrievable contact handled TRU and TRU-mixed wastes currently
stored at Transuranic Storage Area (TSA) of Radioactive Waste
Management Center (RWNC)
- Retrievable remote handled TRU and TRU-mixed waste currently stored
at Intermediate-Level Transuranic Storage Facility (ILTSF) of RWNC
- Low Level Waste (LLW) E - Management and staff have recent experience in designing, constructing,
operating, and maintaining facilities to handle, process, and store LLW
- LLW disposal currently at Subsurface Disposal Area (SDA) of RWHC
Table 1 (Con't) PLUTONIUM DISPOSITION COMPLEX DEPLOYMENT CAPABILITIES AT THE IDAHO NATIONAL ENGINEERING LABORATORY
CAPABILITY REQUIREMENT CURRENT CAPABILITY AT INEL
Waste Management (ton't)
- Hazardous Waste E - Management and staff have recent experience in designing, constructing.
operating, and maintaining facilites to handle, process, and storehazardous wastes
- Mixed Waste E - Management and staff have recent experience in designing, constructing,
operating, and maintaining facilities to handle, process, and storemixed wastes
- Nuclear Waste Interface Management E - Experience managing all forms of nuclear waste is applicable to
management of _nterface with Nuclear Waste Disposal Program
Tritium Recovery Interface
(IF w/Nuclear Weapons Complex Reconfig Program}
- Extraction G - A new facility to extract tritium from targets irradiated in
(within weapons program} the ABWR Pu Disposition Complex is required
_ - Purification G - A new facility for purifying the tritium to specifications
_o (within weapons program) for shipment to the RTF at SRS is requiredI
- Reservoir Loading G - New facility required. It is believed the Replacement Tritium
(within weapons program) Facility at SRS has been designated by DOE for reservoir loading
function for the weapons program.
- Tritium Recovery Interface Management E - Experience managing tritium target development and testing at INEL
under NPR program is applicable to management of tritium recovery
interface whether or not some or all of the tritium recoveryfunctions are located at INEL
Safeguards & Security
- Accountability E - Management and staff experienced in implementing strict
accountability procedures for storage, handling, and transport
of SNM in all forms including uranium and plutonium Iota1,
oxide, components, and waste
- Protection U - INEL dedicated safeguards and security contractor experience, well
trained and equipped to provide necessary protection and initiate
effective emergency responsed to security infractions, is applicable
to providing additional personnel, fences, and guard posts for
Pu Disposition Complex
Table I (Con't) PLUTONIUM DISPOSITION COMPLEX DEPLOYMENT CAPABILITIES AT THE IDAHO NATIONAL ENGINEERING LABORATORY
CAPABILITY REQUIREMENT CURRENT CAPABILITY AT INEL
Safety and Environmental Approval
- Safety Approval U - INEL has experience with safety approval by DOE of nuclear facilitiesincluding reactors.
- INEL technical staff provides assistance to Office of Nuclear Reactor
Regulation (NRR) in instrumentation and control systems, electrical and, xn sorvxco and pro-service inspectionmechanical components and systems " - "
amd testing of piping systems, equipment qualification, radiological
issues, operator licensing examinations and training programs, standard
technical specifications and plant specific technical specifications,
license renewal activities, ALWR issues, and thermal/hydraulic analysis.
Recent emphasis has been loss on operational aspects and more on
ALWR issues.
- DOE and NRC requirements are similar. Experience with DOE safety
approval procedures, and with technical assistance to NRR, is
applicable to extension of capability to accomodate NRC procedures.
- Environmental Approval E - Management and staff experienced with environmental approvalrequirements to obtain Record Of Decision (ROD) by DOE with EPA
concurrence for nuclear facilities including reactors.
- Site data developed for ROD issued to INEL for Environmental ImpactI
Statement for Special Isotope Separation Project, and other
C_ evaluations (EIS and EA reports listing available upon request),
is applicable to Pu Disposition Complex
- Federal/State/Local Permitting E - INEL has experience in permitting at all levels
Transportation- Site Access E - Access by air via Idaho Falls, ID
- Access by road includes two US highways and one State highway crossing
INEL boundaries. -230 miles on-site roadway classified principal arterial
and major collector routes. 500+ ton load capability on INEL roads.
- Access by rail is Union Pacific connection at Scoville Siding to
government-owned spur line linking developed areas within INEL. Gantrycrane at Scoville Siding and spur line handle up to 160 ton loads,
-15' high x 10' wide for heavy equipment delivery during construction.
- Unloading & Inspection E - Management and staff experienced in receiving SNM in all forms
on sxte transport in SNM- On-Site Transport E - Management and staff experienced in - "
in all forms within and between protected zones
- Two US highways and state roadway cross INEL boundaries
- Droposed ABWR plant and MOX fuel fabrication facility siteswithin -3 miles of each other, with no road crossings or other
facilities in between
- Packaging for Shipment E - Management and staff experienced in shipping SNM in all forms
Table 1 (Con't) PLUTONIUM DISPOSITION COMPLEX DEPLOYMENT CAPABILITIES AT THE IDAHO NATIONAL ENGINEERING LABORATORY
CAPABILITY REQUIREMENT CURRENT CAPABILITY AT INEL
Supporting Site Assets
- DOE Site Office E - Experienced in directing nuclear programs and large projects.
52 reactor projects have been completed at INEL.
- Sponsored and participated in meeting with GE Pu disposition team
- Site Contractor Management E - Operational experience managing nuclear processing and large
projects including many special purpose reactors
- Expressed strong support for use of electric power producing
fission reactor for Pu disposition and confidence that complex
could be built and operated successfully at INEL
- Considers plutonium to be a national resource
- Work Force - Technical and production oriented work force can support large
plutonium processing and reactor construction and operation project
- INEL has -12,500 employees of which "4600 hold professional degrees
- Utilities E - Adequate water supply and distribution system for reactor makeup
and complex operations
- Adequate electric power provided by Idaho Power 230 KV transmission
line to INEL to support large construction project
- Technology Development E - Argonne National Laboratory, West is technology center for DOE programs
_J - Quality Assurance E - Experienced in working to DOE quality assurance requirements! °
Also NRC and industry related (such as ASME, ANS, IEEE, NQA1 & 2, etc.)
"_] quality assurance requirements.
- Safety E - Management and staff experienced in meeting or exceeding all health
and safety requirements associated with nuclear programs. Long
history of safe operation at INEL.
- Environmental Protection E - Management and staff experienced in protection of environment.
Environmental restoration programs are operational at INEL. [TBR]
- Community Support E - INEL management believes there is broad community support in Idaho for
nuclear activities based on long history of safe operations at INEL
which will facilitate public acceptance of Pu Disposition Complex
- Idaho Falls community leaders under "Initiative 2000" will support
new nuclear activities at INEL which meet environmental concerns [TBR]
- President of Idaho State AFL-CIO supports present and future
projects at INEL
- Long history of support by Idaho elected congressional, state,
county, and city officials. Congressional delegations unanimously
supported defense-related projects such as Special Isotope Separation
Project (producing plutonium) and Complex 21 at INEL. New nuclearactivities WHICH DO NOT INCLUDE PERMANENT STORAGE OF NUCLEAR WASTE
SUCH AS SPENT FUEL will be welcomed. [TBR]
- Public interest groups include Environmental Defense Institute.
Is requesting large amount of information on radioactive and chemicalreleases and accidents and worker radiation records under FOIA.
DRAFT 15Jan93 file NDEPLOY.
************************ DRAFT TO BE REVIEWED WITH NTS ***********************
Table 2 PLUTONIUM DISPOSITION COMPLEX DEPLOYMENT CAPABILITIES AT THE NEVADA TEST SITE
G = Greenfield. New Capabilities Required.
U = Upgrade of Existing Capabilities Possible.
E = Existing or Planned Capabilities Meet Requirements.
CAPABILITY REQUIREMENT CURRENT CAPABILITY AT NTS
Overall Site Qualification E - Currently operating. Has served for -40 Fears as DOE site for
nuclear weapons testing. Current mission is to maintain readiness
to resume nuclear testing if required.
- Occupies -1350 square miles -65 miles northwest of Las Vegas, NV.
Bordered on three sides by additional 4120 square miles federally
controlled (Nellis Air Force Range).
PuO2 Feed Material InterfaceI
(IF w/Nuclear Weapons Complex Reconfig Program}OO
- Pu Receiving and Storage G - New capability needed for handling Pu in forms for processing.
(within weapons program)
- Pu-to-PuO2 Conversion G - New capability needed for Pu-to-PuO2 conversion processing
(within weapons program}
- PuO2 Feed Interface Management U - Experience managing nuclear weapons testing could be extended
to capablility for management of interface with PuO2 supplier
Mox Fuel Fabrication
- MOX Pellet and Pin Fabrication G - New capability needed for MOX fuel fabrication processing
- ABWR MOX Assembly Fabr'ication G - NOX assemblies could be fabricated in greenfield facility
built for MOX pellet and pin fabrication
- Fresh MOX Assembly Storage G - Fresh MOX assemblies could be stored in greenfield facility
built for NOX pellet and pin fabrication
Tritium Target Fabrication Interface
(IF w/Nuclear Weapons Complex Reconfig Prog)
- ABWR Tritium Target Fabrication G - Target rods could be fabricated using components from commercial
(within weapons program) vendors in greenfield facility built for MOX pellet and pin fabrication
- Tritium Target Interface Management E - Experience managing procurement from commercial vendors is applicable
to management of interface with supplier of tritium targets whether
or not ABWR tritium targets are fabricated on-site
Table 2 (Con't) PLUTONIUM DISPOSITION COMPLEX DEPLOYMENT CAPABILITIES AT THE NEVADA TEST SITE
CAPABILITY REQUIREMENT CURRENT CAPABILITY AT NTS
Reactor Plant Site Qualification
- Reactor Site E - 360 acre site identified for Tritium Supply Site within I0,000 acresfor Complex 21 available in east-central NTS. More if needed.
- Site Qualification U - Qualification of NTS for nuclear weapons testing activities isapplicable towards qualification as operating reactor site.
Has not participated in NPR or other reactor siting evaluations.
- Reactor Plant G - The ABWR plant will be a new facility
- Reactor Plant Operation G - Reactors have not been built at NTS. Reactor operation will hea new function.
Power to the Grid Interface
(IF w/IPP/Utility}
- Proximity of Commercial Grid E - Southern Nevada is hub for power transmission corridors connecting loadcenters and generation systems in Utah, California, Nevada, Arizona
s - Reactor-to-Commercial Grid Transmission G - New -100 mi transmission line and substation needed to link 1350 MWoreactor electrical output to 500 kV power transmission grid [TBR]
- Electric Power Sales G - Agreements needed- Potential value of baseload power is [TBD]
- IPP/Utility Interface Management U - Experience managing power distribution sytem for RTS applicableto management of interface with IPP/Utility. Extension of
capability needed to handle transmission and sale_ of 1350 MWe.
Spent Fuel Storage Interface
(IF w/Nuclear Waste Disposal Program)
- Spent Fuel Storage G - New capability needed for storage of spent fuel at NTS
(within waste disposal program)
- Spent Fuel Transport G - New capability needed for transport of spent fuel
{within waste disposal program)
- Spent Fuel Interface Management U - Extension of capability needed for management of interface withNuclear Waste Disposal Program whether or not spent fuel is
stored or transported by NTS
Table 2 (Con't) PLUTONIUM DISPOSITION COMPLEX DEPLOYMENT CAPABILITIES AT THE NEVADA TEST SITE
CAPABILITY REQUIREMENT CURRENT CAPABILITY AT NTS
Waste Management
- High Level Waste (HLW) E - Management an4 staff experienced in handling HLW [TBR]
- Transuranic Waste (TRU} E - Management and 3tall experienced in handling TRU [TBR]
- Low Level Waste (LLW) E - Radioactive Waste Management Site in Area 5 includes 92 acres
for surface storage and disposal of LLW
- Hazardous Waste E - Management and staff experienced in handling hazarsous wastes [TBR]
- Mixed Waste E - Management and staff experienced in handling mixed wastes [TBR]
- Nuclear Waste Management Interface E - Experience managing nuclear wastes is applicable to management
of interface with Nuclear Waste Disposal Program
Tritium Recovery Interface
(IF w/Nuclear Weapons Complex Reconfig Prog)
- Extraction G - New capability required
(within weapons program)
- Purification G - New capability required
(within weapons program)
!- Reservoir Loading G - New capability required
(within weapons program)
- Tritium Recovery Interface Management U - Similar to spent fuel. Extension of capability needed for manageDent
of tritium recovery interface, whether or not some or all of the
tritium recovery functions are located at NTS
Safeguards & Security
- Accountability U - Management and staff experienced in strict accountability of
nuclear weapons compon-nts. Upgrading needed for accountability
of Pu in forms for processing.
- Protection U - NTS experience in protection of nuclear facilities is applicable to
providing additional personnel, fences, and guard posts for
Pu Disposition Complex
Safety and Environmental Approval
- Safety Approval U - Experienced with safety approval by DOE for weapons testing related
facilities. This capability will need to be upgraded for DOE and
NRC requirements for reactor and Pu processing facilities.
- Environmental Approval U - Management and staff have experience with Environlental Assessment andEnvironmental Impact Statement requirements. Currently preparing data
for EIS for upgrade alternative for Assembly/Disassembly function under
Nuclear Weapons Complex Reconfiguration Program. This exper_ Jnce is
applicable to extension of capability needed for reactor and
Pu processing facilities.
- Federal/State/Local Permitting E - Experienced in permitting at all levels
Table 2 {ton't} PLUTONIUM DISPOSITION COMPLEX DEPLOYMENT CAPABILITIES AT THE NEVADA TEST SITE
CAPABILITY REQUIREMENT CURRENT CAPABILITY AT NTS
Transportation- Site Access E - Access by air via Las Vegas, NV. Desert Rock Airport on NTS has
7500' long x 100' wide runway capable of accepting jet aircraft.
- Access by road is four lane, divided US95 which intersects I15.Nature on-site road infrastructure. Adequate for construction.
- Access by rail via rail head at Las Vegas with Union Pacificline to NTS.
- Unloading & Inspection U - Management and stafC experience in receiving nuclear weaponscomponents is applicable to upgrading of capability needed for
receiving Pu in forms for processing
- On-Site Transport E -Nanagement and staff experienced in on-site transport of SNM- No public roads on NTS
- Packaging for Shipment E - Management and staff experienced in packaging nuclear componentsand wastes for shipment off-site
Supporting Site Assets- DOE Site office U - Experienced in directing nuclear weapons testing activities.
Applicable to extension of capability needed for reactor and
Pu processing activities.- Plutonium considered to be a resource
J
ho
- site Contractor Management U - Operational experience managing nuclear weapons testing activities.Applicable to extension of capability needed for reactor
and Pu processing activities.
- Work Force U - Trained, experienced, educated industrial base could supply skilled
professional, technical, craftspersons. Expansion needed for
Pu processing and reactor operation.
- NTS has -3500 employees
- Utilities E - "9E6 gpd water available at NTS from 14 existing wells.
-2E6 gpd beyond current usage available from Area 6 System for
proposed Complex 21 site. Additional wells can be drilled.
- Two independent 138 kV transmissions lines providing 25-30 NW peak
load, with 10-15 MW to be added within next few years.
Table 2 (Con't) PLUTONIUM DISPOSITION COMPLEX DEPLOYMENT CAPABILITIES AT THE NEVADA TEST SITE
CAPABILITY REQUIREMENT CURRENT CAPABILITY AT NTS
Supporting Site Assets (Con't)
- Technology Development E - Nevada Research & Development Area (NRDA} experienced with technology
development for nuclear rocket program. ENAD facility located inthis area.
- Quality Assurance E - Experienced in working to DOE quality assurance requirements
- Safety and Health E - Management and staff experienced in meeting or exceeding all health
and safety requirements associated with nuclear programs
- Environmental Protection E - Management and staff experienced in protection of the environRent.
- Community Support E - Long-standing support from local cummunities for DOE defense prograls
- Congressional delegation, governor, and state, regional, and local
officials publicly support continued defense-related projects at NTS
- Nevada Test Site Contractors Association supports now programs at NTS
- Political opposition to plutonium storage at NTS. Connected with
opposition to disposal of high level waste at Yucca Mountain.
!
file ODEPLOY.
DRAFT 15Jan94 **************************** DRAFT TO BE REVIEWED WITH ORR ***********************
Table 3 PLUTONIUM DISPOSITION COMPLEX DEPLOYMENT CAPABILITIES AT THE OAK RIDGE RESERVATION
G = Greenfield. New Capabilities Required.
U = Upgrade of Existing Capabilities Possible.
E = Existing or Planned Capabilities Meet Requirements.
CAPABILITY REQUIREMENT CURRENT CAPABILITY AT ORR
Overall Site Qualification E - Has served for -50 years as DOE site for energy R&D and weapons uranaumand litiium production. Current missions are energy R_D, weapons
disaantling, and uranium and lithium storage, waste management,
and environmental restoration.
- Currently operating, with -[TBD] employees
- occupies -54 square miles "17 miles northwest of Knoxville, TN
- Three technology areas: Xl0 (Oak Ridge National Laboratory}, Yl2, K25
Pu Feed Material Interface
(IF w/Nuclear Weapons Complex Reconfig Program}
- Pu Receiving and Storage E - Management and staff experienced with SNM handling including Pu
(within weapons program} - Pu could be stored in [TBD] existing facilities
' U/G - Y12 facilities available for upgrading for Pu-to-PuO2 conversion.- Pu-to-PuO2 Conversion
t_ (within weapons program} New process equipment needed.- Also greenfield available within 10,000 acres proposed for Complex 21 site
just West of YI2
- Management and staff have lab scale Pu-to-PuO2 conversion experience [TBR]
- PuO2 Interface Management E - Extensive experience at ORR in management/handling/transport/accountabilitof enriched uranium, and also Pu and PuO2 [TBR], is applicable to
management of interface with supplier of PuO2 feed material whether ornot Pu-to-PuO2 conversion function is located at ORR
MOX Fuel Fabrication
- MOX Pellet and Pin Fabrication U/G - Y12 facilities available for upgrading and new process equipmentfor MOX blending, pellet mint.ring, and pin loading operations
- Also greenfield available within proposed Complex 21 site
Management and staff have pilot scale experience producing MOX powder
using sol-gel and gel sphere processes
- ABWR MOX Assembly Fabrication U - Existing Y12 facilities adaptable for MOX assembly fabrication [TBR]Could also be included within MOX pellet and pin fabrication facility
located at either Y12 or Complex 21 site
- Fresh MOX Assembly Storage U - Existing YI2 facilities adaptable for fresh MOX assembly storage [TBR]Could also be included within MOX pellet and pin fabrication facility
located at either YI2 or Complex 21 site
Table 3 (ton't) PLUTONIUM DISPOSITION COMPLEX DEPLOYMENT CAPABILITIES AT THE OAK RIDGE RESERVATION
CAPABILITY REQUIREMENT CURRENT CAPABILITY AT eRR
Tritium Target Fabrication Interface
(IF w/Nucxlear Weapons Complex Reconfig Program)
- ABWR Tritium Target Fabrication U - Existing YI2 facilities adaptable with minimal refurbishing [TBR].
(within weapons program) New equipment for pin loading and closure needed.
- Could also be easily included within MOX pellet fabrication facility
located at either YI2 or Complex 21 site
- Management and staff experience fabricating fuel pins is applicable
to assembly of tritium targets using vendor supplied components [TBR]
- Tritium Target Interface Management E -- eRR has experience receiving and preparing tritium targets for
shipment to SRS under the production reactor program.
- Management and staff experienced with lithium handling, and with
coordination of procurement =tom commercial vendors, is applicable
to management of interface _°ch supplier of tritium targets whether
or not tritium targets are fabricated on-site
Reactor Plant Site Qualification
- Reactor Site E -- ABWR plant could be located within proposed Complex 21 site
- Former Clinch River Breeder Reactor (CRBR) site, owned by TVA,
is adjacent to eRR and could be utilized for ABWR
m - Site Qualification E - Data already compiled for Complex 21 site evaluation is applicable
to ABWR plant site qualification
- TVA site was characterized for CRBR plant
- Reactor Plant G - The ABWR plant will be a new facility
- Reactor Plant Operation E - err management and staff have operated several reactors. The new
reactor "Advanced Neutron Source" is planned by DOE at eRR.
Power to the Grid Interface
(IF w/IPP/Utility)
- Proximity of Commercial Grid E - eRR Central Control Facility at K25, formerly used as input
station for >2000 NWe from TVA, could be used as output stationfor 1350 HWe ABWR
- Reactor-to-Commercial Grid Transmission U - eRR has three primary 161 kV substations and eight primary
161 kV transmissions lines criss-crossing the site which could
be upgraded link with the 1350 HWe ABWR plant [TBR]
- Electric Power Sales U - Baseload power could be wheeled through the TVA grid for sale
to a nearby IPP/Utility. Agreements needed.
- The potential value of the baseload power is [TBD]
- IPP/Utility Interface Management E - Experience managing large amounts of electric power provided
to eRR in the past is applicable to management of the interface
with IPP/Utilities for power provided from eRR
Table 3 (Con't} PLUTONIUM DISPOSITION COMPLEX DEPLOYMENT CAPABILITIES AT THE OAK RIDGE RESERVATION
CAPABILITY REQUIREMENT CURRENT CAPABILITY AT ORR
Spent Fuel Storage Interface
(IF w/Nuclear Waste Disposal Program)
- Spent Fuel Storage E - Facilities at ORR for temporary storage, pending ultimate disposal,
(within waste disposal program) of spent fuel from the ABWR Pu Disposition Complex include _TBDJinclude [TBD]
- Spent Fuel Transport E - Management and staff experieced in transport of spent fuel
(within waste disposal program)
- Spen_ Fuel Interface Management E - Experience managing spent fuel from reactors at ORR is applicableto management of interface with US Spent Fuel Disposal Programwhether or not the spent fuel is stored at ORR
Waste Management
- High Level Waste (HLW) E - Management and staff experienced in handling HLW in the form ofirradiated reactor components such as control rods and core
structural elements |TBR]
- HLW is stored for future disposal at the [TBD] facilities at ORNL
- Transuranic Waste (TRU} E - Management and staff experienced in handling TRU- Solid transuranic waste is stored for future treatment and disposal
t_ at the [TBD] facility at ORNL. Solid Pu scrap from the NFS plant
t_ is also being transferred to ORNL. A new facility at ORNL to treat
and package solid transuranic waste to specifications for ultimate
disposal at WIPP has been proposed to DOE.
- Liquid and sludge transuranic waste is stored at the [TBD] storage
tank facility at ORNL. A new facility at ORNL to treat 3-600,000
gallons of transuranic liquid/sludge to specifications for ultimate
disposal at WIPP has been proposed to DOE.
- Low Level Waste (LLW) E - Management and staff experienced in handling LLW
- Hazardous Waste E - Management and staff experienced in handling hazardous wastes
- Mixed Waste E - Managemet and staff are experienced in handling mixed wastes- ORR has the only licensed operating Toxic Substances Control
Act (TSCA} incinerator in the US to handle mixed waste.
currently handling liguid. Capable of handling solid.
- Nuclear Waste Interface Management E - Experience managing all forms of nuclear waste is applicable tomanagement of interface with Nuclear Waste Disposal Program
Table 3 (Con't) PLUTONIUM DISPOSITION COMPLEX DEPLOYMENT CAPABILITIES AT THE OAK RIDGE RESERVATION
CAPABILITY R_UIREMENT CURRENT CAPABILITY AT ORR
Tritium Recovery Interface
(IF w/Nuclear Weapons Compl _x Reconfig Program)
- Extraction G - A new facility is required
(within weapons program)
- Purification G - A new facility is required for purifying the tritium to specifications(withi_ weapons program) for shipment to the RTF at SRS
- Recovery G - A new facility is required at ORR. It is believed the Replacement
(within weapons program) Tritium Facility at SRS has been designated by DOE for tritium recovery
for the weapons program
- Tritium Recovery Interface Management E - The proposed Advanced Neutron Source reactor at ORR will
include a detriiation facility
- Experience managing in-pile experiments and tritium operations at ORR
is applicable to management of tritium recovery interface with the
Nuclear Weapons Complex Reconfiguration Program whether or not some or
all of the tritium recovery facilities are located at ORR
Safeguards _ Security
- Accountability E - Management and staff experienced in implementing strict
accountability procedures for storage, handling, and transport, of special nuclear materials in all forms including metal,
oxide components, and wasteO_
- Protection U - ORR has an existing security force, with experience in PIDAS zones,
which is applicable to providing additional personnel, fences, and
guard posts for protection of Pu Disposition Complex
Safety and Environmental Approval
- Safety Approval U - ORR has experience with the design requirements, safety analyses,
documents, safety reviews and audits, operationaA readiness reviews,
conduct of operations, Price-Anderson ammendmont rules, configuration
management, and quality assurance to obtain safety approval by DOEfor nuclear facilities including reactors. Since the DOE and NRC
requirements are similar, this capability at ORR can be upgraded toaccomodate NRC procedures.
- Environmental Approval U - ORR has experience with the environmental approval procedures which
result in the Record of Decision by DOE. These procedures are still
evolving, and it is anticipated that this capability at ORR will
need to be upgraded for new reactors and Pu processing facilities.
- Federal/State/Local Permitting E - ORR has experience with permitting at all levels
Table 3 (ton't) PLUTONIUM DISPOSITION COMPLEX DEPLOYMENT CAPABILITIES AT THE OAK RIDGE RESERVATION
CAPABILITY REQUIREMENT CURRENT CAPABILITY AT ORR
Transportation
- Site Access E - ORR has access to the site by road, rail, and barge which can
meet construction needs of the Pu Disposition complex
- Unloading • Inspection E - Management and staff experienced in receiving SNM in all forms
- On-Site Transport E - Management and staff experienced in on-site transport within
and between protected zones
- Packaging for Shipment E - Management and staff experienced in shipping SNN in all forms
Supporting Site Assets
- DOE Site Office E - Experience in directing nuclear prograEs and largo projects
is applicable to Pu Disposition Complex
- Site Conttractor Management E - Operational experience in management of plutonium and tritium
processing and technology and large projects including reactors
is applicable to Pu Disposition Complex
- Site contractor management has indicated strong interest in
the Pu Disposition Complex and believe ORR could accomodate
o the program
- Work Force E - Technical and production oriented work force can support large
plutonium and tritium construction and operation project
- utilities E - Adequate water for reactor makeup and complex operations
- Adequate electric power and other utilities to support
major construction project
- Technology Development E - ORNL is multipurpose research laboratory in energy related areas
- Quality Assurance E - Experienced in working to DOE quality assurance requirements
- Safety and Health E - Management and staff experienced in Beeting or exceeding all health
and safety requirements associated with nuclear programs
- Environmental Protection E - Management and staff experienced in protection of the environment.
Environmental restoration programs are operational at ORR.
- Conmunity Support E - Indications are that state and local officials will be receptive
to the handling of plutonium at ORR, based on recent public
hearings for the Complex 21 Program Environmental Impact Statement
file PDEPLOY.DRAFT 15Jan94
******************** DRAFT TO BE REVIEWED WITH PANTEX **********************
Table 4 PLUTONIUM DISPOSITION COMPLEX DEPLOYMENT CAPABILITIES AT THE PANTEX PLANT
G = Greenfield. New Capabilities Required.
U = Upgrade of Existing Capabilities Possible.
E = Existing or Planned Capabilities Meet Requirements.
CAPABILITY REQUIREMENT CURRENT CAPABILITY AT PANTEX
Overall Site Qualification E - Currently operating. Has served for >40 years as DOE site for assemblyand disassembly of nuclear weapons. Current missions are if) fabricate
chemical high explosive components for nuclear weapons, (2) ass®ible
nuclear weapons for the nation's stockpile, {3) maintain and evaluate
nuclear weapons in the stockpile, and (4) disassolble nuclear weapons
being retired from the stockpile. Pantex has also been designated asan interim storage site for Pu pits.
- Occupies "16,000 acres -17 miles northeast of Amarillo, TX
PuO2 Feed Material Interface
(IF w/Nuclear Weapons Complex Reconfig Program)
!
tO - Pu Receiving and Storage G Handling of Pu in metallic form for processing would he new capability
Oo (within weapons program)
- Pu-to-PuO2 Conversion G - Management and staff experienced in handling weapons parts containing
(within weapons program) nuclear materials. Nuclear materials have not been processed in thepast, and this would be new capability for Pantex.
- PuO2 Feed Interface Management U - Experience managing nuclear weapons components could be extended tocapability for management of interface with PuO2 supplier
Mox Fuel Fabrication
- MOX Pellet and Pin Fabrication G - New capability needed for Pu processing for MOX fuel fabrication.- Pantex has experience with process robotics, including 70% automation
of glovebox operations and 100% automation of handling of the pit
in the Pit Reuse Program
- ABWR MOX Assembly Fabrication G - MOX assemblies could be fabricated in greenfield facilitybuilt for MOX pellet and pin fabrication
- Fresh MOX Assembly Storage G - Fresh MOX assemblies could be stored in greenfield facilitybuilt for MOX pellet and pin fabrication
Tritium Target Fabrication Interface
(IF w/Nuclear Weapons Complex Reconfig Prog)
- ABWR Tritium Target Fabrication G - Target rods could be fabricated using components from couercial
(within weapons program) vendors in greenfield facility built for MOX pellet and pin fabrication
- Tritium Target Fab Interface Management E - Experience managing procurement from commercial vendors is applicableto management of interface with supplier of tritium targets whether
or not ABWR tritium targets are fabricated on-site
Table 4 (Con't) PLUTONIUM DISPOSITION COMPLEX DEPLOYMENT CAPABILITIES AT THE PANTEX PLANT
CAPABILITY REQUIREMENT CURRENT CAPABILITY AT PANTEX
Reactor Plant Site Qualification
- Reactor Site E - Land is available for a reactor. Specific potential sites have
not yet been selected.
- Site Qualification U - Characterization of Pantex site for nuclear weapons activities is
applicable towards qualification as an operating reactor site.
Has not participated in NPR or other reactor siting evaluations.
- Reactor Plant G - The ABWR plant will be a new facility
- Reactor Plant Operation G - Reactors have not been built at Pantex. Reactor operation willbe a new function.
Power to the Grid Interface
(IF w/IPP/Utility)
- Proximity to Commercial Grid [TBDJ
- Reactor-to-Commercial Grid Transmission |TBDJ
t_ - Electric Power Sales [TBD]!
- IPP/Utility Interface Management |TBD]
Spent Fuel Storage Interface
(IF w/Nuclear Waste Disposal Program)
- Spent Fuel Storage G - New capability needed for storage of spent fuel at Pantex
(within waste disposal program_
- Spent Fuel Transport G - New capability needed for transport of spent fuel
(within waste disposal program)
- Spent Fuel Interface Management U - Expansion of capability needed for management of interface with
Nuclear Waste Disposal Program whether or not spent fuel is
stored or transported by Pantex
Table 4 (ton't) PLUTONIUM DISPOSITION COMPLEX DEPLOYMENT CAPABILITIES AT THE PANTEX PLANT
CAPABILITY REQUIREMENT CURRENT CAPABILITY AT PANTEX
Waste Management
- High Level Waste (HLW) G - HLW not currently generated or disposed of on-site
- Transuranic Waste (TRU) G - TRU not currently generated or disposed of on-site
- Low Level Waste (LLW) G - LLW not currently generated or disposed of on-site
- Hazardous Waste E - Management and staff experienced in handling hazardous wastes- Other than burning of high exploseve materials from disassembled
weapons, hazardous wastes not currently disposed of on-site
- Mixed Waste G - Mixed wastes not currently generated or disposed of on-site
- Nuclear Waste Interface Management U - Experience of Manager, Amarillo Area Office in TransuranicWaste Program is applicable towards expansion of capability
for management of interface of Pu processing and reactor waste
functions with Nuclear Waste Disposal Program
Tritium Recovezy InterfaceI
t_ (IF w/Nuclear Weapons Complex Reconfig Prog}
- Ext action G - New capability required
(within weapons program)
- Purification G - New capability required
(within weapons program)
- Reservoir Loading G - New capability required
(within weapons program}
- Tritium Recovery Interface Management U - Similar to spent fuel. Extension of capability needed for managementof tritium recovery i_terface, whether or not some or all of the
tritium recovery full. ions are located at Pantex
Safeguards & Security
- Accountability U - Management and staff experienced in strict accountability of nuclearweapons components. Safeguards & Security Directorate includes nuclearmaterial control, nuclear material accounting, and enhancement for
protection of SNN. Also security clearance and classified document
control. Expansion of capability needed for accountability of Pu in
forms for processing.
- Protection U - Pantex experience in protection of nuclear weapons facilities isapplicable to providing additional personnel, fences, and guard posts
for Pu Disposition Complex
- Safeguards & Security Directorate includes security force program.
Experienced with PIDAS zones.
Table 4 (Con't) PLUTONIUM DISPOSITION COMPLEX DEPLOYMENT CAPABILITIES AT THE PANTEX PLANT
CAPABILIT_ REQUIREMENT CURRENT CAPABILITY AT PANTEX
Safety and Environmental Approval
- Safety Approval U - Management and staff experience in safety approval by DOE for weaponsassenbly/disassembly facilities and operations is applicable to
extension of capability and procedures needed for DOE and NRC
requirements for Pu processing and reactor safety approval
- Environmental Approval U - Management and staff experienced in environmental permitting for weaponsassembly/disassembly facilities and operations is applicable to
extension of capability and procedures needed to include requirements
for Pu processing and reactor environmental approval
- Environmental Assessment for storage of pits without HE in progress
- FederalStateLocal Pernitting E - Experienced in permitting at all levels
Transportation
- Site Access E - Access by air via Amarillo, TX- Access by road via US-60, which intersects with 1-40 in Amarillo.
47 miles of paved roads on-site.
- Access by rail is [TBD]. 17 miles of railroad track on-site.
- Unloading & Inspection U - Management and staff experience in receiving nuclear weapons!
t_ components is applicable to extension of capability needed forreceivin_ Pu in forms for processing
- On-Site Transport E - Nanagement and staff experience in on-site transport of nuclearweapons components is applicable to transport of packaged Pu
- No public roads on Pantex site
- Packaging for Shipment U - Management and staff experience in packaging nuclear weaponscomponents for shipment off-site is applicable to extension of
capability needed for shipment of Pu and wastes in processing forms
- Experience includes use of Special Secure Transporter (SST) vehicles.
SST operation is directed from Pantex.
Supporting Site Assets- DOE Site Office E/U - Albuquerque Operations Office experience directing nuclear programs
and large projects is applicable to Pu Disposition Complex
- Amarillo Area Office experience in direction of nuclear weapons
assembly/disassembly related programs is applicable to extension
of capability needed for reactor operation and Pu processing
- Albuquerque Operations Office and Amarillo Area Office consider
plutonium to be a national resource
- Site Contractor Management U - Operational experience managing nuclear weapons assembly/disasselblyis applicable to extension of capability needed for reactor and
Pu processing
- Conmitted to growth by infusing new technologies and broadened base
of technical personnel into the Pantex Plant
Table 4 (Con't) PLUTONIUM DISPOSITION COMPLEX DEPLOYMENT CAPABILITIES AT THE PANTEX PLANT
CAPABILITY REQUIREMENT CURRENT CAPABILITY AT PANTEX
- Work Force U - "3000 employees. 29% with bachelor's degree or above. >50% ofPantex work force are technically skilled oxployees. But not
experienced with Pu processing and reactor operations.
- Utilities E - Adequate water available [TBR]- Adequate electric power for large construction project [TBR]
- Technology Development U - Advanced Technology Office (ATe) provides planning and implementationof new technologies from Battelle, DOE labs, industry, and universities
for activities related to nuclear weapons assembly and disassembly.
Expansion of this capability needed to support Pu processing and
reactor operation.
- Tester Design Engineering Department provides design, development,
modification, and maintenance of automated electronic measurement
systems for testing of nuclear weapons electrical ciruitry
- High Explosives Synthesis Facility develops new chemical processes
for explosives
- Other capabilities at Pantex applicable to technology development
include the Analytical Laboratory, Nondestructive Evaluation (XRay)
Department, Gas Analysis Lab, Explosives Test Site, and others
- Quality Assurance E - Experienced in working to DOE quality assurance requirements
- Safety and Health U - Management and staff experienced in meeting or exceeding all health, and safety requirements associated with nuclear programs
t_ - Experience in safety for weapons assembly/disassembly facilities
and operations is applicable to extension of capability needed
for Pu processing and reactor operation such as criticality
alarm and Pu detection and containment
- Environmental Protection U - Management and staff experienced in protection of environment.Extension of capability needed for Pu prccessing and reactor operation.
- Community Support U - Community Relations Department activities include Information andAwareness, Community Involvement, Employee Communications, and
Educational Development
- Local community leaders in Panhandle 2000 support continuing
current nuclear weapons activities at Pantex
- Pantex is largest employer in Amarillo area
- State political leaders concerned about long term storage of pits
and potential for contamination of Ogalalla aquifer by Pu processingactivities
- Ranching and farming interests concerned about any adverse impact
on source of water
- University of Texas, Austin, Department of Economic Geology conducting
study of geologic structure to conduct water and potential for
contamination. Considers plutioium a resource.
- Public interest groups active around Pantex include Panhandle Area
Neighbors and Landowners (PANAL), Serious Texans Against Nuclear
Dumping (STAND}, Save Texas Agricultural Resources (STAR), andPeace Farm
DRAFT 15Jan94 file SDEPLOY.
************************* DRAFT TO BE REVIEWED WITH SRS *************************
Table 5 PLUTONIUM DISPOSITION COMPLEX DEPLOYMENT CAPABILITIES AT THE SAVANNAH RIVER SITE
G = Greenfield. New Capabilities Required.
U = Upgrade of Existing Capabilities Possible.
E = Existing or Planned Capabilities Meet Requirements.
CAPABILITY REQUIREMENT CURRENT CAPABILITY AT SRS
Overall Site Qualification E - Has served for >40 years as a DOE site for production and processingof plutonium and tritium. Current missions include processing and
storage of fissile materials, tritium processing, waste management,and environmental restoration.
- Currently operating, with -[TBD] employees
- Occupies -300 square miles adjoining the Savannah River -13 miles
southeast of Augusta, GA and -12 miles south of Aiken, SC
PuO2 Feed Material Interface
(IF w/Nuclear Weapons Complex Reconfig Program)
- Pu Recieving and Storage E - Plutonium Storage Facility (PSF} in SRS Separations Area was
(within weapons program) built to receive, store, monitor, retrieve, and sh_p packaged Pu,
and is adaptable to reveiving and storage of Pu metal from a weapons
site for conversion to PuO2 for Pu disposition
- Pu-to-PuO2 Conversion E - New Special Recovery Facility (NSR) in SRS Separations Area was
o (within weapons program) built to convert impure Pu metal or oxide from scrap to pure
t_ Pu nitrate solution, and is adaptable to producing PuO2 to spec
with some refurbishing
- PuO2 Feed Interface Management E - Experience in management/handlingtransport/accountability of
plutonium is applicable to management of interface with supplierof PuC2 feed material whether or not Pu-to-PuO2 conversion is
located at SRS
MOX Fuel Fabrication
- MOX Pellet and Pin Fabrication U/G -- Waste Tank Equipment Gallery (WETG} at Barnwell site can be
upgraded to house multi-story, automated fabrication facility.
New equipment for powder blending and sintering and pin
i_ading required.
- Greenfield sites also available well within SRS boundaries
- Management and staff experienced in Pu processing
- SRS has process automation development capability applicable
to design of MOX fabrication facility
- ABWR MOX Assembly Fabrication U - Could be co-located with either pellet/pin fabrication or fresh
assembly storage. New equipment for pin loading and closure needed.
- Fresh _OX Assembly Storage E - Main Processing Facility at Barnwell site has space for
temporary storage of assemblies until transport to reactor
- Storage space available at existing, not-currently-operating
reactors could meet requirements with some refurbishing
Table 5 (Con't) PLUTONIUM DISPOSITION COMPLEX DEPLOYMENT CAPABILITIES AT THE SAVANNAH RIVER SITE
CAPABILITY REQUIREMENT CURRENT CAPABILITY AT SRS
Tritium Target Fabrication Interface
(IF w/Nuclear Weapons Complex Reconfig Program)
- ABWR Tritium Target Fabrication E - Production Reactor Fuel and Tritium Target Fabrication Facility in300M Area adaptable for assembly from vendor supplied components
(within weapons program) with some refurbishing
- Could also be located within MOX fabrication facility
- Tritium Target Fab Interface Management E -- Experience managing fabrication of tritium targets for productionreactors and in procurement from commercial vendors is applicable
to management of interface with supplier of tritium targets whether
or not ABWR tritium targets are fabricated on-site
Reactor Plant Site Qualification
- Reactor Site E - ABWR plant could be located on site selected for NPR- Alternate NPR sites at SRS also available
- Site also available at Vogle Station, owned by Georgia
Power, just across Savannah River from SRS
- Site Qualification E - Site data already compiled for NPR and Complex 21 evaluationsis applicable to ABWR plant site qualification
- Site at Vogle Station is qualified for reactor
- Reactor Plant G - The ABWR plant will be a new facility!
t_
- Reactor Plant Operation E - Management and staff experienced in reactor operation.K Reactor, last of production reactors, operational until 1992.
Currently in cold standby readiness to restart if directed.
Power to the Grid Interface
(IF w/IPP/Utility)
- Proxlmity of Commercial Grid E - Commercial grid available through current connection to SCE&G lines- Commercial grid also available at Vogle Station, owned by Georgia Power,
just across Savannah River from SRS
- Reactor-to-Commercial Grid Transmission U - Existing high voltage transmission lines crossing SRS need to beupgraded to link with ABWR plant site
- Electric Power Sales U - Power could be brokered through existing DOE organization,Southeastern Electric Power Authority, but changes in legislation
and customer base required to sell power from nuclear plant
to customers with baseload needs.
- Power could he wheeled through SCEaG grid or from Vogle Station
through commercial grid for sale to a nearby IPP/Utility.
- At present, SRS power interface is with SCE&G, not Georgia Power
- The potential value of the baseload power is [TBD]
- Power sales agreements needed
- IPP/Utility Interface Management E - Experience in managing large amounts of electric power providedto SRS in the past is applicable to management of interface
with the IPP/Utilities for power provided from SRS
Table 5 (ton't) PLUTONIUM DISPOSITION COMPLEX DEPLOYMENT CAPABILITIES AT THE SAVANNAH RIVER SITE
CAPABILITY REQUIREMENT CURRENT CAPABILITY AT SRS
Spent Fuel Storage Interface
(IF w/Nuclear Waste Disposal Program)
- Spent Fuel Storage E - P-, K-, and L-Reactor pools now being used to store U/A1 alloy fuel
(within waste disposal program} from production reactors. A proposed plan is to process this fuel
in F-Canyon. Should take -3 years once started. Then pools could
be available for spent ABWR MOX fuel.
- C-Reactor pool currently filled with water and empty of fuel.
Can be used for spent fuel with minimal upgrading to meet
water purification system requirements
- R-Reactor pool is dry and empty
- Main Processing Facility at Barnwell site has large pools
currently dry and eRpty and available. Barnwell County Council
has Economic Development Initiative for "Center for Nuclear
Materials Management", which includes temporary storage of
spent commercial LWR fuel.
- Spent Fuel Transport E - Management and staff experienced in spent fuel transport
(within waste disposal program)
- Spent Fuel Interface Management E - Experience managing spent fuel from production reactors is
applicable to management of interface with Nuclear Waste Disposal• Program whether or not spent fuel stored at SRS
!
t_ Waste Managementt_
- High Level Waste (HLW) E - Management and staff experienced in handling high level waste
(including transuranics} in the forms of solid and liquid from
irradiated fuel reprocessing and irradiated reactor components suchas control rods and core structural eleRents
- Liquid/sludge high level wastes (including transuranics} from
reprocessing operations stored at tank farms (241-F and 241-H)
- Construction of Defense Waste Processing Facility (DWPF} is
complete and currently in cold test mode. Will vitrify high
level liquid/sludge waste (including transuranics} into
glass logs, to be stored on-site until transport to Yucca
Mountain for disposal.
- Transuranic Waste (TRU} E - Management and staff experienced in handling transuranic wastes
- Solid transuranic wastes stored at Solid Waste Disposal Facility
- Liquid/sludge transuranic wastes considered to be high levelwaste and treated as noted above
- Low Level Waste (LLW) E - Management and staff experienced in handling low level wastes
- SRS has Solid Waste Disposal Facility for solid waste including
used equipment
Table 5 (Con't) PLUTONIUM DISPOSITION COMPLEX DEPLOYMENT CAPABILITIES AT THE SAVANNAH RIVER SITE
CAPABILITY REQUIREMENT CURRENT CAPABILITY AT SRS
Waste Management (Con't)
- Hazardous Waste E - Management and staff experienced in handling hazardous wastes
- Mixed Waste E - Management and staff experienced in handling mixed wastes
- SRS has broken ground fen Consolidated Incineration Facility to
dispose of burnable wastes from all site operations
- Nuclear Waste Interface Management E - Experience managing all forms of nuclear waste is applicable to
management of interface with Nuclear Waste Disposal Program
Tritium Recovery Interface
(IF w/Nuclear Weapons complex Reconfig Program)
- Extraction U - Existing facility for tritium extraction from _ tritium
(within weapons program} targets needs upgrading for extraction from ABWR targets
- Purification E/U - Existing tritium purification capability could moot
(within weapons program) requirements, but will soon need replacement duo to age.
Upgrade of Tritium Extraction/Separation Facility (232-H}
has been proposed to DOE.
- Reservoir Loading E - Replacement Tritium Facility (RTF) is now facility!
t_ (within weapons program)
- Tritium Recovery Interface Management E - Experience managing tritium operations is applicable to
managing the tritium recovery interface whether or not some
or all of the tritium recovery facilities are located at SRS
Table 5 (ton't) PLUTONIUM DISPOSITION COMPLEX DEPLOYMENT CAPABILITIES AT THE SAVANNAH RIVER SITE
CAPABILITY REQUIREMENT CURRENT CAPABILITY AT SRS
Safeguards & Security
- Accountability E - Management and staff experienced in implementing strict accountability
procedures for storage, handling, processing, and transport of special
nuclear materials in all forms including metal, oxide, componentsand waste
- Protection U - SRS has existing security personnel, with experience in PIDAS zones,
which can provide protection for Pu disposition facilities with
some upgrades for additional personnel, fences, and guard posts
depending on actual location on-site
Safety and Environmental Approval
- Safety Approval U - Management and staff experienced with requirements, analyses,
documents, and reviews required to obtain safety approval by DOE
for nuclear facilities including reactors. Since DOE and NRC
requirements are similar, this capability can be upgraded to
accomodate NRC requirements.
- Safety approval experience for K-Reactor and NPR programs
applicable to Pu Disposition Complex
- Environmental Approval E - Management and staff experienced with environmental approval
requiremments to obtain Record of Decision by DOE with EPA, approval for nuclear facilities including reactors.
t_ - SRS has established approach of detailed procedures for evaluations,
_J documents, and reviews for each stage of approval processes
- Site data developed for NPR and Complex 21 evaluations applicable
to Pu Disposition Complex.
- Federal/State/Local Permitting E - SRS has pro-active approach to permitting at all levels
Transportation
- Site Access E - SRS has access to site by road, rail, and barge docking facilities,
which can meet construction needs of Pu Disposition Project with
minor upgrading. Savannah River can be made navigable for barge
transport of heavy equipment.
- Unloading & Inspection E - Management and staff experienced in receiving SHlq in all forms
- On-Site Transport E - Management and staff experienced in on-site transport of SNM
in all forms within and between protected zones- Public road crosses SRS
- Packaging for Shipment E - Management and staff experienced in shipping SNlq in all forms
Table 5 (Con't) PLUTONIUM DISPOSITION COMPLEX DEPLOYMENT CAPABILITIES AT THE SAVANNAH RIVER SITE
CAPABILITY REQUIREMENT CURRENT CAPABILITY AT SRS
Supporting Site Assets- DOE Site office E - Experienced in directing nuclear programs and large projects
- Have sponsored tours and presentations to expedite independent
review and evaluation of SRS capabilities for Pu disposition
by GE team
- Have expressed interest in status of Pu Disposition Study
independent of site and reactor type
- Site Contractor Management E - Operational experience managing plutonium and tritium processingand technology and largo projects including production reactors
- Site contractor management has indicated strong interest in
the Pu Disposition Complex and believes SRS could accolodate
the program
- Work Force E - Technical and production oriented work force can support largeplutonium and tritium processing and reactor construction and
operation project
- Utilities E - Adequate water supply and distribution system for reactor makeupand complex operations. Three pumping stations on Savannah River
can deliver 3E6 gpm to SRS.!
t_ - Adequate electric power to support large construction projectOo
- Technology Development E - Savannah River Technology Center (SRTC) at SRS devoted to solvingproduction problems and improving processes of nuclear materials
handling, storage, processing, reactor operation, and waste management
- Quality Assurance E - Experienced in working to DOE quality assurance requirements
- Safety and Health E - Management and staff experienced in meeting or exceeding all healthand safety requirements associated with nuclear programs
- Environmental Protection E - Management and staff experienced in protection of the environment.Environmental restoration programs operational at SRS.
- Community Support E - SRS management believes indications are that state and local officialswill be receptive to new Pu processing and reactor operations of
Pu Disposition Complex, based on response to NPR and Complex 21
proposals- SRS management has found widespread local community support for
operations at the site. Active program of communications with
community leaders and public is maintained.
- Barnvell County Economic Development Initiative includes proposal
for "Center for Nuclear Materials Management" which is to be a
private, commercial engineering laboratory working on ways to
manage waste
- Energy Research Foundation (ERF), located in Col_L_bia SC, is loading
public interest group. Concerns include long torn storage of
nuclear materials and release of tritium contaminated waste.
4.4.1 TRANSPORTATION LOGISTICS FORTRITIUM PRODUCTION
4.4.1.1 Summary
Tritium production tnvolves transportation of enrtched ltthtum andtritium, both
of which are classified as "Other Nuclear HaterJal" and require appropriate
safeguards, security and accountability measures. In addition, the irradiated
target rods must be shippedto the extraction facility and the spent target rods
(after tritium extraction) shipped to the waste dtsposal site. The results of
this evaluation confirmed that the existing and/or planned transportation
infrastructureshouldbe adequatetomeetthetritiumproductionneeds.Further,
relativelyfew shipmentswill be required.Thus,transportationlogisticsare
not likelyto be a controllingconsiderationin sitingthe tritiumproduction
facilities.
4.4.1.2 Background
The purposeof thistaskwas to assessthe transportationlogisticsassociated
with tritiumproductionin the ABWR in the contextof applicablerequirements,
the existingand planned infrastructureand the deploymentoptions under
consideration.The scopeof the assessmentincludedtransportationrelatedto
targetfabrication,shipmentof freshand irradiatedtargetrods,the tritium
productand the spenttargetrodsafterextractionof the tritium.
As discussedin previoussectionsof thisreportthe referenceplanfortritium
productionin the ABWRconsistsof:
• Commercialfabricationof alltargetrodcomponentsincludingtheenriched
LiAl02pellets. The natureand locationof thesefabricationfacilities
was purposelyleftopento maximizeDOE flexibility.,
• Assemblyof the targetrods in a fuel fabricationfacility,but not
necessarilythe MOX fuelfabricationplant.
4.4.1-I
* Loading the target rods in the fuel bundle at the MOXfuel plant or thereactor.
• Discharge of all target rods after a one cycle exposure.
• Removalof the irradiated target rods from the fuel bundle in the reactor
spent fuel pool for shipment to the extraction facility.
• Extractionof tritiumfromthe irradiatedtargetrods in a new facility
located adjacentto the existingSavannahRiver Site (SRS) tritium
facilities.
• Disposal of the spent target rods after extraction as LowLevel Waste at
a DOEor commercial disposal site.
Alternatives considered which could impact the transportation logistics for
tritium production include:
• Fabrication of the LiAI02 pellets and/or assemblyof the target rods in a
facilityor facilitiesco-locatedwiththe reactor.
• Irradiationof the targetrodsfor multiplecycles.
• Co-locationof the extractionfacilitywith the reactor.
• Disposalof the spenttargetrodsafterextractionas corecomponentsin
the federalrepositoryfor spentnuclearfuel.
4.4.1.3 Discussion
The key factorswhich influencethe transportationlogisticsfor tritium
productionare the numberof targetrods irradiatedper year, theirdesign,
enrichedlithiumand tritiumcontentandthe activitylevelof the spenttarget
4.4.1-2
rods. Another potentially important factor is the extent to which the tritium
facilities are co-located with each other or with other related facilities.
The reference ABWRcore design for tritium production contains 872 fuel
assemblies with four enriched ltthtum target rods per assembly. The target rods
have the same external dimensions as the ABWRfuel rods. Each fresh target rod
contains approximately 20 grams of Li e. The average trttium content of a target
rod at discharge after one cycle exposure is approximately 104 curies or about
I gram. The gammadose rate from a spent target rod, one year after discharge
is expected to be on the order of 100 R/hr at a foot.
(_ontro_and Accountabilitv
Both enriched lithium and tritium are classified as "Other Nuclear Material"
under DOE Order 5633.3A, Control and Accountability of Nuclear Material.
Enrichedlithiumis classifiedas CategoryIV,AttractivenessLevel E,which must
have material control and accountabilityfor quantitiesof I kg or mo-e of Lia
(-50 target rods). Since the reference plan involves irradiationof ~3,500
target rods per year the transportationof the enriched lithium for pellet
fabrication,the finishedLiAL02pelletsand the fresh target rods will require
appropriatecontrol,accountabilityand safeguards. In the event that shipment
of the LiAl02pelletsor finishedtargetrods is consideredto involveclassified
configurationor contentthe physical protectionprovisionsof DOE Order 5632.5
would also apply.
The spent target rods, which containabout 85% of their originalLie inventory,
are plannedto be transportedin bundlesof 49 rods. In this configurationthey
will exceed the 100 rem/hr at 3 feet criteriaof 10 CFR 67 and 73.60 and can be
consideredas self protecting.
Under DOE Order 5633.3Atritiummust havematerialcontroland accountabilityfor
quantiLiesof 0.01 gram or more (~100 curies). Quantitiesof tritium greater
than t. grams are classifieda Category Ill,which requiresadditionalcontrol,
accountabilityand safeguards. While individualshipping containersmay have
4.4.1-3
less than 50 grams, the total quantity in a shipment wtll undoubtedly exceed this
limit in order to keep the number of shipments to a reasonable level.
A spent target rod after extraction is expected to contain about 50 curies of
residual tritium. The target rod handling concept for shipping, extraction and
waste disposal is to keep them in a 7x7 bundle which has the same external
dimensions as a BWRfuel assembly. This maintains compatibility with existing
and planned transportation systems. Thus, a spent target waste package would
contain -2,500 curies of residual tritium. While itwt11 be necessary to account
for this residual tritium from a materiai balance standpoint, the waste package
_i11 still have a dose rate that exceeds the self protection criteria of 10 CFR
67 and 73.60.
Safeguards and security provisions for Category III and IV materials require
development of protection provisions to be contained within a site-specific
safeguardsand security plan and/or a Master Safeguardsand SecurityAgreement
(MSSAs). Provisionsrequiredto be in thisdocumentsare identifiedin DOE Order
5632.2A.
A Material Controland Accountability(MC&A)Plan will be also requiredfor the
enriched lithiumand tritium. The level of controland accountabilitywould be
consistent with the economic and strategic value of these materials. An
implementationguide for DOE Order 5633.3A has been prepared which describes
methods for meeting requirementsof this order. Since a MSSA and a MC&A plan
will also be required for the ABWR f,aeland it is expected that the control,
accountabilityand safeguardsrequirementsfor enrichedlithiumand tritiumwould
be incorporatedin these documents.
In additionto DOE Order 5632.2Aand 5633.3A,the provisionsof DOE Order 5633.4,
Nuclear Materials Transactions: Documentation and Reporting, and DOE Order
5633.5, Nuclear Materials Reporting and Data Submission Procedures,are also
applicable and provide additional detail relating to the requirements.
Requirementsfor the scope and contentof MC&A plans are to be determinedby the
Manager, DOE Field Office.
4.4.1-4
Shipmentof CategoryIII quantitiesof tritiumcan be made by SST, or by
government-ownedtruck, exclusive-usecommercialcarrieror rail with the
appropriatecontrols(see Section7 for details). Enrichedlithium,being
CategoryIV,canalsobe shippedbyDOEapprovedcommercialcarriersandinvolves
lesscontrolsthan CategoryIII quantitiesof tritium.
)hiDm_ntof EnrichedLithium.LiAlO2_Pelletsand FreshTaraetRods
DiscussionswithDOE staffat the SavannahRiverSite (SRS)indicatethatoff-
siteshipmentof enrichedlithiumhasbeenveryinfrequent.Shipmentsthathave
takenplacehave involvedshipmentof lithiumin oxideform. The materialhas
beenplacedin a containerand sealedin an inertatmosphere.The containerhas
then beenplacedinsidea metaldrum for shipmentby a commercialcarrierthat
is bondedand certifiedby DOE. Theywere unawareof any near-termchangesin
regulations.
The quantitiesof enrichedlithiumand LiAl02pelletsneededby the ABWR are
smallenoughthattheycouldeachbe accommodatedin a singleannualshipmentif
desired.Theunirradiatedtargetrodswouldbe shippedincontainerssimilarto
thosecurrentlyusedforfreshBWRfuel. Thiswouldrequire25 to 30 containers
to shipthe 3,500targetrodsneededeachyearwhichcouldbe accommodatedintwo
shipments.Becauseof the relativelyfewshipmentsinvolvedit isunlikelythat
transportationwill be an importantconsiderationin sitingthe LiAI02pellet
fabricationor targetrod assemblyoperations.
_hiDmentof Irradiated Target Rodsto the Extraction Facilit,y
As indicated in Section 4.4.7.2 the reference plan is to ship the irradiated
target rods to a newextraction facility at the SRSfor recovery of the tritium
product. An alternate considered was to perform the extraction and initial
purification steps in a facility co-located with the reactor andship the tritium
to SRSfor final processing and storage. In either case the target rods wouldbe removedfrom the fuel assemblies in the reactor spent fuel pool and loaded
4.4.1-5
into transport canisters. These transport containers would hold 49 target rodsand have the sameexternal dimensionsandconfiguration as a BWRfuel assembly.
There are a numberof commerciallyavailable, licensed spent fuel shipping casks
that can be used to transportthe irradiated target rods for e_ther on-site or
off-site shipment. Theserange from a 23 ton, two assembly, l(_gal-weight truckcask to a 70 ton, eighteen subassemblyrail cask. The other option is to design
a special purposecaskwhich could undoubtedlyaccommodatelarger target loadings
than a spent fuel cask becauseof the reduced decay heat and radiation levels.However, since commercial cask shipment is relatively inexpensive (four rail
shipments per year at $4k per day) the investment in a special purpose cask for
target rods maynot be warranted.
Shipmentof Tritium
DOE has two typesof containersfor shippingtritiumas describedbelow. Both
containersarecurrentlyconfiguredto shipthe tritiumas a gas. WhileDOE is
consideringchangingtheformof tritiumshipmentfromgas to a metalhydridethe
capacityof the shippingpackagesis expectedto remainessentiallythe same.
The UC-609shippingpackage(Figure4.4.7-I)provides'containmentand offers
impactand thermalresistancefor shipmentscontainingtritium,in any of its
forms,duringtransportunderbothnormaland accidentconditions.The tritium
to be shippedisplacedwithinanappropriatestoragevessel.The storagevessel
isplacedwithina stainlesssteelcontainmentvesselandthecontainmentvessel
placedwithinan insulatedsteeldrum.
The internalcavityavailableforinstallationof a tritiumstoragevesselis 10
inchesin diameterby 31 incheslong. The UC-60gshippingpackagecan contain
30 gm-molesgas,not morethan25 gm-molesof whichmay be tritium(150gramsor
1.5millioncuries).The containmentvesselissurroundedby aminimumof 2-3/4
inchesof Celotexinsulation.A steeldrum surroundsthe insulationand is the
externalsurfaceof the package. The packageis 25 inchesin diameterand 55
incheshighandweightsa maximumof 500pounds.
4.4.1-6
The containment vessel is considered to be the primary containmentboundaryand
wtll contain the tritium if the packageis exposedto the normal or hypothetical
accident conditions specified tn 10 CFR71. The storage vessel receives no
credit for trttium containment but is to be designed, certified and tested to
provide the maximumassuranceof containment under all shipping conditions.
Tritium is shipped in the LP-50 Packageat low pressure (23.2 psia at 25 C) in
a 50 liter 304L stainless steel container surroundedby an aluminumvessel and
Celotex insulation at least 4 inches thick in a ]6-gauge steel drumwhich is 23.5
inches ODby 40 inches. This packagehas a maximumcapacity of ]93,500 curies
of tritium but current shipments are being limited to about ]25,000 curies oftritium.
The LP-50is expectedto be removedfromservicein about3-4yearsandwill be
replacedby a containercurrentlyunderdevelopment.Thisreplacementshipping
containeris expectedto utilizeuranium-hydrideand have a tritiumcapacity
comparableto the LP-50.
Basedon the size,weightandcapacityof theUC-609shippingcontainer(orits
replacement)it ispossibleto fitthecontractquantityof tritiumintoa single
shipment. However,prudentpracticewouldargueagainstcommittingthe entire
annualoutputto a singleshipment.
Shipmentof SpentTarqetRodsAfterExtractionof the Tritium
Spenttargetrodsareplacedin a canisterwiththe sameexternalconfiguration
as a BWR fuelassemblyandwouldbe compatiblewithbothexistinglicensedspent
fuelcasksandthe transportationsystemplannedfordisposalof commercialLWR
spentfuel. As a result,the transportationlogisticsw_ll be determinedby
whetherthe spenttargetrods are disposedof as low levelwasteat a DOE or
commercialdisposalsiteor as corecomponentsinthe federalrepository.Under
currentregulationseitheroptionis possible.
4.4.1-7
Plywood disk_j_
Heat shield - Plasticplug
Cerablanket insulation
(_ !' - _
Aluminum honeycomb, .... 5.0 in. on each end
3.7 in. on sides
" " Carbon steel drum
\ \ 24 in. insidediameter52½ in. insideheight
_ , / / Rubber I_ads
Vesselcarrier
\\ X
\\
54.5 in. Storage vesseltotal height
\\
I
Containment vesselType-316 stainlesssteel
\ \ _ 18 in. outside diameter1/8-in.-thick wall
Cavity 10 in. diam x 31 in. long
-,j Aluminum tube
\\
Celotex insulation/ / 2.9 in. on sides
4.0 in. on each.snd
l Cavity 18in. diam x 44 in. longI= 25 in. max diam
Figure 4.4.1.1
Model UC-609 Shipping Package
4.4.1-8
If the spent target rods are disposed of as low level waste the transfer cask
must be compatiblewith the packagingand unloadingrequirementsat the disposal
site. Further,unlessthe extractionsite is also permittedfor low level waste
disposal,the cask would have to be licensed for use over public roads. Whi_.q
it may be possible to interfacea commercialspent fuel cask with the disposal
site packagingand unloadingrequirementsa specialpurposecask is likelyto be
a more practicaland cost effectivesolutionin this case.
If the spent target rods are disposed of a core components in the federal
repositorythe followingconsiderationswould apply. A morecompletedescription
of this transportationsystemwas provided in Section4.4.4.
I0 CFR 961 providescriteriafor "standardnuclearfuel" which can be shippedto
the federal repository. This regulationalso allows for shipment of non-fuel
componentssuch as controlspiders,burnablepoisonrod assemblies,controlrod
elements, thimble plugs, fission chambers, and primary and secondary neutron
sources,that are an integralpart of the fuel assemblyand which do not require
special handling. While not directly applicable to spent target rods it is
likelythat they could be classifiedas "core components"if DOE chose to do so.
A legal-weighttruck cask is available for shipment of spent nuclear fuel
assemblies. This cask can hold g canistersof target rods which would require
eight annualshipmentsfromthe extractionfacilityto the MonitoredRetrievable
Storage (MRS).
Two Multi-PurposeCanistershave been conceptuallydesigned under the Civilian
RadioactiveWaste Management System (CRWMS) for rail transportationof spent
nuclearfuel from civilian reactorsites to the MRS or the FederalRepository.
The smaller MPC will hold 24 BWR assembliesand therefor would require three
annual shipmentsof consolidatedtarget rods. The largerMPC would accommodate
40 canisters of consolidatedtarget rods. Two annual rail shipmentswould be
requiredwith this larger MPC for the referencetritiumproductioncycle.
4.4.1-9
Other Considerations
Although the reference design for tritium production involves discharging all the
target rods after one cycle the potential exists to increase target rod exposure
to three cycles. This would reduce shipptng requirements throughout the tritium
cycle. However, since the number of shipments with one cycle exposure is
relatively small a change to three cycle exposure would not substantially change
the transportationlogisticsfor tritiumproduction.
Co-location of various tritium production operations can also reduce the
transportationrequired. While there are other good reasons to consider co-
location,transportationis not likely to be a controllingfactorbecauseof the
relativelyfew shipmentsinvolvedeven with a totallydispersedconfiguration.
From a transportationviewpoint,the differencein location of the extraction
facility (SRSor the reactorsite) affectsthe numberof shipmentsof irradiated
target rods and how the tritiumproduct is transportedto its end use location.
Locating the extraction facility at SRS avoids shipping the tritium in an
attractiveform (e.g purifiedproductgas or metal hydride). However, it could
requirean additionallongdistanceshipmentof the irradiatedtargetrods unless
they were disposedof as Low Level Waste at the SRS site. Again, becauseof the
existinginfrastructureand the few shipmentsneeded,transportationis unlikely
to be a controllingconsiderationin siting the extractionfacility.
The transportationrequiredto disposeof the spent target rods after extraction
is dependenton the waste disposalmethod selected (Low Level Waste or Federal
Repository). Disposalas a low level waste would probablybe less expensiveand
does not competewith spent fuel for priority in the transportationsystem or
space in the federal repository.
Under existing regulationspertaining to the Federal Repository the highest
transportationand disposalpriorityis assignedto the oldestspent fuel. Since
the MRS/Repositorymay just have begun operationat the time spent target rods
would need to be shipped, expanded storage would probably be needed at the
extractionfacility if disposal in the FederalRepositorywere selected.
4.4.]-I0
4.4.1.4 References
DOE Order 5630.13A,Master Safeguardsand SecurityAgreements
DOE Order 5630.14A,Safeguardsand SecurityProgram Planning
DOE Order 5632.2A, Physical Protectionof Special Nuclear Material and Vital
Equipment
DOE Order 5633.2A, Control and Accountability of Nuclear Materials:
Responsibilitiesand Authorities
DOE Order 5633.3A,Control and Accountabilityof NuclearMaterials
DOE Order 5633.4,NuclearMaterialsTransactions:Documentationand Reporting
DOE Order 5633.5, NuclearMaterialsReportingand Data SubmissionProcedures
Certificateof ComplianceUSA/6678/B()(DOE)Rev. O, for the LP-50 ShippingCask
Certificateof ComplianceUSA/9932/B(U)(DOE) Rev. 4, for the UC-609Shipping
Package
DPSPU-74-124-5,SafetyAnalysisReport:PackageLP-50TritiumPackage,May 1975:
Rev 2 issuedApril, 1988
Officeof CivilianRadioactiveWasteManagement(OCRWM)Bulletin,MPC Conceptual
Design Report Submittedto OCRWM, November,1993.
UCRL-52424,SafetyAnalysisReporton Model UC-609ShippingPackage,August 1977
UCRL-ID-111494,SafetyAnalysisReporton Model UC-609B(U) DOE ShippingPackage
4.4.1-II
4.4.2 TRANSPORTATION OF PLUTONIUM MATERIALS TO MOX
FABRICATION FACILITY
A. Off-Site Shipments
One of the feed materials for the mixed oxide fabrication plant is plutonium oxide. The source of
the plutonium oxide is the plutonium metal pits from disassembled nuclear weapons. Present
reconfiguration plans for the DOE weapons complex envision the first module of a plutonium
storage facility to be built by 2001 and a complete plutonium storage and processing facility to be
in operation by 2010. The nature of the stored plutonium (whether it will be metal or oxide) has
not been determined. There are severe problems in storing plutonium oxide for long periods of
time including americium buildup, gas evolution, and the hygroscopic nature of the powder.
Recommendations have been made that Pu02 should not be stored for more than three years.
Whatever storage for is selected, it appears that the source of material for the plutonium
disposition program will be from these new storage facilities.
While the conversion of the plutonium metal to oxide is not specifically addressed in the DOE
Phase 1C Statement of work, it is addressed here to provide a more complete description of
program needs. This conversion could be performed by DOE contractors at the fuel fabrication
site or a separate site. A discussion of the transport of plutonium metal forms follows, as well as
a description of the transport of plutonium oxide to the fuel fabrication plant.
Plul_onium M¢l_tl Transport
The shipment of plutonium metal pits to various DOE facilities has been safely carried out for
years. Since these are U.S. Government shipments and consist of nuclear weapons components,
they are regulated under DOE Order 5610.1, "Packaging and Transporting of Nuclear Explosives,
Nuclear Components, and Special Assemblies." The order states that these materials must be
packaged and transported to provide a level of safety at least comparable to that provided by
packaging and shipment, in accordance with applicable regulations of other radioactive and
explosive material. The references cited are DOT 49 CFR 100 through 179, "Hazardous Materials
Regulations." (These in turn, refer to Title 10 CFR 71, "Packaging of Radioactive Material for
Transport" which promulgates Federal regulations for the packaging of radioactive material for
4.4.2-1
transport); DOT 49 CFR 390-397, "Federal Motor Carrier Safety Regulations;" DOE 10 CFR
871, "Air Transportation of Plutonium," NRC 10 CFR 20-21-73; Interim Management Directive
5001; "Safety Health and Environmental Protection," of 9-29-77 to DOE 5481.1 "Safety,
Analysis and Review System," of 3-20-78.
Some of the key points of the regulations are presented here.
The type of packaging required in the DOE order is governed by the amount and radioactivity of
the fissile materials being transported. Fissile materials are defined as uranium-233 and 235,
plutonium-238, 239 and 241, neptunium-237 and curium-244. (It should be noted that neptunium
and curium are not listed as fissile in 49 CFR 173.403 and 10 CFR Part 71.)
The least restrictive package, is a Type A package. Type A packages cannot exceed a quantity of
aggregate radioactivity determined as A1 for special form radioactivity and A2 for normal form
radioactivity. Values for A1 or A2 quantities are shown in Appendix A, Table A-I, 10 CFR 71.
Special form radioactivity is radioactive material which is either a single solid piece or is contained
in a sealed capsule that can only be opened by destroying the capsule. The piece or capsule has at
least one dimension not less than 5 mm (0.197 inch) and it satisfies test requirements outlined in
10 CFR 71.75. The amount of radioactivity for the plutonium pits is too high to be shipped in
Type A package. Therefore, Type B packages are required. (They include the package and its
radi oacti ve contents.)
Another factor used to regulate shipments is the fissile classification. The material to be
transported is placed into one of three categories based on the controls needed to provide nuclear
criticality safety during transport. The categories include fissile class I which permits transport of
unlimited numbers of packages in any arrangement and requires no criticality control, and fissile
class II which permits the package to be transported with other packages in any arrangement
provided that they do not exceed an aggregated transport index of 50, with individual packages
having a transport index of not less than 0.1 and not more than 10. (The Index is a dimensionless
number rounded up to the first decimal place.) The transport index is further defined as a number
expressing the maximum radiation level in millirems/hour at 1 m from the external surface of the
package, or for fissile class II packages, the same definition for the number is obtained by dividing
4.4.2-2
the allowable number of packages which may be transported together, as determined in 10 CFR
71.59, by 50, whichever number is larger.
The final category is fissile class III which is a shipment of packages which is controlled in
transport by specific arrangements between the shipper and the carrier to provide nuclear
criticality safety. There are special requirements for plutonium shipments. The only requirement
applicable to plutonium metal is that plutonium in excess of 20 curies/package must be shipped
as a solid (which plutonium metal meets.)
The fissile class III shipments of plutonium would require a Type B package. These packages are
specified in 49 CFR 173.416. The package used is per DOT specification 6M which can be used
for only solid or gaseous radioactive materials that will not undergo pressure generating
decomposition at temperatures up to 121° C (2500F) and that do not generate more than 10 watts
of radioactive decay heat. A 6M package is a metal container that can contain no more than 4.5
kilograms of plutonium. This is due to the 10 watt decay heat limitation. It also provides a
double containment.
The Savannah River Site (SRS), for example, has four certified packages that are currently in use.
These are NRC approved packages (certification numbers 9965 through 9968.) These packages
are used for shipping solid metal or alloys of fissile or other radioactive material, as well as oxides,
scrap, or powders. The maximum payload of each of the four packages is 53 pounds. The gross
weight of the packages varies from 193 to 627 pounds. The difference in weight is due to
shielding requirements.
These SRS packages are placed in a rack, called a "bird cage," to maintain distance between
packages and placed on a special truck called an SST (Safe Secure Transporter.) A maximum of
125 of these packages can be shipped on the same SST.
Plutonium Oxide Transport
Plutonium oxide is not a nuclear weapon component and its transport is governed by DOE Order
5480.3, "Safety Requirements for the Packaging and transportation of Hazardous Materials,
4.4.2-3
Hazardous Substances and Hazardous Waste." The DOE Orderincorporates significant parts of
Title 10 CFR 71, "Packaging of Radioactive Material for Transport," which promulgates Federal
regulations for the packaging of radioactive materialfor transport and is the governing regulation
for civilian shipments of radioactivematerial.
Plutonium oxide must also be shipped in a Type B package. The same certifiedcontainers listed
above for plutonium metal shipments are also used to transport plutonium oxide. In addition,
there is another package, Certification #5320, that can be used for shipping plutonium and
americium oxides. This package uses a 10"by 12" diameteraluminum pipe as an inner container,
with a cylindrical stainless steel pressurevessel as the secondarycontainment. Thiscan is limited
to 357 grams of plutonium of any isotopic composition (403 grams of Pu0 powder) or 176 grams
of americium (200 grams Am0 powder.)
B. On-Site Shinments
A draft DOE Order 5480.x was issued in July 1991 to cover on-site "Packaging and
Transportation of Hazardous Materials Substances and Wastes." While not official, the Oak
Ridge Reservation (ORR) is following most aspects of the order. SRS has also adapted some of
the provisions. The order calls for site specific transportation safety manual to be prepared and
maintained for each site. The manual documents safety, health, and environmental protection
measures being taken for all on-site transfers of hazardous materials. While packaging
performance is the preferred way to ensure overall safety, DOE desires an integrated approach
which considers the packaging in combination with specified communication and controlmeasures.
The on-site packaging and transportation programs include the following elements: Identification
of responsibilities, lines of authority and program approval procedures, definition of safe
packaging requirements, descriptions of transportation systems, operational controls, safety
methodology, site descriptions, emergency response and process for non-routine packaging and
transportation activities. ORR has a Transportation Safety Manual outlining these activities and
SRS is studying issuing such a manual. At present SRS has a collection of these procedures. The
4.4.2-4
draft Order states that "Adherence to Federal standards, normally applicable to off site
transportation, is an acceptable approach to meeting on site standards."
Other factors also regulate on-site movement. The least restrictive is movement within a Material
Control Area (MCA). In this case there is less documentation and security measures required.
However, the same stringent safety and criticality controls are employed. In case of movements
between MCA's security surveillance is greater with constant surveillance of the transport'slocation and condition. If the routes selected for the movement can be cleared of other traffic, the
use of dedicated security and emergency response readiness can be reduced. If the transport is
across a public highway on the reservation, which is possible at many DOE sites, the shipment
would probably be treated as an off-site shipment and all the measures required for such
shipments must be followed.
4.4.2-5
4.4.3 TRANSPORTATION OF MOX FABRICATION FACILITY ANDREACTOR PLANT NUCLEAR WASTE
This section presents a study of the transportation of the waste forms from the ABWR and the
reactor complex fuel fabrication plant to on-site storage as well as transport to off-site DOE
disposal facilities. The study included the type of containers and shipping carriers required to
meet the applicable Federal, State and local regulations.
TRANSPORT OF FUEL FABRICATION PLANT NUCLEAR WASTES
Although a new fuel fabrication plant of modern design will be required, it is assumed that some
minimum amount of wastes would still be generated that cannot be recycled within the plant.
During the fabrication process, some of the material must be recycled because the do not meet the
final pellet specifications. This would include materials required for setting up machine
parameters, material for test specimens, material ground off during machining the pellets and
pellets with surfaces defects such as cracks. Either a dry recycle or wet recycle is used depending
on the purity of the material. If the material is out of specification because of damage, such as
chips or cracks, it is ball milled and the powder returned to the process head end. This is a dry
method. However, if it is powder from cleanup of the glove boxes which could contain
impurities, a wet method is used. This wet method consists of dissolving the material and
processing the solution by solvent extraction or ion exchange.
There are three categories of nuclear waste that can arise from fuel fabrication. These are solid
glove box waste, solid waste from retired equipment, and liquid waste from scrap recovery and
from work in the plant analytical laboratory. This waste can be classified as low level wastes,
mixed wastes (radioactive and hazardous waste), and transuranic (TRU) waste. Before the waste
can be disposed, it may be treated for: volume reduction of the waste; reduction of the plutonium
contained in the waste; and solidification of the liquid waste. A typical plant will usually use
compression or incineration to reduce the volume. Plutonium recovery is accomplished by ash
leaching, acid digestion and a washing process. Solidification of the caste can be by cementation or
bitmunization, such as is practiced in French and German plants. Some references state that less
than 0.5% of the throughput of the plutonium will be present in the waste and that more than
4,4,3-1
50% of this plutonium can be recovered. As an example, the new MELOX plant in France is
designed to recycle as much plutonium waste as possible. Low activity liquid wastes are treatedin the Marcoule treatment unit. At MELOX waste which can be burned will be incinerated with
the plutonium being recovered from the ash. The plant does not expect to produce more than 100
m3 of storable waste per year, of which only 4 m3 will require underground storage. MELOX is
a 120 MT/year plant.
Some DOE sites' have facilities to dissolve and recover the plutonium at the present time and
these could easily handle the additional feed stream from an on-site MOX plant. Where future
operations of these facilities at DOE sites are uncertain ,if these programs are not continued, a
plutonium disposition program might have to recreate what are already existing capabilities.
The majority of a fabrication plant waste can be expected to be the equivalent of Class A low level
waste with the small amount being equivalent to Class C low level waste requiting disposal in the
equivalent of Class B/C storage vaults. Any greater than Class C waste and TRU waste would
and could be disposed of at the WIPP or Yucca Mountain repositories. The disposition of mixed
wash from DOE facilities is not determined at this time.
A. On-Site Shipments:
The DOE Orders 5480.x apply to on-site waste shipments as well as to the on-site oxide or metal
shipments. At ORR for example, low level waste that can be compacted is placed in dumpster
type vehicles and driven to the waste burial ground where is compacted and placed into lined
trenches provided with a sealable cap. Non-compatible low-level waste materials are placed in 4'x
4'x 6' steel-boxes and transported to the burial ground. Material such shoe co,,ers, gloves, papers
are placed in stainless steel 55-gallon drums for storage. At SRS a Central Incineration Facility to
handle low-level combustible waste is planned but the project is currently on "Hold". At ORR,
an existing TSCA incinerator is currently accepting only low-level liquid waste, primarily of
contaminated oil. The oil is shipped in 55-gallon drums or in tanker trucks. While the facility
was designed to burn solids as well as liquids because of the large backlog of liquids there are no
existing plants to burn solids.
4.4.3-2
B. Off-Site Shipment:
DOE order 5480.3 covers off-site shipments of radioactive and hazardous waste. It is believed
that the vast majority of the waste generated by the fabrication plant could be retained on-site so
there would be no need to sl_ip off-slte. Where off-site shipment is desired however, a system
similar to that in use at ORR could be used. ORR has a contract with a private company which
can sort out clean material, and either incinerate or melt the remaining low level material. A
combination of drums and boxes, similar to what a private utility uses to handle shipments of low
level nuclear waste from reactors to storage sites, can also be used.
_TRANSPORT OF REACTOR COMPLEX NUCLEAR WASTES
Due to the dedicated waste volume reduction and plant worker radiation exposure reduction
features designed into the evolutionary ABWR, the type of waste from a mixed oxide fueled
ABWR will be similar in type but less in quantity than that generated in current BWRs. As a
benchmark for comparison of waste types, The River Bend unit of Gulf States' utilities is a 936
MWe BWR. This reactor is expected to generate 155 55-gallon drums (1,163 ft3)per year of dry
activated waste (DAW), Class A low-level waste, plus another 12 95 ft3 -boxes (1,140 ft3) of
the same waste. The dose rates on these waste range from 10 mRem/hour - 2 reins/hour. In
addition, about one 120 ft3 High Integrity Container (HIC) shipment/year are usually required for
resins which are a class B low-level waste with dose rates ranging from 8 Reins/hour - 30
Reins/hour. Riverbend also averages 45,207-ft 3 liners/year (9,315 ft 3) of Class A low-level
radwaste sludges. These sludges have dose rates from 0.5 Reins/hours- 10 Reins/hour. The
_proposed 1300+ MWe plutonium disposition ABWR has been designed to emphasize waste
management and would produce less waste than a conventional BWR. Reactors also generate a
limited amount of greater than Class C low level waste (GTCC) in the form of activated metals.
Radiation levels of this material can exceed 30,000 Rems/hour but only average one shipment/year.
This Type B material will also be generated when the reactor is decommissioned.
4,4.3-3
A. On-Site Shipments:
Again, the same DOE Orders regulate on-site reactor waste shipments as described above. On-
Site shipments are made using the 55-gallon drums and metal boxes for the DAW, Class A, low-
level waste as shipping containers. Shipments to storage areas is made by fiat bed trucks or
covered trucks. At ORR, for example, resins from the research reactors on-site are dried,
packaged in metal canisters, and buried. The Class A radwaste-sludge could be shipped in
disposal liners, as civilian reactors do; to ship these materials from the reactor to low level burial
grounds. The GTCC waste is normally shipped in spent fuel shipping casks which are heavily
shielded.
B. Off-Site Shipments:
Off-site reactor waste shipments will be regulated by DOE Order 5480.3 and as noted previously
use the same type of shipping containers and casks as civilian reactors.
4.4.3-4
4.4.4 Spent Fuel Transportation & Logistics
4.4.4.1 Summary
A review has been conducted to determine the current status of the Civilian
Radioactive Waste Management System (CRWMS). Using that information, the
potential impacts on spent nuclear fuel (SNF) discharged from the Advanced
Boiling Water Reactor (ABWR)in a weapons gradeplutonium destruction modeweredetermined.
Although a significant numberof studies andevaluations have been conducted over
the years related to permanent disposal of SNF and high-level waste (HLW), anumberof issues remain to be finalized. Most of the evaluations relatedtothis
program are in the conceptual design stage and are expected to be firmed-up over
the next several years. This includes, but is not limited to: site
characterization evaluations leading to final site selection for the Federal
Repository; repository spent nuclear fuel (SNF) storage capacity, repository
waste acceptance criteria; site selection and design of a Monitored Retrievable
Storage (MRS) facility; design of the Multi-Purpose Canisters (MPC); and, design
of transportation casks for SNFacceptance, handling, transportation, storage and
disposal.
As a result of this review,no significantconcernswere identifiedrelated to
acceptance,handling,transportation,and disposal of spent mixed-oxide (MOX)
fuel from an ABWR in a plutoniumdispositionmode. Likewise,few firm design
requirementswere found which could be used at this time to firm up the ABWR MOX
design and other ABWR designor processissuesto ensure compatibilitywith plans
for the CRWMS.
Compatibilitywith the CRWMS is highlydesirablefor economicand other reasons.
To achieve this, it is recommendedthat an on-going awarenessof the status of
the CRWMS be maintainedfor potentialimpact. The Departmentof Energy (DOE) is
strongly encouraging continuing involvement and input from stakeholdersto
influencedesign and implementationof the CRWMS.
4.4.4-1
I!
4.4.4.2 Discussion
The purpose of this task is to: 1) present the results of a review of the current
plans to ship SNFfrom civilian nuclear power plants to the Federal Repository;
and, 2)assess the range'of options available for spent nuclear fuelfroman ABWR
tn a plutonium disposition modeto meetthe projectneeds. A,significant amount
of the information presented below is still conceptual in nature. It is expected
that studies by DOE and support_contractors over the next several years will
• result in firming up information of relevance to destruction of weapons grade
plutonium in the ABWR.
A comprehensivereview of the entire CRWMS is currentlybeing conductedby the
Secretaryof Energy. Resultsof this review is scheduledto be completedin 30-
60 days and may presentadditionalpolicydirectionof relevanceto this program.
DOE StandardContract
A standardcontract for transportof SNF and high level radioactive(HLW)waste
is required by Section 302 of the Nuclear Waste Policy Act of 1982. This
contract iscontainedin 10CFR g6 which includescontractualprovisionsrelating
to disposalof SNF and HLW from civiliannuclearpower reactors requiredto be
licensed under Sections 103 or I04(b)of the 1954 Atomic Energy Act as amended
(42 U.S.C. 2133, 2134(b)).
Under provisions of this contract, the DOE will make available nuclear waste
disposal servicesto the owners and generatorsof SNF and HLW. In exchangefor
these services,the ownersor generatorsof such fuel or waste pay fees on a full
cost recoverybasis into a NuclearWaste Fund. Thesecontractshad to be entered
into by June 30, 1983 or by the time generationof spent fuel or high level waste
generation is initiated,whichever is later. These contracts provide for: I)
title transfer of SNF and HLW to the DOE at the utilitiessite; 2) shipmentof
the SNF and HLW to the FederalRepository;and 3) disposalof the SNF and HLW by
the DOE followingcommencementof the operationof the repository.
4.4.4-2
It is assumedthat the contractual provisions of 10 CFR961 would be waived for
the spent ABWRHOX fuel since it is owned by DOE. However, the technical,
financial and programmatic aspects of the Standard Contract are likely to apply.
DOEwtll provide a cask for shipment of the SNF and/or HLWfrom the utilities
nuclear power reactor, or such other locat|on designated by the user, to the DOE
st(}rage facility(s). The cask will be delivered to the user sufficiently in
advance of the shipment, suitable for use at the site, meet applicable regulatory
requirements and be accompanied by the following information: 1) written
procedures for cask handling and loading; 2) specifications for user furnished
canisters for containment of failed fuel; 3) training for user personnel in cask
handling and loading; 4) technical information, special tools, equipment, lifting
trunnions, spare parts and consumables needed to perform incidental maintenance
on the cask; and 5) documentation on the equipment supplied by DOE.
The user will be responsible for incidental maintenance, protection and
preservation of the casks provided to the user. The user will also be
responsible for providing all preparation, packaging, required inspections and
loading activities in preparation for transportation of the SNF and HLWto the
DOEstorage fac i 1i ty (s).
A prioritywill apply for shipmentof SNF from generators. This priorityranking
is issuedby DOE in aF annualAcceptancePriorityRanking (APR) reportwhich is
based on the date the SNF was dischargedfrom the reactor. The oldest fuel or
waste on an industry-widebasis will have highest priorityfor shipmentto the
repository. Reinserted SNF will be removed from the APR and rescheduledfor
shipmentin accordancewith a new APR based upon it's permanentdischargedate.
DOEhas indicatedthat it will accept "standardnuclearfuel" asdefined by fuel
specificationsin the contract. These specifications,inessence,establishthe
waste acceptance criteria of spent nuclear fuel for disposal in the Federal
Repository. DOE will evaluate the feasibility of disposing of non-standard
nuclear fuel and inform the utility of any adjustmentsthat may be required.
Standard nuclear fuel specificationsinclude:
4.4.4-3
• Maximumphysical dimensions)
Maximumnominal physical dimensions are as follows:
Overall length: BWR--14 feet, ]1 inches; PWR--14 feet, ]0 inches
" Active"fuel lenqth: BWR--12 feet, 6 inches; PWR--12.feet, O.inches ....,_
_ross sectlon: BWR--6 tnches x 6 inches (not Including -the channel);.,,PWR--9inches x 9 inches
• Non-fuelcomponents
Non-fuel components including, but not limited to, control spiders, burnable
poison rod assemblies, control rod elements, thimble plugs, fission chambers, and
primary and secondary neutron sources, that are contained within the fuel
assembly, or BWRchannels that are an integral part of the fuel assembly, which
" do not require special handling, may be included as part of the spent nuclear
fuel delivery for disposal.
• Minimum cooling time
The minimum cooling time for fuel is five years.
• Non-LWR fuel
Fuel from other than LWR power facilitiesshall be classifiedas non-standard
fuel. Such fuelmay be unique and requirespecialhandling,storageand disposal
facilities.
• Consolidatedfuel rods
Fuel which has beendisassembled and storedwiththe fuelrods in a consolidated
manner shall be classifiedas non-standardfuel.
• Failed fuel
Assembliesshall be visuallyinspectedand those which are structurallydeformed
or have damaged claddingto the extent that specialhandingmay be required,or
4.4.4-4
for any reason cannot be handled with normal fuel handling equipment, shall beclassified as failed fuel.
A revision to the current Standard Contract is expected to be developed and
• published inthe'Federal Register by the DOEfor commentduring CY-Ig94. This
revision is expected to clartfy, modify and add_several_newprovisions to the
contract, including changesto someof the standard nuclear fuel specifications.
This is expected to include a possible modification of the spent nuclear fuel
cooling time, establish criteria for and disposition of failed fuel, address
priority for disposing of SNF from permanently shutdown reactors, address
disposition of consolidated fuel rods, and other potential changes.
SNF Canisters
Studieswere initiatedin 1992 to determinethe feasibilityof using sealed
canisterstoaccommodateSNFduringwasteacceptance,transportation,storageand
disposaloperationsthroughoutthe CRWMS. Initiationof thesestudiesresulted
primarilyfrom concernsof repeatedSNF assemblyhandling and packaging
operationswhichwouldbe requiredwith otheroptions.
DOE iscurrentlyevaluatinga September30, 1993draftof the MPC Implementation
ProgramConceptualDesign Phase Report. Productsfrom this study include
conceptualdesigns for the MPC, the transportationcask, the monitored
retrievablestorage(MRS)facilityand the utilitytransfersystem. The MPC
conceptualdesign includes canister configurationsfor containingboth
pressurizedwaterreactor(PWR)and boilingwaterreactor(BWR)SNF assemblies
in the same MPC. The transportationcask conceptualdesignsprovidecasks
necessaryfor transportingMPCs fromwastegeneratingand storagesitesto the
repository. TheMRS facilityconceptualdesignprovidesfacilitiesfor loading
MPCs and storingthem untilthe permanentrepositorybecomesoperational.The
utilitytransfersystemconceptualdesignprovidesan on-sitetransfersystemat
commercialreactorfacilitiesto load,handleand storeMPCs.
4.4.4-5
The MPCpreliminary design concepts are based on several assumptions:
• A legal weight truck caskwith a capacity of 4 PWRor g BWRassemblies, nominal
2S-ton loaded weight, for facilities with limited cask handling capabilities.
• A mediumsize MPCfor rail transportation with a capacity of 12_PWRor_24 BWR
assemblies (without reliance on burnupcredit for either type of fuel). This
will involve a nominal 75 ton loaded weight MPCin a transportation cask.
• A largesize MPC for rail transportationwith a capacityof 21 PWR (with
relianceon burnupcredit)or 40 BWR assemblies(withoutrelianceon burnup
credit). This will involvea nominal 125 ton loaded weight HPC in a
transportationcask.
• SNFwould be initiallyacceptedat theMRS in theyear2000. Betweentheyear
2000and 2010all SNFwouldbe transportedto the MRS for temporarystorage.
• Prior to 2010, all MPCs receivedat the MRS would be removedfrom the
transportationcask and placedin a steelor concretestoragecask for storage
at the MRS. Any SNF assembliesreceivedin legalweighttruckcaskswouldbe
unloadedat the MRS and placedin eitherof the two MPCs and then placedinto
storagecaskspendingsubsequentshipmentto the repository.
• Beginningin2010,theMRSwouldserveas a stagingarea. MPCsarrivingat the
MRS wouldbe placedon dedicatedtrainsfor shipmentto the federalrepository.
The MPCs consistof a cylindricalshellwithtwo lids,a spentfuelbasketand
a shieldplug. ThespentfuelbasketprovidesstructuralsupportfortheSNFand
a mechanismforthetransferof theheatgeneratedby theSNFintotheMPCshell.
Thespentfuelbasketalsoprovidescriticalitycontrolto ensuretheSNFremains
subcriticalunderall definedcircumstances.
The largeMPCdesignrequiresburnupcreditfor shipmentof PWRSNF. Themedium
12 PWR basketdesignemploysa water gap flux trap arrangementto improve
effectivenessof the boratedaluminumneutronabsorberpanels. Similarlythe
4.4.4-6
smaller size of a BWRfuel assembly allows the borated aluminum panels to come
closer to the center of each fuel assembly which results in the neutron absorber
panels being more efficient. Using this design approach, the BWRbasket
configurations do not require burnup credit for storage or transportation
licensing with either of the two MPCs. Criticality requirements wtll need to be
calculated for spent ABWRfuel to ensure ,compliance,with the preliminary MPC
designs.
A SNF cladding temperature limit of 340 C was adopted as the design limit for
preliminary MPCthermal analyses to ensure that storage limits could be met in
the transportation cask design. This is the 10 CFR72 storage temperature limit
for ten-year cooled fuel. ]0 CFR 73.60 and 67 state that the radiation dose
after a two year cool down period must be at least 100 rem/hr at a distance of
three feet from all surfaces of the fuel bundle. This is to provide protection
_against theft or diversion. _ Structural design of the MPCwas based on a 60 g
acceleration limit for the hypothetical 9-meter drop accident scenario. This
represents the maximumacceleration that light water reactor fuel assemblies must
structurally withstand in a side drop without failing. Calculations of the
cladding temperature and radiation dose for spent ABWRfuel under the above
conditions and structural integrity for the various accidents involving handling
and shipping of ABWRspent nuclear fuel need to be determined to ensure
compliance with these preliminary requirements.
Someconsideration is being given by DOEto design a smaller MPCfor legal weight
truck use. Currently, there are about nineteen reactor sites that are limited
to legal weight truck cask shipments. Most reactor sites will be able to use the
125 ton MPC.
SNF T.ransportationCasks
The Office of CivilianRadioactiveWaste Management (OCRWM)has an objectiveof
developing and placing into operationa system capable of transportingspent
nuclear fuel and high-level waste from the various waste sources to waste
receivingfacilitiesbeginningin January,2000 and subsequentlyto the Federal
Repository.
4.4.4-7
The Department of Energy has sponsored a number of transportation cask design
efforts to accommodatethe various assemblies expected to be accepted for
disposal. The MPC is sealed and placed inside the transportation cask for
shipment by rail to the MRSor Federal Repository. At this time, there are
conceptual designs for two transportation casks. One is to transport the large
125 ton MPC and the second is to transports,the 75 ton MPC.-These_wetghts
represent the under the hook weight at the reactor spent fuel storage pool and
includes the MPC (with outer lid removed), SNF, water, transportation cask body
and lifting yoke.
Design temperature limits and acceleration values for preliminary MPC
transportation cask analysis and design are the sameas used for MPCanalysis and
design (i.e. 340 C and 60 g acceleration). The cask systems must be certified
by the Nuclear Regulatory Commission (NRC) and must comply with limits imposed
by various'agencies including the DOE, NRCand the Department of Transportation
(DOT).
MRS Storageof SNF
Without the availabilityof a Federal Repository, inventories of SNF have
continuedto build in the storagepools at reactorsites,forcingmany utilities i
to take actionsto increasethe SNF storagecapacitiesat their sites. The most
prevalent action taken by the utilitieshas been to re-rack their spent fuel
storagepools with highercapacitystorageracks. A few utilitieshave used fuel
assemblyconsolidationas a means of increasingthe capacityof their pools,and
a few more have installeddry storagefacilitiesof varioustypes on their sites
to permit storageoutsideof their reactorstoragepools. Currently,fiveonsite
spent-fueldry cask storagefacilitieshave site-specificlicensesfrom the NRC
and are in operation.
The DOE was authorizedby Congressin the NuclearWaste PolicyAmendmentsAct of
1987 to develop a Monitored Retrievable Storage (MRS) facility to provide
temporary above-ground storage for a limited amount of SNF from commercial
reactors. For planningpurposes,the MRS iscurrentlyscheduledto receivespent
nuclearfuel from utilitiesin January, 2000. A recommendationfor locationof
4.4.4-8
the site of the MRSis currently scheduled for September1994 with start of
construction in September1998.
The MRSfactllty conceptual design provides facilities for temporarily storing
MPCscontaining SNFunttl the permanentFederal Repository becomesoperational.
MPCsarriving at the HRSare removedfrom the transportation caskrplaced wtthtn
a disposal container designed for permanentdtsposa] and then placed within a
vertical steel or concrete structure for temporary storage at the MRS. The MRS
conceptual design also allows handling of bare SNFassemblies.
A dedicated factltty for inspecting, testing, andmaintaining the cask systems
was recommendedby the General AccountingOffice as the best meansof assuring
their operational effectiveness, safety, and regulatory compliance. In 1987,
OCRWMrequested a feasibility study be madeof a CaskMaintenanceFacility (CMF)
that would perform the required functions. It is currently envisioned that the
CMF will be integratedwiththe MRS facility.
With regardto the ABWR,itwouldappearthatmaximizingthe spentfuelstorage
capacityin the reactorbuildingwouldbe prudentbecauseit wouldminimizethe
handlingnecessaryto placespentfuel back intothe reactor,if desired,for
additionalpowerproduction.This wouldalso appearto be cost effectiveas
comparedto constructionof a centralizedstoragefacility. In any event,a
minimumtenyear storagecapacityat the ABWRsitewouldappearto be necessary
beforeshipmentto theMRSor repositorywouldbe allowedbecauseof a numberof
factors,including:I) existenceof the currentlargebacklogof spentnuclear
fuel;2) DOE'srequirementto givepriorityforshipmentof theoldestspentfuel
to the repository;3) possiblehigh priorityto be given to utilitiesfor
disposingof SNFfrompermanentlyshutdownreactors;and,4) sufficientcooldown
to meet the maximumcladdingtemperatureof 340 C.
Currently,defenseSNF is precludedfrom being storedin the MRS. It is
expected,however,thatthisprohibitioncouldbe eliminatedas discussedbelow.
4.4.4-9
Federal Repository Storaae
Site characterizationof YuccaMountain,Nevada is currentlyin progressto
determineit'ssuitabilityfordevelopmentasa FederalRepository.The Federal
Repositoryconceptualdesignwillaccommodate70,O00MTUof waste:63,020MTU of
SNF; 6,340MTU of DefenseHlgh-LevelWaste_(bHLW)and.640MTU of West-Valley
High-LevelWaste(WVHLW).Theearliestcurrentscheduleforcomenclngoperation
of the FederalRepositoryis 2010.i
The repositoryisbeingcreatedundertheNuclearWastePolicyAct (NWPA)of 1982
(U.S.C.Title42 I0101,as amended)andDOE was designatedas the ownerof the
facility. The law initiallyprecludedwastegeneratedas a resultof defense
activities.However,in 1985PresidentReagansignedan authorizationto allow
acceptanceof defenseHLW into the facility. Scopinghearingsare now in
progress to allow DOEto accept defense SNFassemblies into the repository.
DiscussionswithDOEstaffwereinitiatedto determineiftherepositoryisfully
committed.Thisinformationiscurrentlynotavailablebecausetheabovestorage
capacityis stillconceptual. By the year 2000, about40,000MTU of spent
nuclearfuel will existat nuclearreactorsites. By the time the last NRC
licensefor the currentgenerationof nuclearreactorsexpires,an estimated
totalof 87,000MTU willhavebeengenerated.Ifthe YuccaMountainrepository
is limitedto 63,020MTU of SNF,additionalpermanentdisposalcapacitywillbe
required. SNF for permanentdisposalfrom the ABWR may be placedin either
facility.
The waste package design criteria for the Federal Repositoryhave been
incorporatedinto the MPC conceptualdesign in order to meet repository
conceptualrequirements. A numberof importantdesign requirementsof the
repository,however,will not be finalizedfor severalyears. One of these
includesthe thermalloadingstrategyto supportthe licensingprocess. This
issueinvolvesthe thermalloadingof the repositoryin the near-fieldandfar-
fielddue to heat generatedby the emplacedwaste. Becauseof the schedule
associatedwith thermalloadingstudies,designof the MPC is expectedto be
finalizedbeforeit is knowniftheMPCdesignis compatiblewiththe repository
4.4.4-10
design. It may be necessary to repackage the SNFat the MRSor at the repository
in order to meet the thermal loadtng criteria. Alternatively, the MPCdesign
could be subsequently modified to be compatible with repository requirements.
MOXfuel wtll have a different heat load than U02 fuel. Decay heat calculations
as a function of time will need tobe performed for the-ABWRfuel as comparedto
BWRfuel assemblies from commercial nuclear plants.
A number of interfaces exist between the MPC and the repository facilities.
These include the possibility of a disposal overpack, waste handling building,
waste package transporter, and the subsurface emplacement operations. Each of
these interfaces were evaluated and considered in the MPCconceptual design.
In federal laws, regulations, and departmental directives, the US Congress, the
r Nuclear Regulatory Commission, and the US Department of Energy have developed
criteria for retrievability of waste emplaced in a geologic repository for high-
level radioactive waste. In response to these criteria, the Yucca Mountatn
Project is expected to have a capability to retrieve emplaced waste as a planned
contingency operation.
10 CFR 60 requires that "... a nuclear criticality accident is not possible
unless at least two unlikely, independent, and concurrent or sequential changes
have occurred in the conditions essential to nuclear criticality safety." One
of these changes must be the addition of amoderator- water. Concern is limited
to accident scenarios involving flooding of the repository.
During repository operation, bare fuel assemblies would be vulnerable to flooding
only in the hot cells of the waste handling building. Careful facility design
is expected to reduce the risk of a criticality accident during fuel handling
operations virtually to zero. In a11 other near-term repository environments,
the fuel is protected by watertight casks or disposal containers. Under post-
closure conditions involving intact spent nuclear fuel assemblies, it is unlikely
that criticality will be found to be a problem for any container configuration.
4.4.4-11
This wtll need to be vertfled for ABWRMOXfuel as compared to commercial BWR
fuel assembltes.
In the longer term (hundreds to thousands of years after repository closure),
there is a potential concern for container failure and flooding. In the very
long term, the 'containers and the_fuel..assembltes _themselves may_.have
disintegrated; tnthat case, the physical configuration of the fuel at the ttme
of emplacement is more or less academic. Current plans for analyzing post-
closure'criticality scenarios are based on the assumption that water Intrusion
has occurred. Criticality concerns with ABWRMOXfuel wt11 need to be determined
as compared to commercial BWRfuel assemblies for these conditions.
4.4.4.3 RecommendedFollow Up Actions
The most cost effective-approach for spent fuel managementwould be to ensure
compatibility of the spent ABWRMOXfuel assemblies with the CRWMS. To achieve
this goal the following actions are recommended:
• Maintain cognizance of the status of the Civilian Radioactive Waste Management
System (CRWMS). One of the first design finalization activities involves
issuance of an RFP for design of the MPCin the Spring of 1994 and award of the
contract in late 1994 or early 1995. Compatibility of ABWRMOXfuel with the MPC
design(s) would be critical.
• Defense SNF is currently not allowedwithin the MRS or Federal Repository.
Actions are being initiatedby DOE to allow defense SNF into both facilities.
The status of these initiativesneeds to be monitored and follow up actions
taken as necessary.
• Modificationsto the StandardContractare expectedto be publishedfor comment
during CY-1994. The modifiedcontractshould be reviewedand commentsprovided
to ensure that the StandardContractadequatelyenvelopesdesigncharacteristics
of spent ABWR MOX fuel.
4.4.4-12
• The MPCsystem (canister and transportation cask) must be licensed by the NRC
under a numberof regulations, including, but not ]imtted to: 10 CFR50, Domestic
Licensing of Production andUtilization Facilities; 10 CFR60, Disposal of High-
Level Radioactive Wastes tn Geologic Repositories; 10 CFR71, Packaging and
Transportation of Radioactive Material; and, IOCFR 72, Licensing Requirements
for the Independent Storage of Spent.Nuclear,.Fuel.-and_High-LevelRadioactiveWaste.
An effectiveinterface:should be established with the DOE Contractor(s)
developing the safety, environmental and other documentationfor this licensing
activity. Information related to the ABWRHOXfuel mayhave to be provided to
the Contractor(s), in addition to the specific issues identified below. This
will ensure that issues pertaining to spent ABWRMOX fuel are adequately
enveloped by the licensing documentation.
• Criticalitycalculationsneed to be performedfor spentABWR fuelto ensure
criticalityrequirementsaremet forthe MPCdesigns,transportationto theMRS
and repository,and storagewithinthe repository.
• Claddingtemperaturesof the ABWR MOX spent fuel assembliesneed to be
determinedto ensuremaximumtemperaturesdo notexceedthe MPC designlimitof
340 C followinga ten-yearcooldown.
• Structuraldesigncapabilitiesof the ABWRMOX spentnuclearfuelwithinthe
MPC need to be determinedto ensurestructuralintegrityduringhandlingand
transportationaccidents.
4.4.4.4 References
I0 CFR Part 961, StandardContractfor Disposalof SpentNuclearFuel and/or
High-LevelRadioactiveWaste
SAND86-2357,OGR Repository-SpecificRod ConsolidationStudy"Effecton Costs,
Schedules,andOperationsat the YuccaMountainRepository,December,1988
4.4.4-13
SAND89-7009,AlternativeConfigurationsfor the Waste-HandlingBuilding at the
Yucca Mountain Repository,August, 1990
ORNL-TM-II019,FeasibilityStudy for a TransportationOperations System Cask
MaintenanceFacility,January, 1991
PNL-SA-19567,US Programfor Managementof Spent Nuclear Fuel, April, 1991
............_'DOE/RW-O31IP,'_.A;,Monitored,RetrievableStorage.Facility:Technical Background
Information,July, 1991
SANDB7-2777,RetrievalStrategyReport for a PotentialHigh-LevelNuclear Waste
Repository.Yucca Mountain Site CharacterizationProject,December, 1991
" " PNL-8072,ForeignExperienceon Effectsof ExtendedDry Storageon the Integrity
of Spent Nuclear Fuel, April, 1992
DOE/RW-0419,1992 Acceptance PriorityRanking,May, 1992
DOE/RW-0407,Designingthe MRS, March, 1993
DOE/RW-0422, FY 1992 Annual Report To Congress,Office of Civilian Radioactive
Waste Management,July, 1993
OCRWM Bulletin,MPC ConceptualDesignReport Submittedto OCRWM, November,1993
4.4.4-14
4.4.5 COMPARISON OF U.S. AND :,NTE_'.NATIONAL
TRANSPORT REGULATIONS
No clear regulatory orenvironmental framework :exists-in .the-,US.for the ..use of recovered
plutoniumin MOX fuels, thereforetherehas been no developmentof a commercialinfrastructure
to support transport, safeguards or security for such transports. The "system" for the movement
•.........:"ofsuch materials isoperated-bythe.US:Departmentof Energy (DOE) and.the..US .Departmentof
Defense (DOD). The "system"was designed to support the transportof plutonium for fabrication
into nuclear warheads and the transport and deploymentof assemblednuclearwarheads.
The development of a commercial system will require a stated US policy for the civil use of
plutonium in MOX fuels and the development and completion of an Environmental ImpactStatement to reflect this action.
Although transportof radioactivematerialdates back to the beginningof the nuclearindustrythe
rapid development of nuclear plants and international trade in fuel cycle services such as
enrichment,reprocessingetc., have led to the evolution of an international transport infrastructure
to service the industry. In particular there is experience outside of the US of international
commercial transport both of plutonium and MOX fuel assemblies under the IAEA regulatoryframework.
Advances in package design and technology have been led by increasing emphasis on safety
assuranceand compliancewith transport regulationswhichin manycases exceed those appliedto
other dangerous goods. In the case of certain materials, security during transport has equal
emphasiswith safety in order to prevent theft or diversion of the cargo. Such security poses
special problems, and can only be described in general terms for obvious reasons.
4.4.5-1
4.4.5.1 National and International Regulations
A TransportRegulations
A review of the regulationsrelatingto plutoniumtransporthas identifiedthe following which are
mostapplicable:
National (US)a) 49CFR Pt 170 to 178 - Transportation
b) 10CFRPt 71 - Packaging of radioactivematerialfortransport and transportationofradioactivematerialundercertain conditions.
c) NUREG 0360 - Qualificationcriteria to certify a package for air transportation of
plutonium
d) US Department of Transport Specification 6M International
e) International MaritimeDangerous Goods (IMDG) Code.
f) InternationalCivil AviationOrganization(ICAO).
g) International AtomicEnergyAgency (IAEA) Safety Series.
No. 6 - Regulations forthe Safe Transportof RadioactiveMaterial.
A study of regulations a, b, e, f and g shows that these are essentially consistent in terms of
package, labelingetc. requirements,because they reflect or cross referenceto IAEA Safety series
No. 6. The basic principle of Safety Series No. 6 is that the safety of the package is vested in its
design and is therefore independent of transport mode i.e., land, sea or air. However within the
US the adoption of the mode specific NUREG 0360 together with the Murkowski amendment
places more test requirements on packages designed for air transport of plutonium. US DOT
Specification 6M is limited to packages containing very small quantities of material (less than
10Wheat output).
B. IAEA Regulations
It is important to note that the arrangementsfor transportationof certain plutonium by BNFL are
in accordance with the IAEA regulations to which the US is also a signatory. The transport of
irradiated fuel from Japan to Europe by the BNFL subsidiaryPacific Nuclear Transport LTD
(PNTL) is also carriedout underarrangements which comply with IAEA regulations.
As already stated according to IAEA philosophy the safety of the consignment is vested in the
packaging. As such they are subjected at all levels to stringent controls on design, manufacture
and operation.
4.4.5-2
Packaging varies according to the nature of the material carried and its radiological
characteristics. Complex methodologies involving dose uptake through direct or indirect
radiationcontact, pathways to the environmentetc., have been derived to classifyeach radioactive
elementor combinationof elementsaccording to hazard so as to permitthe selection of the most
appropriatetype of packaging.
In general, like the products of reprocessing, plutonium requires packages which have to
demonstratethat they retain their integrity in the most severe accident condition as defined by a
'..series of sequential tests: The so called IAEA type (B) tests involve a rigorous regime of impact
onto an unyielding target followed by an all engulfing fire to prove that the containment system
remains leak tight to the prescribedlimitsand that radiationlevels from the damaged package do
not pose an unacceptable threat to the public following such an accident.
........... Plutonium transportpackagingcalls for diversepackagingtypes to caterfor its many forms, from
powder to complete MOX fuel assemblies for fast and thermal reactors; each must serve the
conditions of safety and protection of the public. Although in very different forms, the
containmentstandardof packaging is the same.
The requirements of the IAEA Regulations, with regard to the standards of leak tightness, are
verydifficult for plutonium. With the mixtureof isotopes derived fromreprocessing of LWR fuel
the regulations restrictthe allowable leakage of plutonium dioxide to approximately 0.004 mg/h
under normal conditions of transport,andto 3 mg per week under accident conditions.
C. Package Approvals
Although this overlaps the section on package availability,it is appropriate to consider the topic at
this stage as it is clearly influencedby regulatory matters. Currently non-US packages would not
necessarily be approved by a US competent authority. However, 49CFR does allow for foreign
approved packages to be re-validated for use in the US by the appropriate US competent
authority. It also specifies that an application for re-validation must be made at least 45 days prior
to the required date for use of the package. There are no foreign packages presently qualified to
NUREG 0360 and therefore the air transport of plutonium is not seen as being applicable to this
project. The Murkowski amendment to NUREG 0360 effectively stops air transport in the US.
Consequently, only land and sea options are being considered.
4.4.5-3
4.4.5.2 Packages to Transport Plutonium or Plutonium Bearing Materials
A review of US packages has been carded out. It is doubtful if any of these could be used for
international transport as none are validated by IAEA Competent Authorities. There are no
Safkeg (2816C-16 Kg) type packages available and following the Murkowski Amendment to
" " NUREG 0360 (21.12.87), which for all practical.purposes.precluded.air,transport;of plutonium,
no developmentprogramshave been pursued.. Similarly,there are no packages in the scope of the
BNFL 1680 package.
....Tostart a plutonia campaign, the existing European packages could formthe basis for a transport
program. However, none of these are approvedfor use in the US and the US licensing process is
expensive and time consuming. Furthermore,thereare some significantdesign requirementse.g.,
double containment boundary, which have to be proven to meet licensing requirements. (It is
....... worth noting that.a Croft,AssociatesSafkeg was once validated for US use.)
Clearly for the movement of plutonia in any significant quantities a development program will
have to be started to design, license and fabricate a new generation of plutonium transport
packages. These will be doubly contained packages licensed by USNRC. They will undoubtedly
be truck or railtransported to the nearest ocean shipping port.
A search of European packages to answer an inquiry for BWR Fuel in general would not
necessarilycover the particularfuel design required.
Without more detailsof the reactorwe can only list packages for BWR Fuels of varying designs.
Firstly, there is a list of powder and metal transport packages alreadyin use for domestic and/or
internationaltransport, followed by severalfuel assemblypackages.
4.4.5-4
CONTENTS PACKAGE MAX CONTENTS(SUBJECT TO CONDITIONS)
PuO2POWDER GB/2816c/B(U)F 18 Kg MAX. *GB/3405A/B(U)F 2.5 Kg MAX. *GB/1680/B(U)F 72 Kg MAX.//
TNB/0145/B(U)F 1.5 Kg *
MOX POWDER, } GB/2816E/B(U)F 10 Kg MAX. *PELLETS }AND SMALL RODS } TNB/OI45/B(U)F 4.5Kg MAX. *PLUTONIUM METAL GB/2816E/B(U)F 4.6 Kg MAX. *
GB/3405A/B(U)F 4.5 Kg MAX. *BWR ASSEMBLIES TN 17 8 ASSEMBLIES /
EX ,4 12 ASSEMBLIES /FS 74 4 ASSEMBLIES//
• Approved to IAEA Regulations generally for all modes of Transport// Under application for IAEA Regulations/ In development stage for fresh MOX fuels. Already approved for irradiated fuel assemblies.
For PWR assembly transport the TN/0176 and COGEMA FS 69 are also used for international
transport in Europe (2 assemblies per package).
4.4.5-5
4.4.5.3 Vehicles
A. Air Transport
AirTransport- US
• As statedearlierthe Murkowskiamendmentto NUREG 0360 effectivelystops theair transportof
plutoniumin the US althoughairtransportis used effectivelyin Europe.
Air Transportof Plutonium- Europe
Transport between countrieson MainlandEurope can generally be accomplishedby road alone
but for particularly long shipments and for current shipments to and from the United Kingdom
(includingsome UK domesticshipments), airtransport is efficientlyemployed. By this means,the
timeto accomplish an internationaltransport is reducedto a few hours, during which time the
materialis removed entirelyfrom the public view with consequent security benefits. Although it
is not necessary to obtain Competent Authority package approval for countries which are
overflownby suchtransports, it is necessaryto obtaina non-scheduledflight clearancefrom thosecountriesaviationauthorities.
B. Land Transport
Road-US.
Our currentknowledge indicates that the entire production and fabrication of plutonium has been
controlledby the US DOE. Similarlythe transport of plutonium has been controlled by the US
DOE Transportation SafeguardsDivision (TSD) based in Albuquerque.
This group has been responsiblefor the transport fleet, the physicalprotection measures built into
the fleet and the armed escorts that accompany each shipment. They also provide the
management,schedulingof vehicles (tanks and trailers) trainingand maintenance. There are no
commercial alternativesinvolvedin plutoniumtransport.
We understand that Tri State Motor Transport (TSMT) used to provide high security services
with armed escorts and armored vehicles. Changes in USNRC policy with respect to safeguards
and a decliningmarket has led to TSMT discontinuing this service.
Currently TSD is providing weapon returns transport for the USDOE. If this work load declines
they may be able to provide a safeguarded transport service under appropriate sub-contract
4.4.5-6
arrangements.
Road TransportExperience- Europe.
Plutonium in its many forms attractsthe most difficultsecurity requirementsduringtransport. In
Europe all plutonium materials from powderto fuel assemblies in Category I quantities (2 kg or
more) are transported in more or less the same way, since the regulations do not currently
recognize the differencein form in which plutoniummaterialsexist; (for example MOX could be
considered as possessing an intrinsic,security characteristic resulting from .the dilution .of
plutonium by up to 20 times ina uranium matrix). Vehiclesused to transport these materialsby
road are specially constructedaccording to national standards, to provide an effective barrierto
attempts to penetration by an adversary,or by special devices to resist theft of the vehicle itself,
details of which cannot be given here. In general these vehicles present, as far as possible, the
.............same appearance as other haulage vehiclesof similartype. Irrespective of any escorting forces the
vehicles themselves are in constant communication with their operations control center and in
most cases they are tracked by automatic systems to give a. constant indication of their position
and status during transport. Special package tie-down systems are included in the vehicle
construction. Where "standard"packages are concerned it is possible to consider an integratedhandling system which is able to load or unload vehicles in despatch on receipt facilities with
minimumcloseuptake to transportworkers. During road transportsthe vehicles are accompanied
by armed escorts whose purpose is to further enhance the security of the shipment as well as
being an additional communication channel to the operations center. In some cases this escort, is
provided by a special constabulary which is specifically empowered by law to protect such
transports. They may also, in addition, be accompanied by the civil police force who can
generallyperform extra duties such as smoothing the traffic flow to allow the unhinderedtransit
of the security vehicle and its escort.
Rail
While rail transportation does provide benefits in terms of carrying very heavy loads over long
distances it also has its own unique safeguards.problems e.g., evasion of a potential threat is
difficult. It has not been possible to pursue this any further within the scope of this study.
Rail transport of Category I materials outside licensed sites is technically feasible under the
guidelines provided in INFCIRC/225 and is covered by legislation in European nations, however
it is not known to what extent actual transports are undertaken using this mode. The use of
through rail transport, like sea and air, will usually involve a road movement either at one or both
4.4.5-7
ends since many facilities do not possess a rail head or sea terminal or airport within the site
boundary and so road transport can never completely be replaced.
Sea Transport
BNFL Transport Division has many years experience of transporting .nuclear.materials by.sea
using PNTL's purpose built fleet of vessels. BNFL Transport Division.vessels.have completed
over 130 voyages and covered around 4,000,000 nautical miles.
" - The latest revision ofthe International Maritime Organization (IMO) regulations contain limits on
the quantities of, for example, MOX fuel, which can be carried by certain types of vessel. The
IMO formulated a code for transporting irradiated nuclear fuel, plutonium and high level
radioactive wastes in October 1993. The requirements of the code will be promulgated tomember states in 1994.
The code applies to new and existing ships regardless of size including cargo ships of less than
500 tons gross tonnage, engaged in the carriage of irradiated nuclear fuel, plutonium and high
level _'adioactivewastes in flasks approved in accordance with the applicable regulations for the
safe carriage of radioactive material adopted by the IAEA and carried in accordance with class 7
of the international maritime dangerous goods (IMDG) code, schedules 10, 11, 12 or 13.
For the purpose of the code, ships carrying materials covered by this code in flasks have been
assignedto three classes depending on the total radioactivity to be carried on board:
Class INF 1 - Ships carrying such materials with an aggregate radioactivity less than 4 000
TBq.
Class INF 2 - Ships carrying irradiated nuclear fuel or high level radioactive wastes with an
aggregate radioactivity less than 2 x 106TBq and ships carrying Plutonium
within aggregate radioactivity less than 2 x 105 TBq.
Class INF 3 -Ships carrying irradiated nuclear fuel or high level radioactive wastes and ships
carryingPlutonium with no restriction on the aggregate radioactivity of thematerials.
4.4.5-8
iL
All ships in addition, regardless of size, carrying materials covered by this code should comply
with Solas 1974 and the following requirements of the INF' code on:
Damage StabilityFire Protection
Temperature ControlStructural Considerations
Cargo SecuringArrangements
Electrical SuppliesRadiation Protection
Management, Trainingand Ship Board EmergencyPlan
BNFL is to classify its fleet of shipsunder the requirementsof INF 3.
The area of physical protection will require agreement between the appropriate UK and US
authorities on responsibility. Thereare a numberof contributingfactors such as type and country
of registration of vessel.
In the absence of any definiterouting requirementsit is not possible to determineany port specific
problems. However, "friendly"ports will need to be established.
4.4._9
4.4.5.4 Physical Protection Requirements
The UK has physicalprotectionplans/systemsthat:
a) complywithinternationalstandardsfor protectingnuclearmaterialin transit,and
b) are approvedby the relevantgovernmentagencies.
The internationalstandardsare:
INFCIRC/225/Rev 2: "The PhysicalProtection of Nuclear Material" and
INFCIRC/274/P,,ev1""Conventionon PhysicalProtection of Nuclear Materials"
to which the USA is also a signatory. Figure4.4.5-1 and 4.4.5-2 refer.
The principal requirementof a nuclear industryphysical protection system is to protect special
nuclear material against theft and radiological sabotage. Additionally, as a consequence of
obtaining an approved physical protection plan, that document would become a guide for the
implementationof the physical protectionregime.
The provisionof physicalprotection equipment and proceduresis dictatedby the level and nature
of the "threat" situations attributed to a process, plant or transportation system under
consideration. Detailed risk evaluation techniques exist within specialized security organizations
who, byvirtue of their knowledge, experience and relationshipwith Governmentdepartmentsand
agencies, are able to evaluate and quantifyphysical protection proposals against the "threat". An
acceptable physical protection plan has to comply with the regulatoryrequirementsoutlined in ---
above. A typical document hierarchyis shown in figure 4.4.5-3.
A. Communications
The communication network needs to be established which will define both routine and
emergency channels together with key individuals who would be involved throughout the whole
transport operation. Appropriate hardware on land vehicles.and ships will enable contact to be
maintained with an Operations Control Center.
B. Safeguards.and Security
This is an important part of the total documentation. Its purpose is to demonstrate that
satisfactory predetermined plans involving equipment and personnel (supported by back up forces
4.4.5-10
if required)can adequatelyrespondto safeguardemergencyevents.
A point to be resolvedherewill be the authorityandresponsibilityof the safeguardsand security
forces assigned to these shipments. At present, within the US, this authority and responsibility
restswith the USDOE and individualsselectedto providethe serviceare deputizedas Federallaw
"enforcementagents. Authorityand responsibility,forsea transports.wiUneed to be defined.......
C. Plannina
Thereare two aspects to the planningoperation. The first-involvesthe overallcampaignplanning.
The objective of this is to decrease vulnerability,and maximize the ability of any response forces.
To fulfill this objectivethe resultant informationfeeds into the safeguards contingency plan. The
second involves the planning for each movement which uses a codification system for sensitive
informationbeing passed between those parties involved.
Marine and land contingency plans provide essential information on sea and land transport,
engineering and health physics. This information is required to mount a response to a
conventional or radiologicalproblemon a ship or vehicle.
D. EmergencyResponseArrangements
Emergency response arrangementshave been developed for sea and land transport. Procedures
have been developedand refinedas a result of regularrealistic emergencyexercises.
Sea Transport
Although there is a high level of safety resultingfrom the PNTL ship design, it is necessary to
nlake contingencyplans for a major incidentinvolvinga nuclear fuel carrier. A full world-wide
emergencyresponse systemincludes:
a) round the clock expert adviceavailableto the ships masters;
b) ship position, headingand speed reported automaticallyto the operations center;
c) emergency response team on standbyat all times;
d) specializedequipment available;
e) equipment andtechniques developed from regular realistic emergency exercises;
f) world-wide salvagecover;
g) salvage location and telemetry system assists in vessel location and provides informationto the salvageteam regarding the condition of the ship and cargo.
4.4,5-11
LandTransport
Procedureshave been developed for emergencyresponse purposes for land transport. As for sea
transportthese have been refinedas a resultof regularemergency exercises.
Forland transport purposes the experiencegained fromthese exercises, together with the existing
emergency response documentation, would be used as the basis for defining and developing
appropriate arrangementsfor use in the US.
4.4.5-12
4.4.5.5 Licensing and Safeguards
Due to the lack of specific detailsit is not possible to answerthis directly,however the following
generalconditionsapply:
.... Regardingexport to the UK, itwould be,necessaryto know..whether,ornot the_material
would be undersafeguards alreadyin the USA and if not whetherthere was a need to keep
it out of safeguards on import to the UK. If there is no reasonto bringit into safeguards
then the US/UK Defense Agreement-couldbe used to effect the transfer (if it was .for
militarypurposes). If the material was to be used for non-nuclearpurposes then it could
possibly be brought into safeguardsand then exempted. If it is alreadyunder safeguards,
and its intendeduse is nuclear,then it will have to come underUS/EuratomAgreementfor
Co-operation,which will meanit will at all times be subjectto US controls.
As far as export licensingis concerned,a US export license will be needed irrespectiveof
safeguardsstatus. Only the authorizingagency will differ, dependingupon whether it's a
civil or non-civilpurpose. We understandthat licenses are issued at federalratherthanstate
level, therefore these should not change asa result of exports from different states. _-
We would also need to establish from the US whether they need a UK import certificate from
HMG before export is allowed as proof that the UK is in a position to receive the material. We
are unable to comment at this stage, on the conditions which would apply within or between
states in the USA as these may differ. Further advice needs to be sought to provide
comprehensive guidance on both State and Federal laws.
Overall, the biggest obstacle may not be the legal/regulatory framework within which exports take
place, but the political perspective within the US as to whether such movements are acceptable.
4.4.5-13
4.4.5.6 Routing Requirements
Without knowing specific destinations, definitive routing requirementsobviously cannot bedescribedhere.
....... In 'generalinstructionsto sailor-travel a particular.route,or _segment._would.be.transmittedby
t_ secure means. Information on the route and the _itinerary_, _ " _edas Confidential
Information in accordance with national guidelines.
Emergency arrangements,e.g., ports of call for ships,-vehicle breakdown/incidentsupport, etc.
would need to be established in a security contingency plan. These arrangementswould have to
be pre-determinedand appropriatepre-notificationrequirementsdefined.
4.4.5-14
INFCIRC/225/REVI EUROPE UNITED STATESNRC 10CFR
6.1.2(a) Shortest Journey Time / /6.1.2Co) Minimize Transfers / /6.1.2(c) Avoid Regular Schedules / /6.1.3 Code Names / /
6.2.1 Advance Notification / /6.2.2 Advance Authorization / /6.2.3 Selection of Route / /6.2.4 Locks and Seals / /6.2.5 Search of Vehicle / /6.2.6 Written Instructions / /6.2.7 Measures after Shipment / /6.2.8 Communication / /6.2.9 Emergency Action / /6.2.10 Escorts or Guards / /6.2.11/2 Advance International / /
Agreements
Mode Specific Provisions
6.3.2 Road / /6.3.3 Rail / /6.3.4 Sea / /
Figure 4.4.5-1. Physical Protection Requirements
4.4.5-15
IAEA UNITED KINGDOM UNITED STATESINFCIRC/225/REVI MINIMUM STANDARDS NRC 10CFR
Load to Secure YES ANSI or ISO
CompartmentContainer
One or more Guards: Case-by-Case YESArms not specified Provisions Armed
Type of Vessel: British Vessel Container ShipNot specified British Master
Minimi_ ports of call YES YES
Locks and Seals: YES To be InspectedInspected regularly Inspected regularly "whenever possible"
Communications: Case-by-Case Reports everyNot specified Provisions 6 hours
Figure 4.4.5-2. Physical Protection Requirements Special Provisions by Sea
4.4.5-16
Pu/MOX SHIPMENT DOCUMENTATION
hd TEROPEN TRANSPORTDCX_UMENT PLAN
II' IJ, ,- 1 J IMATRIX PLAN
• 3_dTEn . - .........(,_r_ort I
- I _L I,.,,,..lr DOCUMENTS)
PRE-SHIPMENT SYSTEMS TO PREVENTION OF TRANSPORT HEALTHPLANNING | COUNTER UNAUTHORIZED CONTINGENCY PHYSICS UNDER EMERGENCY HEALTH
DOCUMENTATION I ENTRyUNAUTHORIZEDDECEITREMOVALBY RESPONSEMANUAL MANUAL ESCORT ROUTE SITUATION PHYSlC,_MANUAL
1 ........)ETECTION OF ENGINEERING ENGINEERINGJNAUTHORIZED | EMERGENCY EMERGENCY_EMOVALBY / RESPONSE SITUATION RESPONsEEMERGENCY
;TEALTH OR FORCEI MANUAL AT LOCATION MANUAL/
TYPICAL DOCUMENT, SYSTEM HIERARCHY
5.0 SAFETY AND ENVIRONMENTAL APPROVAL
5.1 PU DISPOSITION COMPLEX S/_"ETYAPPROVALWITH TRITIUM PRODUCTION
5.1.1 Fuel Licensing Considerations
The ABWR certification provides for the operation ofalternate fuel designs provided that specific criteria onthe fuel and control blades are satisfied (NEDE 234011-P-A,Ammendment 22). These criteria evaluate not only the
thermal/hydraulic performance under transient and accidentconditions, but also stability considerations, impact on the
power/flow map and operating limits. NRC approval of eachreload is not required, provided that documentation that thecriteria are satisfied is available. NRC normally recievesa copy for information.
The ABWR SAR is based on the application of 8x8 UO 2 fuelrods. Licensing considerations for the initial MOX core isdiscussed in Section 2.3.1. Such considerations include
applicability of calculational methods and results of theanalysis.
Assuming that the MOX core has recieved prior NRC approval,operation for tritium production may still require anadditional specific submittal for NRC approval prior toinitiation of operation for the first production cycle.Additional study is required to determine whether theinitial tritium production core could be loaded withoutseparate NRC approval.
5.1.2 Potential Design Modifications
No design modifications to the ABWR have been identifiedspecifically required for the initiation of tritiumproduction or the disposal of plutonium. Therefore, theplanned certification of ABWR design satisfies the safetyapproval of the ABWR-PDR part of the complex.
5.1.3 Probabilistic Risk Assessment Impact
The Probabilistic Risk Assesment (PRA) contained in the ABWR
Safety analysis (Reference 3, Chapter 19) is based upon thecertified ABWR design. Since no design modifications arenecessary for Plutonium Disposal, this analysis remainssatisfied. The different core design and tritium productionlead to some other differences which are discussed below.Detailed review of differences should be conducted toconfirm that the PKA conclusions remain valid.
D__ecay Heat
Section 3.4.4.3 indicates that for both the plutonium
disposal cycle and tritium production cycles, decay heatwill be lower than for the core evaluated in the PRA.
Therefore, the current evaluation bounds the ABWR-PDR case
and no new analysis is necessary.
Fission Product Inventory
Section 4.5.1 of reference 1 provides a discussion of severeaccident source terms applicable to the PRA, but does notinclude a discussion of the fission product inventory. Table2.7-3 of reference 1 provides the inventory of spent fuel 5days after reactor shutdown, but does not provide anisotopic breakdown of fission products. The fission productinventory of the fuel assumed in the ABWR PKA is summarizedin Table 2A-6 of reference 3. Comparison of theseinventories with comparable values from the MOX core areneeded to confirm that bounding inventories of fission
products exist in the core evaluated in the PRA. Nosignificant differences are expected.
Fuel Pool Assist
RHR operation following startup in the Tritium Cycle willrely upon the Fuel Pool Cooling assist mode of the RHRsystem. In this operating mode, the RHR is not availablefor automatic initiation in low pressure core flooding modeor the suppression pool cooling mode. To satisfyoperability requirements of the Technical Specifications,however, manual realignment mode of the system may be
relied upon in the event that low pressure injection or poolcooling is required.
The potential for manual realignment has a slight impact onthe core damage frequency evaluated by the ABWR PRA.Sensativity studies indicate that core damage frequencywould be increased by much less than 10%. Therefore, thisimpact, although it represents an increase in risk, can beconsidered insignificant.
5.1.4 Licensing Considerations for the Disposition
Complex
Safety regulators, and therefore assumed organizationsresponsible for requirements for the disposition complex,are summarized in the table 5.1-1. Licensing considerationsfor DOE-Defense and waste management are not considered in
this phase of the project since such activities are assumedto be beyond the boundaries of the facilities underconsideration. Only the Fuel Fabrication and PowerProduction portions of the complex are addressed in thisreport.
Application of Certified ABWR Design to the ABWR-PDR
Assuming the NRC licenses the operation of the powerproduction complex, any ABWR design changes required by thePu/Li production must be evaluated against the Inspections,Test and Acceptance Criteria (ITAACs) to establish the basisfor safety approval of the ABWR-PDR portion of the complex.Since no modifications, other than the fuel design, havebeen identified, the certified design is considered to beintact for the ABWR-PDR.
In addition to submittals for the fuel design, certain sitespecific data requires approval to complete requirements foran operating license. Further, a discussion of NEPA actionsrelative to the ABWR-PDR plant as well as the fuelfabrication facility, must be considered for final NRCapproval. Additional discusion is provided in Section 5.2.
The need for initiation of a tritium production cycle withinsix months of notification places scheduling considerationson the licensing of the tritium production core. Priordesign and approval of this core would be prudent so thatonly hardware and operational considerations will need to betaken into account.
DOE Requirements
Section 4.2 of the Reference 2 summarizes the applicablerequirements for the DOE regulated portion of the complex.A regulatory line of responsibility will need to beestablished between NRC and DOE for approval of the complex.A concept to approach approval is described in Section 5.2,but the ABWR certification effort and basis needs to be
integrated into the concept.
No specific regulatory requirements have been identifiedwhich apply to the target rods or handling of exposedtargets other than DOE and NRC requirements associated withthe handling of radioactive materials. In order assure thattargets remain intact during operation and handling,fabrication quality controls are expected as a purchaserequirement, but not subject to regulation. Section 3.3provides additional information on target fabricationrequirements.
5.1-3
5.1.5 References
1. "Study of Pu Consumption in Advanced light WaterReactors", GE Nuclear Energy, NEDO-32292, May 13, 1993.
2. "Study of Pu Consumption in Advanced light WaterReactors, Compilation of Phase ib Reports", GE NuclearEnergy, NEDO-32293, September 15, 1993.
3. "Safety Analysis Report, Advanced Boiling WaterReactor", 23A6100 Rev I, Amendment 31.
Table 5.1-1
Plutonium Disposal ComplexRegulatory Agencies
II llrl
Disposition Complex Segment RegulatorWeapons Reciept Facility DOE-Defense
Target Fabrication Facility DOE-Defense
MOX Fabrication Facility DOE-Nuclear EnergyPower Production Facility NRC(ABWR-PDR)
Tritium Recovery Facility DOE-DefenseWaste Storage Facility DOE-Waste Management
, ..... , ,,
5.1-5
5.2 IMPACTOF TRITIUIt PRODUCTIONON ENVIRONMENTALAPPROVAL
As discussedin the Phase IA report (ReferenceI) the only incrementaleffects
of tritiumproductionon environmentalapprovalare relatedto the leakageof a
small amount tritium to the environmentfrom the ABWR. Since only one ABWR is
necessaryto achieve the tritiumproductiongoal, thls assessmentassumesonly
one tritium-producingreactor.
Currently,there are no federalor state regulationscontrollingthe releaseof
hydrogen-3 (tritium) to the environment with respect to chemical hazards
(hydrogengas or water/watervapor). However,becausetritium is a radioactive
isotopeof hydrogenthat could cause radiationdose to the public,releases to
the environmentare regulatedunder severalfederaland state statutes.
State regulationspertinentto this issuevary from state to state,but generally
follow the federal regulationsvery closely. For that reason, only federal
regulations will be considered in this discussion. The applicability of
regulations also depends upon the specific approval authority to operate the
facility. A facility operatingunder DOE authoritymust comply with standards
issuedby both DOE (DOEOrders) and the EPA (Code of FederalRegulations,Title
40, EnvironmentalProtection). A facility licensedby the NRC must be capable
of demonstratingcompliancewith standardspresentedin Titles 10 (Energy)and
40 of the Code of FederalRegulations. A brief descriptionof the dose limits
is as follows"
Departmentof Enerqy
DOEOrder 5400.5
• 100 mrem/year effectivedose equivalent (ede) from all DOE sources and
exposure modes
• 10 mrem/yearede from airbornecontribution(by referenceto 40 CFR)
4 mrem/year ede from drinkingwater pathway
5.2-I
• Best Available Technology for ltquid discharges to surface waters
• Best AvailableTechnologyfor liquid dischargesto sanitary sewers
40 CFR61 subpart H (National Emission Standards for Emissions of Radtonucltdes
Other than Radon from Department of: Energy Facilities)
• 10 mrem\yearede from airbornecontributionfrom entire site
NuclearRequlator.YC.ommission
10 CFR20
• 100 mrem/year total effectivedose equivalent (excludingsanitary sewer
releases)
• Annual average concentrations do not exceed Table II (10 CFR 20
Appendix B) limits at the unrestrictedboundary
• Annual external dose not to exceed 50 mrem at the unrestrictedboundary,
continuously occupied area
40 CFR 61 subpartI (NationalEmissionStandardsfor RadionuclideEmissionsfrom
Facilities Licensed by the Nuclear Regulatory Commission and Federal
FacilitiesNot Coveredby SubpartH)
• 10 mrem\year ede from airborne contribution (all radionuclidesexcept
radon) from entire site
• 3 mrem\yearede from airbornecontribution(radioiodines)
Basically, there are no regulatory requirements specific to tritium. For
environmentalapproval purposes, the impact of tritium on the environmentis
included as a component of the total radionuclideimpact. For this report
5.2-2
however, the impact of tritium released to the environment as a result of tritium
production activities ts assessed as an incremental impact.
This report ts a refinement of the Phase 1A assessment on tritium impact to the
environment. That assessment determined that, when the Incremental tritium
Impact was added to the other radtonucltde impacts, the.total impact was within
the standard levels specified above. If the impact of tritium is considered
alone, the impact was projected to be a_factor of 25belowtheregulatorylevels.
In actuality, the impact would be much less without the very conservative
assumptions used in the first assessment. Thts section introduces and discusses
those refinements that bring the impacts closer to reality.
[nvironmentalImpacts
The refinementof the tritiumconcentrationinthe reactorcoolantin Section3.2
effects the quantity of tritium released to the environmenteach year during
normal operations. Table 5.2-I provides a listing of the revised tritium
environmentalreleasequantities. These quantitieswere used to determinethe
environmentalimpact of tritiumproductionusing the ABWR.
Table 5.2-1. Tritium Release Quantities, ;, , , ..... , . ,, ,,,,, _ --., ,,
Tritium Source Operational Period Annual Quantity
Released (Ct)'1m i ' iiiii i iiiii i __ i imli i
ReactorBuilding Dur.!ngPower 30.6
During Refuel!ng 24.3
Turbine Building Dur!ng Power 346.9
Total 401.8
By far, the most restrictiveregulation pertaining to the tritium production
activity is the 10 mrem/year ede limit to the maximally exposed person.
Therefore,the impactof this activitywill be assessedagainst the regulatory
limit of 10 mrem/year. This limit is found in 40 CFR 61 in both subpartH (DOE
Facilities) and in subpart I (Nuclear Regulatory Commission -licensed
5.2-3
facilities). These regulations are better known as the National Emlsston
Standards for Hazardous Air Pollutants (NESHAPs).
These impacts were assessed using two calculattonal methods approved by the EPA
for documenting compliance to NESHAPSstandards. These are AIRDOS-PC(Version
3.0, November, 1989) (Reference 2) and CAP88-PC (Version 1.00, 1989)
(Reference 3). Due to the absence of site meteorological data, worst case
meteorology was assumed. Where the two codes gavediffering results, the highest
calculated value was used.
No liquid dischargesare expectedto containtritium;only the gaseouseffluent
from the plant is expectedto containtritium. Chemically,the releasedtritium
is assumed to be within water molecules. This assumption gives the most
restrictivevalues for healthdetrimentfrom exposure. Ingestionand inhalation
are the only two significantpathwaysintothe body and the only ones considered
in this assessment. These pathways result in only an internalradiationdose.
AIRDOS-PC determined the maximallyexposed member of the public would receive
0.0092 mrem/year ede from tritium at ]000 meters downwind from the facility
stack. This is at least a factor of 1,000 less than the 10 mrem/year limit.
However, because the airborne dose contributionfrom the entire site must be
included in the 10, no conclusionscan be drawn except that tritiumproduction
should have a very minimal impacton the environment.
When put into a risk perspective,a radiation dose of 0.0092 mrem ede would
relate to a lifetime fatal cancer risk of 3.7 E-9 for the maximally exposed
individual. Ariskto the entireexposedpopulationcannot be projectedwithout
offsite populationdata.
The Phase IA report assessedthe tritium impact at two potentialsite boundary
distances. These were at 400 meters for a standardNRC-licensedcommercialpower
plant and at 10,000 meters for a typicalDOE Site boundary. The CAP88-PC code
determinedan annualdose of 0.0014mrem ede at the 400 meter distanceand 0.0028
mrem ede at the 10,000 meter distance.
5.2-4
OccuDatlonal Impacts
Therefinement of the tritium concentration in the reactor coolant in Section 3.2
also effects the quantity of tritium released tnto the ABWRfacilities eachyear
during normal operations. This in turn dtrectly effects the occupational
exposure to factltty workers. Table 5.2-2,provtdes a ltstlng of the,revisedfacility tritium airborne concentrations. These quantities were used to
determine the occupational impact of-tritium production using the ABWR......
Table 5.2-2. Facility Tritium Air Concentrations
Plant Location Operational Period Air Concentration:c
Reactor Building Power 7.84 E-08
Duri Refuel i n_ 1.86 E-07Turbine Build' ,er 1.55 E-07
As can be seen, the concentrations are significantly below the derived air
concentration (DAC) of 2 E-05 pCi/cc and are an order of magnitude below the
tritium concentration (2 E-06 pCi/cc) that DOE requires for respiratory
protection of workers. Theseconcentrations are an average over the room air
volume. It is possible that, near the sourceof the tritium entry into the air,
concentrations could be higher. However, it is unlikely that workers would be
exposedto concentrations that would require respiratory protection.
Facilityworkerswouldbe exposedto the Table5.2-2air concentrationsfor a
limitedperiodof timeand ina limitednumberof locationsin theplant. Within
the reactorbuilding,workersshouldbe exposedonlyduringrefuelingactivities
where tritiumatoms becomeairbornein water moleculesfrom evaporationof
coolantwater. In the turbinebuilding,tritiumbecomesairborneduringpower
operationfromminorsteamleaks. Theseare the only two areaswheretritium
contaminantsare expected.
5.2-5
From the projected concentration levels and the anticipated annual occupancy in
each area, the worker radiation dose from tritium can be determined. Table 5.2-3
lists the collective dose to workers from tritium airborne contaminants. The sum
of these doses total 0.184 rein, which relates to a risk of 7.4 E-5 for incurringa latent fatal cancer.
To placethis dose in perspective, it tsassumedthat four.workers receive the
entire dose from tritium exposure (4 workers at 2000 hours per year ~ 8,500
person hours). The average dose to an individual worker is 0.046 rem per year.
This maximumdose would be two orders of magnitude less than the DOEannual limit
of 5 rems and is a factor of 12 less than the DOEannual dose goal of 0.5 rem.
This worst case scenario indicates no significant impact to occupational workers
from tritium production.
Table 5.2-3. Incremental Occupational Doses from Tritium Productioni
Location Person Tritium DAC EDEHours Concentration Fraction (Rem)
ci/cc , iill ii i
Reactor Bldg. - 5,000 1.86 E-07 0.0093 0.116Refuelin_
Turbine Bldg. - 3,500 1.55 E-07 0.0078 0.068During Power,,, i
Total 8,500 0.184
Summary
This sectiondeterminedthat the maximallyexposedoffsiteperson could receive
a radiationdose of 0.0092mrem per year fromABWR tritiumproductionactivities.
This is a very small percentageof the EPA dose limit of 10 mrem ede per year.
The impactto occupationalworkers is projectedto be a collectivedose of 0.184
rem per year. Spread over several facility workers, this is a very small
percentage of any occupationaldose standard. Therefore, with respect to
environmental,health,and safety impact,tritiumproduction in the ABWR would
presentno technicalbarrierto projectapproval.
5.2-6
References
I. Study of Pu Consumptionin Advanced Light Water Reactors- Evaluationof
GE Advanced Boiling Water Reactor Plants,NEDO-32292 May13, 1993
2. U.S.E.P.A, User's Guide for AIRDOS-PC, Version 3.0, EPA/520/6-89,035,
December, 1989, Office of Radiation Programs, Las Vegas, NV.
3. U.S.E.P.A., User's Guide for CAP88-PC, Version 1.0, 402-B-92-001, March,
1992, OfFice of Radiation Programs, Las Vegas, NV.
5.2-7
5.3 ABWR PU DISPOSITION COMPLEX SAFETY APPROVAL PROGRAM
5.3-1 Program Assumptions
The safety approval program assumed for the ABWR Plutonium Disposition Program willconsist of the submittal of a General Electric Integrated Safety Analysis Report (ISAR) and aDOE Integrated Safety Evaluation Report (ISER) safety review process. This programassumes that DOE wiI1 b¢ responsible for the overall safety approval program with input asrequired from other government safety agencies, committees or boards such as the NuclearRegulatory Commission (NRC), the Defense Nuclear Facilities Safety Board (DNFSB), or anadvisor5' board on Pu Disposition safety similar to that established for the New ProductionReactor (NPR) program.
It is also assumed that the ISAR/ISER documents will cover the entire ABWR Pu Dispositioncomplex including the mixed oxide (MOX) fuel fabrication facility and DOE approvals willallow early start of construction and operation of the mixed oxide (MOX) fuel fabricationfacility before construction/operation of the ABWR reactor facilities.
5.3-2 Integrated Safety Analysis Report Development and Submittal
The Integrated Safety Analysis P.,epon will serve as the single, primary safety evaluationdocument provided by General Electric for the Pu Disposition Complex. This document willcontain all design and analysis information required for review of the safety adequacy ofthe ABWR reactor, MOX fuel fabrication tad complex support facilities.
The ISAB, will be developed and submitted in a phased process to support DOE comprehensivereviews and conclusions that are necessary for readiness reviews conducted by DOE at keydecision points established for authorization to proceed to the next program phase. Asshown in Figure 5.3-1 these decision points include authorization to proceed with Title IIdetailed design, to begin site preparation activities, and to proceed with construction,startup and operation. Each ISAR submittal required to support these key decisions isbriefly described below.
• ISAR Submittal 1- Identification of Safety Requirements and Criteria andEvaluationMethodology- consists of ISAR sections describing the safetyrequirements and criteria and evaluation methodologies that will bc used to verifythat the required level of safety has been achieved.
• ISAR Submittal 2- Initiation of Site Preparation Activities- this submittal willsupport the start of site preparation activities and will include preliminary designinformation for all safety systems, seismic, meteorologic, hydrologic tnd geologiccharacteristics of the site, anticipated maximum levels of radiological and thermaleffluents the complex will produce, proposed major features of the emergencyresponse plan, and the level I probabilistic risk assessment,
• ISAR Submittal 3- Authorization for Substantial Construction- will integrate all thepreviously submitted material into the complete ISAR tad include information that issufficiently detailed to permit DOE to reach definitive safety conclusions. In addition,the Construction Safety Verification Plan (CSVP) including the principal verificationinspections, tests and analyses and acceptance criteria will be provided, The ISAR, atthis stage, will necessarily have less detail than contained in commercial nuclearpower plant operating license applications, but wiI1 be more detailed and completethan a typical application for a construction permit, The basic level of dotal1 isshown in Figure 5.3-1.
• ISAR Pro-Operational Amendments for Authorization to Load Fuel and PerformStartup Testing- these preoperational amendments will provide detailed informationnecessary to demonstrate the readiness for plant startup testing and the conduct of
5.3-1
operations. Revisions to previously submitted design and analysis material willreflect the as-built complex and will include the final complex technicalspecifications as well as the final accident management and emergency responseplans.
• ISAR Amendments to Support Complex Operations- amendments will be provided tosupport resolution of outstanding issues required for the transition to productionoperations. The information submitted is expected to be minimal at this point, assafety issues will have been resolved early in the program, the as-built plantreconciled against the approved design, and operational issues addressed prior to fuclload and startup testing.
5.3-3 DOE Reviews O,,'ersight Reviews and Integrated Safety Evaluation Repon
Based on the submitted ISAR material, review and audit of additional engineering and safetyanalysis material, and answers to initial DOE questions, the DOE will issue drafts ofIntegrated Safety Evaluation Report (ISER) segments in the sequence of ISAR issuediscussed above. These draft ISERs will describe the DOE safety review and conclusions,including any conditions and confirmatory items necessary to verify safety features• Thedraft ISER segments will contain open items that require resolution prior to issuance of theISER. Input to the ISER will include results of oversight reviews by safety oversightagencies _nd committees such as the NRC and the DNFSB.
5.3-4 Safety Verification Program
A Safety Verification Program (SVP) will be established to identify, document and trackopen and confirmatory items that result from the generation and review of the design andsafety analysis information contained in the ISAR and to track to completion the tests,inspections, analyses and acceptance criteria used to verify the as-built plant confomlswith the approved de.sign. This program will include a Design Safety Verification Program(DSVP) and a Construction Safety Verification Program (CSVP). The DSVP will beimplemented during the design and analysis phase of the project and will include trackingand closeout of open items requiring resolution as well as genetic industry issues arising inthe commercial nuclear power industry or from other DOE nuclear production facilities.The CSVP will be used to document and track plant construction and performanceconfirmatory tests, inspections, analyses and acceptance criteria used to verify the the as-builtplantis in conformancewith the approveddesign.
5.3-5 ABWR ReactorSafetyApprovalProgramSchedule
The ABWR reactorsafetyapprovalprogramscheduleisshown on Figure5.3-2The ISARsubmittalsare phasedto supportthe DOE safetyreviewprocessforISER developmentand tosupportkey program decisions.Durationsshown presentan aggressiveschedulebased onthenearlycompletedreviewof the GE AI3WR StandardSafetyAnalysisReport(SSAR) by theNRC. The ABWR isscheduledtoreceivefinaldesignapprovalfromtheNRC in Ib'94.Also,theABWR First-Of-AKind Engineering(FOAKE) forthe standardU.S.plantdesignhas beenstar_ed In addition, two ABWRs are currently under construction in Japan and are scheduledfor operation in 1996 and 19_1, This considerable experience in design, licensing andconstruction of the ABWR gives a measure of confidence in the proposed schedules.
5.3-6 ABWR Pu Disposition Program MOX Fuel Fabrication Facility Safety Approval Schedule
As shown in the overall ABWR Pu Disposition Program Schedule (Figure 5.3-3 ) theauthorization for start of construction of the MOX fuel fabrication facility is scheduled atapproximately the 34th month after start of design versus the end of the 36th month for theABWR reactor complex. This schedule is predicated on expeditious development of facilitydesign and early environmental and safety approvals. Again, the schedule for this facility is
5.3-2
considered aggressive but can be achieved with a national commitmcnf and proper use ofexisting technology both in the U.S and abroad. As noted ia Sections 2.2 and 4 thetechnology and infrastructure for MOX fuel fabrication at both DOE and foreign sitescurrently exists and can be drawn upon to expedite the schedules. Although licensingapprovals of some modern facilities in other countries have been extensive, it is believed thatthere are no unresolved technical issues that would delay approval of the safe operation ofthis facility on an environmentally acceptable DOE site,
5.3-3 iII
STARTTITLEII BEGINSITEPREP AUTH.C_RUCTION STARTUP OPERATION
ISARSUBMITTAL1 ISARSUBMITTAL2 ISARSUBMITTAL3 PRE-OPAMENDM'I"S AMENDMENTSi i l =
|ill
' • Safety requirements • Tille I safety • Generalarrangement - Conductofop's - Resolveopen itemsi andcriteria systemdesigns
• System functional• Analysismethods - Site interface descriptions * Summaryof startup • CompleteCSVPtests
• P&lDs• Radlologicaland - Final E-Plan
thermal effluents • Controllogicdiag.'s
u_ • Major featuresof • ALARA/radlation -Tech specs;,,,,J= E-Plan protectionprogram - CSVP.resulls
• Security plan• Lever1 PRA• Accidentmgmt plan
• Prelim. E-Plan
• DBA's
• Severeaccidentassessment
• Prel. level 3 PRA
• Drall tech.specs
• CSVP inspections,lests,analyses
,, ,,
FIGURE 5.3-1 ISAR SUBMITTAL SEQUENCE & GENERAL CONTENT
Months 6 12 18 24 30 36 42 48 54 60 (;G 72 78 84 90 93
START I ROD SEE EIrARI"PREP CONSTRUCTION FUELLOAD OPERATION
ENVIRONMENTAL EiS ] I ,_L,co_s'reuCnON I I
SAFETY I i Y AUT,ORIZA_ON I I
ISAIVISER I I I I I
SAFETYCRITERIA ISARsue ! ISERI I I ! I
sITEPREP. i tSARSL_Z ,, / wseez I I I I
CONSTRUCTION ilSAR SUlB3 ._ ISER3 I I i
AMENDMENTS [ AMENDMEN_$ ' "' • / PRE-OP' 1 1
SAFETYVIERIF. I I | I IL_
I I _ I IDESIGN(DVSP) [ DESIGN/ANALYSISOPENITEMS l
I I I I I
CONSTRUCTION(csvP) [SUBMITCSVP i JREV_WCSVP I PERFORMCON:>rlRUCTJONiPERFORMANCEVERIFICA31ON , [
t i _ I IENGINEERING lilLE I hTLE !! I COMPLETETITLE !1 J
i I I I I
PROCUREMENT l, . . PROCUREMENT I I I
CONSTRUCTION ] , i IlilLE nlSUPERVISION i I I i
I SITE_I I ICONSTRUCTION!, PREP I CONSTRUCTION t ]
TESTING/ I ISTARTUP I _E-OPTESTS SrA_TUP1
FIGURE 5.3-2 ABWR REACTOR PU DISPOSITION SAFETY PROGRAM SCHEDULE
' Activities / Year
DEVELOPMENT
SAFE'rY/ENVIRONM.
EIS
-- ConslructionSAFETYVERIF.
FUELI rCENCING
REACTOR - Title 1DEIGN
Site ConstructionOONSTRUCnON
PRE.OPTEST
OP.READINESSREV_
_, STAFTI1JPL_
OPF.RAllON
FUEL FAB PLANT
DESIC__
_ON
PRE-OPTEST
OP.REAOINESSREV.
STARTUP
OPERATION
MILESTONES
Rx Env. SI art- Reac- ReactorTil. I RE_ of torCorn- Fuel Fuelplele F'ab load
,, IIFIGURE 5.3-3 ABWR PU POSITION PROJECT SCHEDULE-ONE
5.4 ENVIRONMENTAL PERMITI'ING PLAN AND SCHEDULE
5.4.1 Introduction
This section contains a preliminary plan and schedule for obtaining theenvironmental permits and approvals required for the constructionand operation of the Advanced Boiling Water Reactor (ABWR) Pu-disposition facility at a "greenfield" site and at an existing DOE facilitysuch as the Savannah River Site (SRS). The Preliminary Schedulepresented here assumes that ABWR safety approvalsand permittingwill use Department of Energy (DOE) Orders and Regulations asprimary guidance but will also be compatible to those of the NuclearRegulatory Commission (NRC).
The Plutonium Disposition Complex (PDC) at either a greenfield site orat an existing DOE complex will consist of two co-located facilities: aMixed Oxide Fuel Fabrication (MOX) Facility and the Advanced BoilingWater Reactor (ABWR) and Power Block. The two generic sitingalternatives discussed will generally bound the schedule uncertaintiesfor the combined facility configuration.
A specific site should fall within these bounds with. the "degree ofdifficulty" of the selected site largely determined by site ownership andcontrol. In the case of the ABWR using Plutonium as fuel,implementation possibilities range from nearly impossible for a green-field site to quite possible at an existing DOE site such as the SRS,ORNL, or INEL.
At the preliminary schedule level of analysis used here, an existingDOE site offers significant schedule acceleration (two to five years),more certainty, and a lower level of complexity for the environmentalapprovals process.
The permitting process for the ABWR at an existing DOE facility wouldmostly bypass the site selection, evaluation, field data collection effort,and public involvement process required by a greenfield site. No fielddata would have to be collected. Non-nuclear permits would also beless complex and time consuming. In addition to providing a site withthe required security, skilled work force, and nuclear supportinfrastructure; a DOE site would have many required permits alreadyin place which could be modified to accept the ABWR.
Because of the extreme sensitivity of the public to siting nuclearfacilities, any greenfield site would require extensive publicinformation and relations programs to begin to have a chance to gainapproval. Some communities near existing DOE site are familiar with
5.4-1
DOE operations and are believed to be supportive of expanded siteoperations.
5.4.2 Assumptions
It is assumed that the general criteria for permitting the ABWR assumethat the facility should be licensable by the NRC even though it wouldbe DOE project and DOE rules and procedures apply...DOE and otherapplicable regulatory agency standards and codes that are morerestrictive than those of the NRC (10 CFR) are considered as applicable ....
Additionally, it is recognized that the environmental evaluationprocess for this facility will follow the requirements and guidelines ofthe National Environmental Policy Act (NEPA). The procedures forobtaining a project Record of Decision (ROD) and approval to proceedthrough the NEPA process, although not strictly a permit, requiressignificant time and resources to complete. The permits and approvalsnot strictly part of the NEPA process require a favorable ROD to becomeeffective and the conditions attached to these permits and approvalsare incorporated as enforceable conditions under the ROD.
The assumptions used to build the environmental permitting schedulepresented in this report are described below divided into four areas:general facility siting, environmental assessment process,environmental permits and approvals, and nuclear licensing andapprovals.
5.4.2.1 Facility Siting
1. It is assemed that the "Facility" will consist of the ABWR andpower block co-located with a MOX fuel fabrication plant.
2. In accordance with the DOE Pu Disposition RequirementsDocument the ABWR will be located at a "greenfield" site nearKenosha, Wisconsin.. For contrast, the ABWR could be locatedat an existing DOE facility which would provide a trained andexperienced work force, a nuclear complex supportinfrastructure and nuclear materials security from resourcesalready on site.
3. Because of the unique nature of the nuclear materials (weaponsgrade Plutonium) employed as fuel as compared to a commercialpower reactor, nuclear materials security is a primary concern.
5.4 -2
4. Generated electric power (-1300 MW) will be accepted by thelocal grid in either the greenfield site or the DOE site. The abilityof the local area to adsorb the ~1300 MW will be a siting criteria.
5. The permitting schedule will address two plant site possibilities:1) a "greenfield" site and 2) a specific DOE controlled site. Thegreenfield site will require a generic approach to permitscheduling as site specific factors effecting permitting are notdefined. The DOE site location will incorporate as much sitespecific information as necessary to define a preliminary permitschedule.
5.4.2.2 Environmental Impact Assessment Process
1. DOE will be required to prepare two Environmental ImpactStatements: 1) a Programmatic EIS for the ABWR/PuDisposition Concept and 2) a site specific EIS for theABWR/MOX fuel fabrication facility at an existing DOE facilityor at a greenfield location.
2. The preparation of these two EIS's can proceed in parallel, not inseries. Other site specific EIS's could also be prepared as required.
3. The EIS process (one or more) ;vill take a maximum of two yearsto complete (for an existing DOE site) and four to five years for agreenfield site.
4. Environmental background data for candidate DOE sites hasalready been collected and analyzed and will be made availableto the ABWR EIS team as required. Extensive field surveys willnot be required. In contrast, a greenfield site will requireextensive data collection efforts to support an EIS(s).
5.4.2.3 Environmental Permits and Approvals
1. It is assumed that the environmental permitting process willaddress the facility as a whole as described in the EIS(s).
2. Sufficient design information exists and engineering help isassumed to be available to support the various permit andapproval applications either at a greenfield site or at an existingDOE facility. A green- field site will require the collection of sitespecific data.
3. DOE will assume the role of Lead Agency and will be responsiblefor inter-agency coordination.
5.4-3
4. The Preliminary Permit and Approval Schedule will addresscurrent regulations only. Impending and future regulatoryactivity will be incorporated with time, as the PreliminarySchedule is reviewed and revised to incorporate DOE experienceand judgement.
5. The Preliminary Permit Schedule will incorporate estimates forpublic review and comment of the EIS(s) determined fromexperience. The schedule will assume a beneficial effect ontiming that will result from an effective Public Relations andInformation Program.
5.4.2.4 Nuclear Safety Approvals
1. DOE Orders and Regulations will be the primary requirementsfor licensing. NRC procedures will be applied as appropriate andthe NRC will be kept informed of all safety approval andenvironmental activities. Materials developed specifically forNRC review will benefit the overall and site specific NEPAcompliance process and expedite permitting.
2. NRC comments and concerns will be addressed by the ABWRenvironmental and EIS teams. Applicable NRC Regulatory andSafety Guides will be referenced and used as guidelines asnecessary.
5.4.3 Applicable Regulatory Standards
, Based on the permitting criteria and approach described above, thefollowing are the major regulatory policy/organizations/agencies andtheir standards that apply to the ABWR:
• NEPA - Section 102 (2) (C)
• DOE - Orders No. 5400.5, 5480.11, and 5820.2A
* NRC - 10 CFR 20, 10 CFR 51, 29 CFR
• EPA - 40 CFR 60, 40 CFR 61, 40 CFR 403 - 471
• STATE* - All applicable state regulations
• LOCAL*- All local agency standards
* - where the facility is sited.
5.4-4
This list is not complete or exhaustive, but sites the majorgovernmental agencies involved in order to indicate the depth andextent of requirements and coordination efforts necessary forcompliance.
5.4.4 Environmental and Permitting Requirements
5.4.4.1 Environmental Impact Statements (EIS)
The National Environmental Policy Act (NEPA) requires identification ....and assessment of impacts to the environment from all major projectsproposed, such as this plutonium disposition project. In order to fulfillthis requirement, the project will eventually require a comprehensiveEnvironmental Impact Statement (EIS), which will involve asignificant amount of information to be collected or developed forassessing such impact.
For a project of this type and magnitude, the following resources andissues must be addressed and characterized, for determining theproject's impact"
* Land / Geologic Resources
• Air Resources and Noise
* Water Resources
* Land Use, Recreational, and Visual Environment
• Biotic Resources and Endangered Species
• Cultural Resources
• Radiological Impacts
• Nonradioactive Hazardous Materials
• Socioeconomics
• Transportation
• Waste Management
• Decontamination and Decommissioning
5.4-5
* Decontamination and Decommissioning
NEPA requires a 90-day public comment period after publication of theDraft EIS in the Federal Register. During this period, no decision on theproposed action can be made or recorded. NEPA has a similarrequirement of 30-days after publication of the final EIS, prior to theRecord of Decision (ROD).
Realistically, the total time period, from start or Notice of Intent (NOI)to prepare EIS to obtaining a ROD, is two years on an average to fouryears or more for a controversial project.
5.4.4.2 Permits
The following are the individual permits that are needed for anABWR facility at an unspecified greenfield site. The permit list
represents a conservative estimation of the overall permittingrequirements for the facility. A DOE site, such as the SRS, would alsorequire a similar list of permits and approvals but many of these wouldbe modifications of existing site permits and easier and less timeconsuming to acquire.
• Section 404 ( Clean Water Act)/Section 10 (Rivers and HarborsAct) Permit
• National Pollutant Discharge Elimination System (NPDES)Permit
• Air Quality Construction Permit
• Section 401 Water Quality Certification
• Stormwater Permit
• Hazardous Air Pollutants Emission (NESHAP) Permit
• Permit for Construction/Operation of Domestic Wells
• Permit for Construction/Operation of a Public Water SupplySystem
• Underground Storage Tank (UST) Construction Permit
• Permit to Construct Sewage Treatment Plant
• Sanitary Landfill Permit
5.4-6
• Radioactive Waste Transport Permit
• Hazardous Waste (RCRA Parts A&B) Permit
• Notice of Construction/Alteration to FAA
• Radioactive Material Transport (DOT) Permit
• Permit to Construct Solid Waste Management System
A brief description of each permit follows, and a permit scheduledepicting the milestones in their acquisitions is attached (Figure 2).
5.4.4.2.1 Section 404 / Section 10 Permit
Generally, the U.S. Army Corps of Engineers (COE) is involved inprojects when construction occurs in a waterway or wetland. Anyproject which has the potential to discharge dredged or fill materialinto the waters of the U.S. is required to obtain a permit from the COEas authorized by Section 404 of the Federal Clean Water Act (CWA).Section 404(h) of the CWA allows transfer of administration of thispermit program to qualified states. There is provision for public noticeand opportunities for public input in the permitting process, as theregulatory purpose of Section 404 is to balance public and privatebenefits and interests against resulting impact on aquatic environment.
Section 10 of the Rivers and Harbors Act authorizes COE to regulateactivities and issue permits for projects involving construction innavigable waters of the U.S. after notice and opportunity of publicinputs, similar to Section 404.
Time requirement for review and approval is approximately one year.
5.4.4.2.2 National Pollutant Discharge Elimination System(NPDES) Permit
The NPDES permit is administered _ander Section 402 of the FederalClean Water Act (CWA) provisions promulgated by the U.S.Environmental Protection Agency (EPA). Under Section 402(b) of theAct, states can administer their own permit programs under thedelegated authority from the EPA, providing a regulatory framework toenforce standards for protecting water quality.
5.4-7
The NPDES Permit regulates the point source discharge of pollutantsinto the waters of the U.S. Typically, industrial discharges regulatedunder the NPDES program include process wastewaters, contaminatedarea drainage, and stormwater during construction. Criteria andstandards for the NPDES permit system are described in 40 CFR 125.The permit application requires information on water use, wastewaterflow, characteristics and disposal methods, planned treatment andimprovements, stormwater treatment, plant operation, material andchemical used, and other pertinent information. Depending on projectcomplexity, the processing time for an NPDES permit varies from 6 -12_months. A public hearing may be required.
The agency specifies conditions in the NPDES permit on issuance,which include technology-based effluent limitations for the waste-streams as well as water-quality based limitations for the receivingwater. New industrial facilities are generally required to meet bestconventional (control) technology (BCT) for conventional pollutantparameters (e.g., COD, BOD, TSS, pH, oil and grease), and best availabletechnology (BAT) for toxics and non-conventional pollutants. Thepermit conditions will include monitoring requirements andprovision for additional technical requirements to checkconformance.
5.4.4.2.3 Air Quality Construction Permit
For the protection of air quality, the Environmental Protection Agency(EPA) sets air pollution standards that apply nationally through CleanAir Act (CAA) and its subsequent amendments. Additionally, state andlocal governments , through air pollution control districts (agency),have broad responsibilities for implementing air pollution controlstandards and regulations within their jurisdictional boundaries.
Each proposed new or modified air contaminant source must undergoa new source review. As part of this review, PSD (Prevention ofSignificant Deterioration) applicability is determined. If PSD review isrequired (generally applicable for facilities emitting more than 100 tonsper year of a regulated pollutant, or as designated by the local agency), aPSD application must be submitted and a permit obtained beforebeginning project construction.
The air permit requires identification of all stationary sources in thefacility, type and amounts of pollutants produced, and air pollutioncontrol equipment used. The permit processing time (for review andapproval) ranges from 6 weeks (no PSD) to 6 months (with PSD)assuming no additional data collection. The construction permit asissued is generally in effect until the completion of construction, after
5.4-8
which the agency maintains compliance through the issuance ofoperating permits.
5.4.4.2.4 Section 401 Water Quality Certification
A Water Quality Certification is required for a Federal License orPermit to conduct any activity that may result in a discharge intosurface waters, pursuant to Section 401 of the Clean Water Act (CWA).The federal agency is provided a certification from the state that thesaid discharge complies with the discharge requirements of federal lawand the aquatic protection requirements of state law. Generally, it takes4 - 6 months to obtain the Water Quality Certification.
Activities requiring this certification include construction in navigablewaters and discharge of dredged or fill materials into state waters,including wetlands. COE will be the federal agency to request thiscertification, the timing of which will be tied to the Corps permitapplication review. Public notice for the water quality certification isincluded with the Corps public notice.
5.4.4.2.5 Stormwater Permit
The stormwater permit will be required to address the water qualityconcerns related to any stormwater discharges associated withindustrial activities. This permit requirement implements theregulations set forth by the Environmental Protection Agency (EPA) inSection 301 and Section 402(p) of the Clean Water Act (CWA),primarily contained in 40 CFR 122.26, and is administered by moststates under the delegated authority of the EPA. Time required toobtain a stormwater permit may vary from state to state; however, amaximum of 6 - 8 months is presently estimated.
Development of a stormwater management plan is required for thefacility under this permit, to cover both construction and operation.
5.4.4.2.6 Hazardous Air Pollutants Emission (NESHAP) Permit
This permit, administered by the EPA, covers construction of any newsource of radionuclides or modification to any existing source, underthe National Emission Standards for Hazardous Air pollutants(NESHAP) in Section 112 of the Clean Air Act (CAA). The 1990
amendment of the CAA however indicates that EPA is not required topromulgate standards for radionuclide emissions from a sourcecategory licensed by the NRC, if EPA determines that the NRCregulatory program provides an ample margin of safety to protectpublic health. States retain the right to adopt or enforce standards that
5.4-9
are more stringent than the applicable federal standards. The standardfor DOE facilities is that activities causing radionuclide emissionsshould not result in an effective dose to the public greater than 10millirems per year.
Activities causing such emissions can be the exposure of the primaryreactor coolant to the atmosphere or conditions that result in itsdegassing, and neutron flux in the spaces .adjoining the reactor vessel.Information to be furnished for this permit application includeslocation of the source, technical inputs regarding nature, design,emission estimates, treatment and control-measures to minimizereleases, list of radioactive materials used at the facility and otherpertinent details.
Prior to construction, an initial site study must be performed todetermine the offsite impact by estimating radionuclide emissions
-using an approved EPA procedure. Review and approval process fori this permit generally takes 4 to 6 months.
5.4.4.2.7 Permit for Construction/Operation of Domestic Wells
Construction, modification, and/or expansion of any domestic waterwell are activities that, if undertaken for this facility, will require thispermit. The permit is administered by the State Department of Health,and regulated under the State and Federal Drinking Water Act, andprimary drinking water regulations as applicable.
Any monitoring wells or dewatering wells do not require a permit,although a water well record may be necessary for submittal to theagency, if dewatering wells pump more than 70 gallons per minute(gpm). Also, a statement to the effect that the well will be drilled by astate-certified well driller is required to be furnished. Total timerequired to obtain this permit is approximately 3 months.
5.4.4.2.8 Permit for Construction / Operation of a Public WaterSupply System
As above, the legal authority for the administering of this permit arethe State and Federal Drinking Water Act, and primary drinking waterregulations. Construction, modification, or expansion of a public watersystem as well as its operation are covered by this permit.
For this facility, water supply systems include fire water, domesticwater, demineralized water, cooling water (makeup and circulating),chilled water and heavy water. The permit application will be requiredto furnish complete information regarding general location plans,
5.4-10
surface and/or ground water sources, water treatment plant, waterdistribution systems including improvements and appurtenances,design criteria and calculations.
Approximately 6 months will be required for review and approval ofthis permit to be issued for construction. In addition, operationalapproval must be received following construction, before placing thewater system into operation. The agency issues a written approval after.conducting a final inspection of the completed construction /modification of the water supply system.
!
During operation, chemical and bacteriological self-monitoringrequirements are generally imposed on the facility. A surface water andground water supply operation report form (including water qualityand water production information) is required to be submitted to theagency, at a frequency determined by the nature of the system and asdesignated in the permit.
5.4.4.2.9 Underground Storage Tank (UST) Construction Permit .
Installation of new underground storage tanks which will storeregulated substances (including petroleum) are covered under thispermit, the administering of which is the responsibility of the StateDepartment of Health. The objective is primarily protection ofdrinking water and environmental health and safety. The legalauthority is the Safe Drinking Water Act, and the activity is alsoregulated by the applicable State Underground Storage TankRegulations. The entire permitting process takes 4 - 6 weeks.
5.4.4.2.10 Permit to Construct Sewage Treatment Plant
Construction of an onsite sewage treatment system for this facility, forthe handling, treatment and disposal of sanitary wastewater generatedat the facility rest rooms, showers and dining areas will necessitate thispermit to be acquired. The state department of health andenvironmental control, responsible for the control of water pollutionwill administer this permit. State Pollution Control Act and applicableEPA guidelines will be the regulatory bases for this permitting process.
The application will be required to include physical site description,nature, quantity and characteristics of the waste, treatability of thewaste, details of the treatment system, point of discharge and its impacton the receiving water, and other information relevant to the proposedtreatment method as it relates to NPDES or other permits.
5.4-11
Construction permit for the sewage treatment plant typically takes 4 - 6months for issuance. However, a Permit to Construct will not be issueduntil the NPDES permit (Section 4.2.2) becomes effective. A Permit toOperate must be issued by the agency prior to startup of the sewagetreatment plant at the facility, and the level of operator required will bebased upon the classification received in the Permit to Construct.
5.4.4.2.11 Sanitary Landfill Permit
As the facility generates solid waste, both during construction andoperation, onsite disposal of such waste at a new or modified sanitarylandfill area will necessitate acquiring of this permit. Theadministering authority for this permit will be the State Department ofHealth and Environmental Control, in accordance with the existingState Landfill Regulations. Time requirements for review and approvalis approximately 6 months to a year, depending en the complexity ofthe issueand related factors, such as the public hearing process.
5.4.4.2.12 Radioactive Waste Transport Permit
Transportation of radioactive waste from the facility offsite withinand/)r out of state will require this permit, as well as a 72-houradvance, written notification of such waste shipments. Department ofEnergy (DOE) as the owner of the facility, may have special agreementwith the State Department of Health and Environmental Safety as theadministering agency of this permit, to delineate the conditions forconformance to the permit requirements. The conditions to fulfill suchspecial agreement are considered open at this time. The time requiredfor review and approval of this permit is postulated to be about 3months.
5.4.4.2.13 Hazardous Waste (RCRA Parts A & B)
Storage of any waste, designated as hazardous, for longer than 90 daysonsite, as well as any treatment and/or disposal of such waste insidethe boundary of the facility are activities that will require theacquisition of this permit. Wastes generated at the facility will need tobe characterized at the source, and categorized as either RCRA or non-RCRA wastes. RCRA and/or potential RCRA wastes will be collected inRCRA tanks and the project will have to decide if any of the aboveactivities ( storage over 90 days, treatment, disposal) will take placeonsite to trigger this permit.
The State Department of Health and Environmental Safety responsiblefor solid and hazardous waste management within the state will be theadministering agency for this permit, regulated under the legal
5.4-12
authority of the State Hazardous Waste Management Act andapplicable regulations.
Generally, Parts A and B applications will be reviewed within 60 days.A draft permit (or denial) is subject to a public review and commentperiod of 45 days, when a public hearing may be requested, prior to theDepartment's final decision.
5.4.4.2.14 Notice of Construction / Alteration to FAA
A notification will be required to the Federal Aviation Administration(FAA) for any construction or alteration in the facility at more than 200feet in height above ground level. Existence of cooling towers(especially natural-draft type) and/or stack(s) for exhaust emissions ifany will have the potential to fall in this category.
The notification will be required to include the location of the saidstructure with respect to the nearest city/town and airport(s), its heightand elevation above MSL, and other prominent terrain features in thevicinity. Unless informed otherwise, the notification must besubmitted at least 30 days before the earlier of : a) the start date ofproposed construction/alteration, or b) the date a state constructionpermit is to be filed.
5.4.4.2.15 Radioactive Material Transport (DOT) Permit
This permit will apply to the activity of transporting the plutoniumfrom its storage in different parts of the country to the fuel fabricationcomplex collocated with the ABWR in this facility. However, a specialagreement between DOE and the Department of Transportation (DOT)may be made, to cover this permit requirement by imposing specialconditions unique to this facility for compliance during the operatingphase. In that case, the review and approval time required will beexpected to be minimal. This assumption will be confirmed as moreinformation becomes available regarding this issue.
5.4.4.2.16 Permit to Construct Solid Waste Management System
This permit will be required in conjunction with a Sanitary LandfillPermit to cover construction or modification of any solid wastedisposal unit onsite. The State Solid Waste Management Regulationswill govern the requirements of this permitting process, and theadministering authority for this permit will be the State Department ofHealth and Environmental Control.
5.4-13
In addition to the general description, location, zoning, process flowetc., and type, quality and quantity of solid waste generated in thefacility, additional information to fulfill the application requirementsfor this permit include names, addresses and telephone numbers of thedesign engineer and the company official directly responsible for theplant Solid Waste Management System.
Time for review and approval of this permit may vary, from 4 monthsto a year, depending on the complexity of the system and the outcomeof public hearings.
5.4-14
GREENFIELD SITE =I
ISITING LAND FIELD DATA I:::::::::::::::::::::::::::::::::::::::::::::::::::::::::::::::::::::::::::::::::::::::::::::Ao, ::::::::::::::::::::::::::::
iI I
_ iiiiii i iii iiii i i i iiii
PROGRAMATIC EIS I SITE SPECIFIC EIS III --- -- L--._,li il
CONS'IttUC'IION
YEARSl I I i I I I
0 1 2 3 4 5 6
DOE SITE
iii iiii
SITE SPECIFIC EIS
i ii
PROGRAMATIC EIS
, r, ,l_ll, _ o. =..., i.., ... °. o ,. , .
,.'0. " ".'., ' .0" ' " ,o' °.' ' ".. ".'o'.'.' • '.' ".. " • "." °0' " •
III
CONSTRUCTION
FIGURE 5.4-1 This graphic presents two optimized permitting schedules for the ABWR
complex located at a "greenfield" site and located at an existing DOE facility such as theSavannah River Site. The view shown here is "optimized" to the extent that some sequential
tasks are conducted in parallel in order to save time. The durations of the individual tasks arepreliminary estimates based on agency requirements and past experience with similaractivities. Both scenarios assume that all activities "go according to plan" and the durationsrepresent the minim_m time to free the project for construction. In the case of the greenfield
site, the various public relations and awareness programs conducted in support of the project areassumed to be successful to the extent that public opposition is not a major factor in scheduling.
5.4-15
2 - 1 0 1 2 3 4 YEARS
I J I I ! IPROGRAMATICEIS _,,?--:_:"'-::"'_:":/-:--'"'"'_-'/_:h:h:::::l Ra3
ISITING STUDIES & SELECTION :::::::::::::::':':! i
LANDACQUISITION i:: ;: ;: :: :: ::-:: :: ;:lI
FIELD STUDIES t:-:--_'-;--'-'_:'_-'-:i-:;..ai I":.:.':.:::.":.":.:::.:.::::--':'":-".:-":':.':'":'":'-':':::-":':J
SITE SPECIFIC EIS iII
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PUBLIC WATER SUPPLY PERM I t v I_PJ_T _:.::.: .-:u, :pRE:vAi ; 6i:'IUSTCONSTRUCTIONPERM .::::::::::::::::::::::_
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SOUD WASTE MGM. SYS. PERM. I F].[.S.[.[.[.S.S.].S.].]t::_.::_.::..::_.::_.::_.::_.::..::.i_.i .i
FIGURE 5.4-2 Preliminary environmental permit schedule for an ABWR greenfield site. The major environmental permitsthat could apply to an ABWR site are listed above with time lines determined for the average case. This presentation assumesthat adverse public involvement in the EIS/permit process (which could stop the process altogether) can be kept to a minimum.The permit time lines shown in the diagram start at the time the permit is submitted to the responsible agency. Preparation ofall the permits for submittal is assumed to start about year 3 - 3.5 during the field studies program. A similar list of permits
,ould also apply for a DOE Facility site but the permit proce: be simplified because existing site Permits could beto accommodate the ABWR.
0 1 2 3 4 5 6 YEARS
, i -! I i I IPROGRAMATICEIS :_.:-"_:._.'.._.'.'-..:---_--_.._-.":--.:._.:."---:._:_::_:1RI3D
iSITING STUDIES& SELECTION _ i
SITE SPECIFICEIS ,-::.-::.-::.-::.-::...;:.-::.-i:.-::.-::._-::.-::.-::.-::.-::.-::.-::.-::.lR00I
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.......................
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FAANOTIFICATION :::::::::::::::::::::::::::::::::::::::::::::::I
SOUD WASTE MGM. SYS. PERM. ==========================================================II
FIGURE 5.4-3 Preliminary environmental permit schedule for an ABWR at a DOE site. Compared to a greenfield site (Figure5.4-2), considerable schedule compression is possible at an existing DOE facility such as the SRS. This graphic shows the mostfavorable case for the ABWR complex located at a DOE site. The rational behind a slightly over 2 year schedule is the largebody of engineering and safety information available for the ABWR coupled with essentially all the information required forenvironmental assessment and permitting is in place at a DOE site. Instead of new permits, a DOE site would involve ,medification of existing site-wide permits and all background information necessary for the environmental assessments hasalready been collected and analyzed.
6.0 DEVELOPMENT REQUIREMENTS
6.1 Development Requirements Overview:
In summarizing the development requirements in Phase 1A of this study, it was concluded
that the proposed means of dispositioning weapons plutonium by conversion to mixed
oxide and its use as fuel in an ABWR entailed no new technology and needed only
verification or validation of already existing technology that had not been used in this
country for some time. Detailed evaluations conducted during Phase 1B and 1C have only
served to confirm this position. Specifically, it has been found that:
• No system level changes are required for the ABWR to utilize MOX fuel;
• Technology to fabricate MOX fuel from Plutonium oxide is well understood although
such fabrication has not been carried out in this country in almost two decades;
• The technology that is being implemented to fabricate MOX fuel from reprocessed fuel
being implemented in other parts of the world is readily adapted to making MOX fuel
from weapons Pu while requiring less shielding and is potentially easier to handle.
By the same token, unlike reprocessed plutonium, weapons plutonium is potentially moreattractive for diversion.
Limited development/engineering verification tasks were outlined in the Phase 1A and
Phase 1B reports in the following areas:
1. Safeguards:
Disposition Process Simulation, Literature Survey
Software Development for Safeguards C&I Integration
2. Fuel Cycle
Validation of Nuclear Methods by Monte-Carlo Techniques and Benchmark Data
Confirmatory Testing for MOX Fuel Fabrication (with and without Gd)
Lead MOX Fuel Pin Tests
3. Tritium Production
6.1 - 1
No new areas requiring development or engineering verification were identified during this
phase of the study. Further detailed evaluations however were conducted in specific areas
during this Phase of the study. These include the specific development tasks needed for
lead fuel tests (given in Section 2 of this report), long-lead engineering tasks associated
with the MOX fuel fabrication plant and development tasks for tritium production. No
development tasks were found needed for the reactor system.
6.2 Development Requirements for MOX Factory
The factory for fabrication of MOX fuel is on the critical path to the disposition
process. As pointed out earlier, a certain level of safeguards implementation would have
taken place by the time the pits are converted into MOX pellets as it would take several
weeks to convert this material back to a weapons usable form. For this reason, it is
desirable to implement MOX fabrication as quickly as possible.
The technology for making MOX fuel is already being implemented in foreign
countries. Evaluations have been conducted on adapting and ! lproving this technology for
weapons Pu disposition. One area which has been identified for further evaluation
involves the surveillance equipment used for safeguards and security control. Unh_:e
plutonium from reprocessed fuel which is highly 7 active, weapons plutonium has a low
level of 7 activity. For this reason, material accountability or surveillance cannot be
dependent upon instrumentation based on _, activity per se. Alternative, positive
interrogation measures have to be implemented, including the use of Cf scanners. While
such scanners have been used in the past, their use has principally been for fully assembled
fuel rods. Their application to on-line measurements is expected to require additional
development and engineering verification. Similarly magnetic sensors have been used for
identifying Gd bearing rods and once again, some additional development and engineering
validation for application in the proposed MOX factory will be needed.
Recent advancements in both laser and fiber optic technology have been
incorporated into real time measurement systems for nuclear material. Although still in the
development phases, this technology appears to be directly applicable to the accountability
and certification measurements required during fabrication of MOX fuel. These systems
are particularly attractive because all of the electronics are located outside of the
contaminated area and fiber optics are utilized to transfer signals for evaluation. Current
development by national laboratory personnel has resulted in prototype measurement
systems for metal and isotopic compositions. Since these types of systems provide
measurement results in minutes instead of hours or even days associated with analytical
chemistry measurements utilized for MOX fuel fabrication, process flow rates can be
6.2- 1
improved and in process storage requirements reduced. These modifications in the MOX
fuel fabrication techniques are expected to • _ult in a more cost effective facility and
reduced operating costs.
Resources for preliminary engineering required for this task are estimated to be 4
engineers for a period of 9 months.
6.2 - 2
6.3 DEVELOPMENTPLANSFORTRITIUM PRODUCTIONi
6.3.1 Background
As part of the close-out of the Tritium Target Development Project (TTDP) plans
were prepared for completing development of the tritium target to support a wide
range of deployment options (Apley 1992) (Reference _1). .The development phase
of this completion plan consisted of the following six areas"
• Evaluation of deployment options
• Target rod design
• Target rod fabrication• Tritium extraction
• Handling, storage,transportationand waste characterization
• Preparationof a target rod qualificationpackage
• Lead test assembly(s)
The basic premiseunderlyingthe TTDP completionplan was that the initialphase
would be completed prior to selection of a new tritium production goal and
deploymentoption. As a result, it includedtasks such as preliminarycore and
fuel cycle design, safety assessments,evaluation of changes required to the
plant and its operatingand safety documentation,cost and scheduleestimates,
etc. for a range of plant and core designs.
However,this preliminarydesign and analysiswork for tritiumproductionin the
ABWR has been incorporatedin the PlutoniumDispositionStudy. Further, the
final core design, safety analysis, etc. needed for tritium production is
includedas part of the ABWR completioncost. Thus,the developmentrequiredfor
tritium production in the ABWR is less than that envisioned in the TTDP
completionplan.
Another important difference relates to the need for Lead Test Assemblies
(LTA's). The targetdevelopmentprogramfor the NPRwas originallyplannedsuch
that the requiredlevel of confidencein targetperformancewould be providedby
theATR tests plus the designand vendorqualificationprograms. In thiscontext
6.3-I
LTA's were considered as a tool to evaluate extended burnup oN other performance
enhancements during operation of the NPR but not as a prerequisite to startup.
The final TTDP completion plan, however, was structuredto support a broader
rangeof optionsincludingan off-the-shelfcapabilitythat could be implemented
on short noticein existingLWR's. Itwas this goal that led to recommendingone
or more LTA's to supportthe range of potentialdeployment scenarios.
The ATR tests demonstratedthat the getter-barriertargetdesignmet or exceeded
all of its designrequirements. Further,the targetrodsoperateundermuch more
benignconditionsthan the fuel rods. For example,post-irradiationexamination
of the target rod (Lanning1992) (Reference2) confirmedthat its claddingdid
not experienceany chemicalattack or mechanical interactionwith its internal
componentsand is essentiallya free-standinggas-pressurizedtube.
As discussedlater it is plannedto modifythe standardABWR fuel bundlehardware
to facilitate remote installationand removal of -3,500 target rods a year.
Under currentNRC practiceLead Use Assemblies(LUA's)are irradiatedto confirm
the performanceof fuel assemblyhardwaremodifications. LUA's do not normally
addressthe rod performanceotherthan mechanicalconsiderationssucha vibration
and wear. However, since prototypic target rods would be available from the
large lot fabricationdemonstrationit is proposedthatthey be includedin these
LUA's. This will provideaddedconfidencein the overalltarget rod performance
plus a final validationof the fabricationqualificationprogram.
LTA's should be considered for inclusion in the ABWR at startup to explore
extendedtarget burnup capability. While we believethe getter-barriertarget
design is capable of much higher burnups, irradiationof the target rods for
multiple cycles is not necessaryto meet the tritium production requirements.
As a result,extendedburnup testingof target rods has not been costed as part
of the developmentprogram.
6.3-2
6.3.2 Target Development Tasks
The following sections provide a brief description of the applicable developmenttasks and their estimated costs based on the information available at the
completion of the TTDPin early 1992. The costs have escalated to January, 1994
at 4%/yr.
{ommercia! LiAI02Pellet Fabricationpemonstr_tion- $O._M
The purposeof this task is to optimizethe pellet fabricationprocess,qualify
vendors and finalizethe pellet procurementspecification.
Provide a Benchmarked,ProductionNDE System- $I.1M
The TTDP successfullydevelopeda prototypeNDE system for verificationof the
quality of the barriercoating on the target rod cladding. This task provides
a production-scaleNDE unit for use in the large-lotfabricationdemonstration.
It is presumed that the same unit, if successful,would be used in production.
Larqe Lot Tarqet Rod FabricationDemonstration- $2.7M
Fabrication of approximately I00 complete, full size target rods, including
barrier-coatedcladding,getters,pellets,and liners. These rods would be used
to validatethe componentfabricationand rod assemblyprocesses,qualifyvendors
and finalizethe facilitylayoutsand procurementspecificationsfor full scale
production.
CompleteThe D2Z.I2 Correlation- $2.7M
Through a comprehensive series of ex-reactor experiments validate that
deuterium/tritium(D2/T2)correlationscan be reliably used to predict the
effects of time, temperatureand pressureon the permeationof tritiumthrough
aluminized barriers and the extraction from and adsorption in the target rod
materials. This data is needed to finalizethe mod31s for target rod behavior
6.3-3
during normal in-reactor operation and transient conditions and during
extraction.
Validate the Pello1;/Getter/Barrler Kinetics Model - $2,7M
Conduct the ex-reactor experiments neededto validate the kinetics model and the
performance of the integrated target rod system.
Full-lenqthD2 ExtractionTests - $3.)M
Validation of existing extraction models and assumption would be achieved by
constructing a prototypic length extraction furnace for testing target rods with
deuterium-loaded components, The results of these tests would be used to
finalize the design of the production-scale extraction furnace.
Fqll-length Breach Tests- $0.6M
Conductex-reactortransientbreachtestingof full-sizetarget rods to confirm
the safetymargin betweenexpectedperformanceduring in-reactortransientsand
conditionswhere target rod claddingfailurecould result in ejectionof LiAl02
from the rod.
Fue.l/TarqetBundleMechanicalDesiqn and TestincI - $6M
The standard ABWR fuel bundle hardware will be modified to facilitate the
installationand removalof -3,500targetrods ayear. Since the targetand fuel
rods have identicalexternaldimensionsand the ABWR fuel bundlewas originally
designed for remote reconstitutionno difficultiesare anticipated. However,
based on currentNRC practiceex-reactorflow, vibrationand mechanicaltesting
of the modified fuel bundle hardware and irradiationof six LUA's is proposed.
The LUA's would containstandardBWR uraniafuel plus target rods from the large
lot fabricationdemonstrationand would be irradiatedin existing BWR's. The
costs for this task include design and ex-reactor testing of the modified
6.3-4
subassembly hardware, fabrication of the urania fuel rods and the necessary
safety and licensing documentation.
AutomatedBundle DisassemblyHardwar@- $0.5M
The ABWR fuel bundle was designed to be disassembled remotely using manual
tooling. However,this operationwas intendedfor occasionaluse rather than a
high volume productionoperation. Experienceindicatesthat with the current
manual tooling a single crew can process four bundles per shift. While this
would be adequate using multiple crews, it is proposed to develop an semi-
automated disassembly machine to minimize worker exposure, provide higher
throughputsand minimizeoperatingcosts and the potentialfor errorsor hardware
damage.
ProjectManaqement and Restart- $2.5M
It is assumed that the target developmentprogramwould be performedby a DOE
laboratoryteam such as the PNL-Westinghouseteam that conductedthe TTDP. This
task coversthe programrestartand managementcosts for completionof the target
developmentfor the ABWR.
Compile the Target Rod QualificationPackaqe- $2.5M
This is the final deliverablefrom the target developmentprogram and provides
the basis for final design of the target rod, the fabricationand extraction
facilities,procurementspecificationsand safety review/licensingsubmittals.
It includesthe resultsof testingand analysisused to validate the target rod
design and its performance models, the basis for and results of vendor
qualification demonstrationsplus the information necessary to support the
handling, storage, transportationand waste characterizationassociated with
tritiumproduction in the ABWR.
6.3.3 Other Considerations
6.3-5
In order to extend the burnup capability of the target rod it wtll be necessary
to perform adequate non-destructive and destructive examinations of irradiated
full-length rods. Based on the post Irradiation examination results from the
TTDP (Lanntng 1992) the key examination capabilities would be:
• Visual and dimensional Inspection
• Sipping to check for cladding leaks
• _ " Neutron radiography to verify the mechanical Integrity and location of the
LiAI02 pellets
• Rod puncture and plenum gas analysis
• Rod cut-up, metallographyof cladding and internal rod components and
measurement of Li6 depletion and tritium retention in the LiAlO2 and
getters.
The first two items are fairly standardexaminationcapabilitieswhich can be
performedat a number of locationsincludinga reactorspent fuel pool. Neutron
radiographyfor the TTDP was performedin the HFEF-Northfacilityat DOE's ANL-
West site. While the TTDP target rods were only 4-feet long the HFEF-North
neutronradiographywas setup to handle TREAT loopswhich are much longer. Thus
it is conceivablethat a full length ABWR target rod could be radiographedat
HFEF-Northusing multipleexposures.
All of the destructiveexaminationsfor the TTDP were performedin the PNL hot
cells at Hanford. The in-celltritiumenclosuresnecessaryfor this work were
sized for 4-footlong rods and have been removed. However,both the PNL and ANL-
West hot cells are large enoughto installthe in-celltritiumenclosuresneeded
for destructiveexaminationof full lengthABWR target rods. Based on the TTDP
experienceit is expectedthat this capabilityfor full length ABWR target rods
could be establishedin either of these facilitiesfor about $5 million. This
presumesthat existingmetallographicand analyticalfacilitiesfor radioactive
samplescould be used which was the case for the TTDP.
6.3-6
Whtle the above target rod examinations are not unique to the ABWRor essential
to meeting the tritium production requirements, experience Indicates that if
tritium production is a long term mission it is prudent to plan for these typesof examinations.
6.3.4 References
1. Apley 1992: PNL-8142, "Tritium Target Development Project Executive
SummaryTopical Report" - Appendix G, September 1992.
2. Lanntng 1992: PNL-8133, "Final Report on the WC-1 LWRTarget Rod
Irradiation Test and Post Irradiation Examinations (Task 3), July 1992.
6.3-7
7.1 SAFEGUARDS REQUIREMENTS FOR PuAND TRITIUM TRANSPORT
Introduction and SummaEy
The previous report in this series provided a framework for assessingthe impacts of facility siting and processing technology on thesafeguards and security of various options for Pu disposition byirradiation in an ABWR. This framework actually must cover more thanthe Pu Disposition Complex alone. The evaluation must also encompassthe critical interfaces with the Nuclear Weapons ComplexReconfiguration Program for the Pu feed material used in MOx
fabrica£ion and for tritium production and recovery (if needed), inorder to provide a complete assessment of the safeguards and securityissues. The present report continues the evaluation by furtheranalyzing the safeguards and security of the siting options for thereactor and fuel cycle facilities with the objective of narrowing theoptions presented in the earlier report. Using reasonableassumptions it was possible to narrow the list of siting options fromsix to four.
The safeguards & security issues involved in the tritium productionoption are also evaluated. If the configuration of the target rodsin the production core remains unclassified, it is expected thatirradiated target rods could be shipped in LWR spent fuel casks withthe level of safeguards and security normally associated with spentfuel shipments today. If the configuration is classified, an LWRshipping cask could still be used but additional escort requirementswould apply. Shipping tritium product from an on-reactor-site tritiumextraction facility to the Savannah River Plant does not, in and ofitself, require the use of an SST. However, given the strategic andeconomic value of tritium in a single shipment and the relatively fewshipments needed, use of an SST is considered prudent.
7.1-1
Sitinu of Fuel Cvale Facilities and the Reactor
The previous report in this series (Ref.l) provided a initialframework for assessing the impacts of facility siting and processingtechnology on the safeguards and security of Pu disposition byirradiation in an ABWR. In this report analysis of the safeguardsand security (S & S) impacts of the various siting options outlinedin Ref.1, and repeated in Fig.1 herein for convenience, is continued.
In the earlier report it was found possible to express a Figure ofMerit (FOM) for the S & S assessment as:
FOM = f(CD, DP, TR, SE, t) (i)
where CD is the dispersion of the complex (i.e. the number of sitesinvolved), DP is the diversion potential of the material beingprocessed during the Pu disposition fuel cycle, TR is thetransportation risk of a single intersite transfer, SE is the SystemEffectiveness of the anti-diversion measures (probability of alarm,proper situational assessment, and neutralization) at any point inthe process and t is the time required to process the entire i00 Mg(megagrams) of excess weapons plutonium. The complex is not yetsufficiently defined to quantitatively evaluate figures of merit.However, several of the variables permit intuitive assessment. Forexample, the longer the time required to process the entire availableinventory, the greater the diversion risk. Similarly, it may bepossible to intuitively evaluate the siting options presented in theprevious report and down-select from the six theoretically possibleoptions to some smaller number. This logic is demonstrated below.
In Figure 1 the Pu flow for all options starts with WeaponDismantlement and we assume; for this report, that this operationwill always take place at the pantex Plant in the interests ofnational security. The final location in the complex is the reactorsite, which we shall call Site A. Thus the Pu flow is always fromPantex to Site A, possibly with intermediate stops. Additionally, weassume that the Pantex site is not acceptable as the site of the Pudisposition reactor (Site A), either because of limited water supply,site size and surrounding population, or the potential requirementsfor transparency in verification of pu disposition and the consequentpreuence of foreign nationals on-site. These assumptions eliminateoption 1 from further consideration.
The five remaining options can be further reduced if we invoke otherarguable, yet reasonable assumptions. We assume that transportationrisks, based on safety, safeguards, or cost are excessive if threesuccessive shipments of Category I Pu are required to deliver eachbatch of fabricated pu fuel to the reactor and that this
transportation continues over the program lifetime. Option 6 canthen be eliminated from consideration.
7.1-2
IOom.o"_ne_,-' ' | C_v._o_ !,.ea s1 ._IM.t,i Mox..r. ___ Fuel Fueled,, : ,, ___ Fabrk:atk3n OYI'ION I
SINGLE SITE
Weapons I@A3X MOX
Dismantlement Conversion Fuel Fueled OPTION Z2 SITES
Metal MOX j MOX
Conversion Fuel ,, _ ! FueledF_:_l:ion ALWR OPTION $
2 SITES
Metal MOX
Dismantlemen Conversion Fuel FueledALW_ OPTION 5$ SITES
MOX MOXMetal Fuel FueledConversion Fabdca_on ALWR OPTION 6
4 SITES
Yucca Mountain
Reposltorg I I
FIG.I FACILITY/LOCATION OPTIONS FORABWR DISPOSAL OF WEAPONS PLUTONIUM
Four options remain; three options with a single Pu shipment in thefuel cycle (options 2,3,4) and a single option (#5) with two pushipments in the fuel cycle. These options are shown in Figure 1 ofthis report and detailed as follows:
One shipment :
° (2) Metal conversion and MOx fabrication at PANTEX; reactor atSite A
• (3) Metal conversion at PANTEX; MOx fabrication and reactor atSite A
• (4) Metal conversion, MOx fabrication and reactor at Site A
Two sh iDment s_• (5) Metal conversion and MOx fabrication at Site B (SRS, RL,
LANL);reactor at Site A.
T1_ere are no other options for the fuel cycle facilities exceptforeign sites or (presumably) "politically difficult" sites likeRocky Flats or LLNL.
From a safeguards and security standpoint, this analysis demonstrates; the importance of the selection of Site A for the reactor. Pantex and
Site A are the end points for the pre-irradiation fuel cycle anddefine the shipping lanes for plutonium (Pantex m Site A) and spentfuel shipments (presumably from Site A I Yucca Mountain). Candidatesites for the metal conversion and fuel fabrication steps might beeliminated because of large shipping distances, or be enhanced by theexistence of easily modified or upgradable existing facilities.
The importance of shipping distances and locations for Site A and the
fuel cycle facilities (Site B) is emphasized by examining separately,the safeguards requirements for transport of feed material andcompleted MOx subassemblies. Shipments of Pu feed material for MOxfabrication do not appear to present a logistics problem other thanthe requirement for Safe Secure Transport (SST). Feed materialshipments will be classified either Category IB or IC SNM dependingupon the shipment form (metal is IB, PuO2 is IC) as shown in Figure2, reproduced from DOE Order 5633.3A (Ref.2). Intersite shipmentswill be made under the auspices of the Transportation SafeguardsDivision of the Albuquerque Operations Office (SST); intrasiteshipments will escorted by field element couriers or contractorsecurity forces as specified in Site Master Safeguards and SecurityAgreements required by Order 5632.2A (Ref.3).
If the fuel cycle facilities and the reactor are co-located,shipments of MOx subassemblies from the fabrication facility to thereactor may be made under the supervision of site protective forces;SST is not required. If the reactor and the fuel cycle facilitiesare not co-located, S/A shipments will be treated as Category IID SNM(Ref.4) and SST is required by DOE Order 5632.2A (Ref.3).
7.1-4
,, i iii
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fUflZ P RflDIICT._Pits, major compeflefll:l,huttoflo, Jnqotn _2 >-0.4<2 _O.2<O.4 <0.2 _S _l<S _O.4<1 <0.4re<notable metal, ndirect ly convertiblematerials
t •
III O ll-ORltOl5 NRTEII I AI, B
Carbides, oxides,solutions (>_ 25g/1),nit<area, el<., I[uoi C 2:6 ]_.2<6 >--0.4<2 <0.4 >-20 >i5<20 :_2<6 <2elements and annembl/ene
-4 alXoye and mixtures, UF,
or ur£._(_> 501 enriched|| , ii • i i i i
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Solutions (1 to 3S9/1 )pro<can rneldtlearoqp, LrLng extensive D H/A >16 >3<16 <3 H/A _SO >_8<50 <8reprocenaLngt moderate|yirradiated material, FUn,(except _aote), UF, or UF_
..J.,_Ot < SOt enrich.dL_._ - .....
_ _m mXZnZALS N/A Nix Jqla * n/A n/a nlA *llLghIy I.clcadlLlted rares,lOllltlon8 I-_ I g/l), Eu:an/um containing e Reportable quantit:lLoo e_o • Reportable quontLklto ere lepmrlLdblo< 201k U-235 |any form any Cat_ego[y I1/ rtgilcd3Loel at Q_mt|tteequanr.[r.y) Cat::egow ZY regmJrdZeaJg of amount.
amount.
1 The lower limit tar Category IV is equal to reportable quantities in this Order.
2 See paragraphs 3b end 3c for HC&/trequirements for tritium and depleted uranium.Bee
!
Fig. 3 Nuolear Haterlal Bafeguarda OategorLes "0
Subassembly shipment from Site B to Site A might raise alternativerisk issues. Each subassembly (S/A) will _eigh about 200 Kg. in thecrated shipping condition, with each shipping _rate holding two S/As.Since SST trailer loadings are limited to 10,0u_ _.(~4500 Kg.), 14SST shipments/year would be required to provide annual core loadings(324 S/A) for the two-reactor case. Commercial carriers, presumablywith larger load limits, might reduce shipment risk (directlyproportional to the number of shipments) even if escort vehicles wererequired.
It is not possible to further reduce the viable siting options usingS & S criteria alone, although option #5 may be le_ desirable sinceit involves two Pu shipments (one Category IB and one Category IID)in the fuel cycle. It does, however, provide DOE more flexibility inSiting of thereactor and the fuel. cycle facilities which arenowindependent.
This analysis has been carried out in accordance with the specificrequirements of current DOE Orders (5632.2A and 5633.3A) governingprotection, control and accountability of SNM. But current Orders donot specifically address the unique circumstances presented by theuse of Pu in ABWRs. Both Orders 5632.2A and 5633.3A provide for thegranting of exceptions in accordance with Order 5633.2 (Ref.5) suchthat more flexibility could be introduced into the SST requirementfor the shipment of MOx subassemblies. Also, it is difficult toascertain how these Orders might be modified in the future and howthey may be impacted by international safeguards requirements and thepossible needs for transparency and bilateral symmetry in Pudisposition (See Sect.7.4).
7.1-6
Safeauards & Security Reauirements for the Tritium
Production ODtion.
A preliminary evaluation has been made of the safeguards and securityissues involved with the production of tritium in an ABWR. Asindicated by footnote 2 in Fig.2 (and paragraph 3b of DOE Order5633.3A), tritium is classified as a "nuclear material of strategicimportance" although it is not a "special nuclear material"° However,as a very costly material, tritium accountability is required tohundredths of a gram (I00 Ci) (op.cit.). Quantities of tritium inexcess of 50g. are treated as equivalent to Category III specialnuclear material (SNM); all other "reportable" quantities of tritiumare Category IV.
An earlier report of this series (Ref.6) showed that, if needed,tritium could be produced in goal quantities in an ABWR using auranium-, rather than Pu-, fueled core. Table 6.3-1 of Ref.6 showsthat each tritium target rod will contain 11.4 KCi (~1.2 g) oftritium at end of a reactor cycle. Consequently a single rodconstitutes a Category IV material, while the shipment of - 1/3 ofthe annual production from an ABWR core (~1.4 Kg) must be treated asa Category III material. The target rod, having a radioactivitylevel in excess of I00 R/h at a distance of 1 meter is considered asAttractiveness Level E.
The reference concept for the tritium production option assumes thattritium target rods will be shipped from the reactor site (Site Aherein) to the Savannah River Site (SRS) for tritium extraction,purification, and storage. Based on current DOE guidance certainaspects of the target rod fabrication, irradiation, and tritiumextraction for an actual production operation would be classified butthe general configuration of the target rod would be unclassified.Thus, for purpose of this safeguards and security evaluation, it isassumed that shipments of target rods would not be considered asinvolving "classified configurations".
An alternate tritium production concept calls for co-location of thetritium extraction facility with the reactor. In this latter case,tritium gas, probably absorbed on uranium metal ("hydride beds"),would be the shipment form for tritium. Such shipments would notinvolve classified materials.
In either of the above cases, each shipment could be as large as 1/3of the annual ABWR production of -4 Kg.(4 x 107 Ci) (Ref.6). Ashipment of that size (-1.4 Kg.) has a value in excess of $20 millionand, on this basis, certainly warrants special protective measures.Irradiated tritium target rods are highly radioactive, requiringshipment in a spent fuel cask and are, therefore, inherentlysafeguarded; extracted and purified tritium gas in approved shippingcontainers is predominantly a low-level beta- emitter and not self-safeguarded.
DOE Order 5632.2A (Ref.4) provides the following baselines protectionrequirements (paraphrased) for Category III and IV quantities of SNMin transit:
7.1-7
(i) Domestic shipments of classified configurations ofCategory III quantities of SNM shall be made by Safe SecureTransport (SST) approved by the Albuquerque OperationsOffice.
(2) Domestic shipments of unclassified Category IIIquantities of SNM may be transported as specified in (I)above, "as deemed prudent and appropriate, byagreements between the Manager, Albuquerque OperationsOffice and the respective Heads of Field Elements."
(3) "Packages shall be sealed with tamper-indicating seals."
(4) Domestic shipments of unclassified Category IIIquantities of SNM not transported by SST.may be shippedeither by government-owned truck or by exclusive usecommercial carrier, or by rail. At least two escorts, atleast one with a "Q" clearance and the other with an "L"clearance must accompany the shipment.
(5) Domestic shipments of Category IV quantities of SNM may beshipped by either SST, or commercial carrier providedproper shipment traceability and package dispatch andreceipt requirements are met.
We can conclude from DOE Order 5632.2A that tritium shipments fromthe reactor to the Savannah River Site will require Category IIIsafeguards and security unless the annual production is subdividedinto unreasonably small quantities. The following restraints willlikely apply :
(i) If target rod configuration remains unclassified, or evenif the configuration is classified, target rods couldprobably be shipped in a spent fuel cask since theradiation level would make them self-protective. Anexemption from shipment by SST may be required.
(2) If tritium gas or hydride beds are shipped, SSTmay be required. As a minimum, cleared escorts will berequired.
Finally, enriched lithium-6 feed material and unirradiated targetrods (if the configuration is unclassified) are treated as CategoryIV S_4 and may be shipped by commercial carrier without escort, butwith proper notification, shipment traceability, and packagereceipts.
7.1-8
7.4 APPLICATION OF SAFEGUARDS AND SECURITY CONCEPTS
TO THE ABWR IN A PLUTONIUM DISPOSITION MODE
This evaluation was based on full compliance with existing DOE ordersrelated to safeguards and security of SNM. However, these orders areexpected to continue to evolve and they do provide for exemptionswhere appropriate. While the existing DOE Orders provide a broadframework covering all aspects of SNM protection involved in thePlutonium Disposition program, they were not developed tospecifically address some of the unique features of this program.These include the transportation of large numbers of mixed oxide fuelassemblies containing a few percent plutonium and the transportationof irradiated tritium target elements.
Other factors which are likely to influence the safeguards andsecurity plan for a US. Plutonium Disposition program include:
• Bilateral or international agreements related to the disposalof excess weapons plutonium,
• The need for reciprocity and symmetry in implementation ofthese agreements.
• Other applicable experience and standards such as the IAEASafeguards system, and
• The need for prudent practice to assure the public that thesafeguards and security measures applied are adequate.
In this context we believe it is appropriate to plan for fullimplementation of existing DOE Orders is this area but continue toexplore alternatives that may be less expensive yet provide adequateprotection.
7.1-9
7.5 Reference8
i. GE Nuclear Energy, "Study of Pu Consumption in Advanced LightWater Reactors; Evaluation of GE Advanced Boiling Water ReactorPlants, Compilation of Phase IB Task Reports", NEDO-32293, RFPDE-AC03-93SFI9681, September 15, 1993.
2. DOE Order 5633.3A, "Control and Accountability of NuclearMaterials", February 12, 1993.
3. DOE Order 5632.2A, "Physical Protection of Special NuclearMaterial and Vital Equipment", January 17, 1989.
4. USDOE, Office of Safeguards & Security, "Guide to DOE Order5633.3, Control and Accountability of Nuclear Materials, DraftGuidance", April 1990.
5. DOE Order 5633.2, "Control and Accountability of NuclearMaterials: Responsibilities and Authorities", January 29, 1988.
6. GE Nuclear Energy, "Study of Pu Consumption in Advanced LightWater Reactors", Evaluation of GE Advanced Boiling Water ReactorPlants", NEDO-32292, RFP DE-AC03-93SFI9681, May 13, 1993.
7.l-lO
8.1 Cost and Schedule
II 8.1.1 IntroductionI
The Phase 1C tasks concerning project costs and schedules included the following:
• Review and update of the structuresaccount of the Phase 1A baseline ABWR capitalcost estimate.
• Development of pre-conceptual cost estimates & schedules which address the fourprimary Phase 1C alternative cases being studied, which are: (1) disposition of 100MT of Pu in 25 years, (2) disposition of 50 MT 1% in 25 years, (3) disposition of100 MT within the ABWR license of 40 years and reactor lifetime of 60 years, (4)disposition of 50 MT Pu in 40 and 60 years.
• Development of a typical cost comparison matrix to compare costs of the projectassuming a "Greenfield" site with an operating DOE site.
• Development of preliminary transportation costs for transporting plutonium feedmaterial, new and used fuel assemblies, and plutonium waste materials to the 1%Disposition complex or to appropriate off-site-facilities.
8.1.2 Data From Available Existing Cost Studies:
Account number 21 (Structures and Improvements) of the General Electric capital cost estimate
for the ABWR reactor complex which was used as the basis for reactor capital costs for the Phase
1A is being reviewed. Information which had been developed for recent ABWR commercial
proposals, cost studies for the GE Simplified Boiling Reactor (SBWR), and the quantities
developed from the construction of the ABWRs in Japan is being used as the basis for the review.
Appropriate cost updates being made use the DOE cost guidelines as outlined in the "Cost
Estimate Guidelines for Advanced Nuclear Power Technologies," ORNL/TM-1007.
The ABWR capital costs will be updated using a cost database that is consistent with the
information being developed for the ABWR First-of-a-Kind Engineering (FOAKE) program and
other programs. The following sources of data are being utilized to update the capital costestimate for the ABWR:
8.1-1
a. Refinement of bulk commodities initially developedfrom the construction of the ABWRsin Japan (K-6/7).
b. Cost data which had been deveioped for a recentABWR commercial proposal.
c. Recently completed detailedcapital cost estimate for GE Simplified Boiling WaterReactor(SBWR) using the DOE cost guidelines as outlined in the "CostEstimate Guidelines forAdvanced Nuclear Power Technologies"(ORNL/TM-10071/P.,3)andBechtel's experiencein constructing both fossil and nuclearplants.
d. Historical data on field installation costs for equipment and materials where appropriate.
The updatedcapital cost estimate for the ABWR will be based upon the above mentioned DOE
guidelines. The level of detail is a function of resourcesandtime. Forexample,as an integralpart
of the work performedfor FOAKE,a detailedcost estimate is plannedfor completion in
June/Julyof 1994. An earlier completion date of April/May 1994 to supportthe Pu disposition
study is feasiblewith less details. All cost estimates will be presentedin the EnergyEconomic
Data Base (EEDB) format. Additional interactionwith DOE and ORNLis anticipatedin orderto
assureproperapplicationof the guidelines to the new cost data. These activities areexpected toprovidethe most creditable cost estimate to date for an ALWR.
8.1.3 Development of Phase IC Alternatives Cost Estimates and Schedules:
Capital costs for the four basic alternativecases defined above for Phase 1C are being developed.These estimates will use the revised GE ABWR one-reactor estimate discussed above as the
baseline case. Estimates for the other multiple reactor cases will then be factored from thisbaseline estimate.
Project schedules for the four Phase 1Ccases areshown in Figures 8-1 through 8-5.
Cost cash flows for the updated costs over the project duration for each of the four cases will
also be reported.
8.1-2
The approach to revenue calculations is being defined. The ORNL-developed Pu Disposition
Lifecycle Cost Analysis Program used for the DOE TRC Phase 1A Pu Disposition Study Report
of July, 1993 will be used as the basis for the calculations in order to integrate the results withORNL.
8.1.4 Development of Greenfield Costs versus Operating DOE Site Costs:
As a part of the Phase 1C scope of work, the ABWR Pu Disposition Study team has continued to
investigate the economic utilization of existing DOE facilities for the Pu disposition mission. As
part of this investigation, a comparison is being developed to show the relative program costs
expected if the Pu disposition complex was located at an operating DOE site as opposed to the
"Greenfield" EPRI standard Kenosha, Wisconsin location assumed for previous studies. The sites
being evaluated include Hanford, INEL, LLNL, LANL, NTS, ORR, Pantex and SRS.
Although costs developed for the upgrade and use of existing facilities are preliminary rough
order-of-magnitude costs, the comparison is expected to show that considerable savings in initial
program capital costs are possible by utilizing the existing facilities and infrastructure at an
existing DOE site.
8.1.5 Transportation Costs:
Transportation costs for plutonium feed material, new and spent fuel assemblies, and plutonium
waste materials are being developed.
8.1-3
Aotivltlel I Your
DEVELOPMENT
SAFETY/ENVIRON.
EIS/vne,'K*ml -oj_"SARConstr'uctka_
SAFETY VERIF.
FUELUCENCING
DESI(3N
CONSTRUCTION
PRE-OP TEST
OP. READINESS
STAFrFUP
OPERATION
OO __ DESIGN
,_ _RMCTION.i_ PRE-OP TEST
OP. READINESS
STAFFrUP
OPERATION
MILESTONES
Fix Env. Start- Reac- Fuel Fab Reactor
Tit. FIDDup of -for Plant Oper-- Fuel - Fuel Oper* Ilion- Fab - Load
FIG. 8-1 ABWR Pu Disposition ProjectSchedule- One Reactor-60 Years
DEVELOPMENT
ENVIRONMENTAL--
_o¢,rcons¢.tme_Jm'ts op_SAFETY.
I_tleI Tdle n Titte IIIREACTOR DESIGN
REACTOR NO. 1CONSTRUCTION
PRE-OP TEST
ORR/STARTUP
OPEP,ATK_N
REACTOR NO. 2
CONSTRUCTION
PRE-OP TEST
ORRPSTARTUP
OPERATION
OO
_, FUEL FAB PLANT i.ii.IIiiDESIGNCONSTRUCTION
PRE-OP TEST
ORP,/STARTUP
OPERATION
MILESTONESRx Env. Start- First Fuel Reactor
Tit. I RODup - Reac- Ptl--Crop. Fuel -tot - ComplMe
- Fab - Fuel
- Pit. -Load
Figure 8-2 ABWR Pu Disposition Overall ProjectSchedule- Two Reactor Case- 60 YEARS
' ,, Ill III
DEVELOPMENT i •,
SAFETY/ENVIRONM. J
EIS t
SAIl P4_r.Con=sl."Amendm'ts- ODer.,Design Constrt_cilon
SAFETYVERIF. ' '
,,. == liiiFUELUCENClM3
- REACTOR DESIGN -'qt_ 1 l_t_ It- r- _le i. -ii I
-pEACTOR.O._I " ( (ii i i-- OC_S_qUCTICN -- I Site Prep. Construct;onI I r
PRE-OP/ORWSTARTUP rLr-I-"l
CPERATI_ Z
REACTOR NO_2
(X_S.I_IUC_0N ! B
PRE-OP/ORR/STAFITUP i , , ,J
REACTORNO. 3
-- O(_IS_ [ ,!o_ PRE-OP/ORR/STARTUP ! _ r't
!OPERA'nON !
Cr,_-pTitle I.I1.!11DESIGN , ,
PRE-OPTEST I_
OP.READINESSREV. L-
STAR/UP _ j
OPERATION [1
.,. .,=s ,, ,t.I *i 4,I !
Rx Env. Start- Reac- Reactor _Tit. I _ "up of tor C_J_-a_,,_Corn- Fuel Fuel ComplmDplots Pal) load
I Plant II I
' FIG. 8-3 ABWR Pu Disposition Project Schedule-m
•- Three Reactor Case-25 Yearsi i l i i s a a I i l i J * i * I i t I I | | | |
A©tivltloe I Year
DEVELOPMENT
SAFETY/ENVIRONM.
EISSAR
REACTOR DESIGNREACTOR NO.1
siloPre_ConsmJc,on
ST_X:XDERATIGN
REACTOR NO. 2
STARTUPK:X_ERATIC_
REACTOR NO. 3CCNS_tJC11CN
STARTUP_3:ERA11ON
FIEACTOF_NO. 4
CChBmUCTICNSTARTUP_OPERATION
REACTOR NO. 5
oo STARTUP£)PE3_TR_
-- REACTOR NO. 6"..3CCNSmL_11_
STARTUPK)PERATIC_I
FUEL FAB PLANT
DESIGN
CCNSIW_'IlONST_
OPERAllCW
MILESTONESRee_or
FIG. 8-4 ABWR Pu Disposition Project Schedule-Six Reactor Case-2S Years
i
_ II lIt' I ill I i I _.
_ ii__ ' __,- ...........I-
,.--III
_" -- 0
ii __ i
.,..,,
innl n nnnll n L,.
(,/')RINII INI
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illll i i L-
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-- 0
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" .I_ ;_- -_ ........
" _ '" . _I,-, F,, 0
o ooo_ ,..._',--_ ....._, ,, ir ..............
_ _._ ,...........,_ y_o"i
8.1-8
APPENDIX A: COMPLIANCE OF MOX FUELED GE9 ASSEMBLY WITH
AMENDMENT 22 OF NEDE-24011-P.A (GESTAR II)
Introduction:
This report presents generic information relative to the GE9 fuel design employing MOX
fuel with less than 10% core average Pu enrichment and analyses of GE ABWR deploying this
fuel. The report consists of a description of the fuel licensing acceptance criteria as specified by
Amendment 22 of GESTAR II (NEDE-24011-P-A, General Electric Standard Application for
Reactor Fuel) and the basis for generic compliance of the GE9 fuel design employing MOX fuel
with those criteria. It is realized that these criteria were developed for standard urania fuel and the
criteria have to be evaluated for applicability to MOX fuel. The effect of Pu addition to the urania
fuel on the physio-chemical properties of the fuel have been descried in Section 2.1 of the main
body of this report. Urania fuel designs which meet the criteria of Amendment of 22 are approvedfor use in BWRs.
In some cases, satisfaction of these licensing acceptance criteria requires cycle-unique
analyses which must be performed after the core loading pattern for that cycle has been specified.
For those cases, the generic information contained in this appendix will be supplemented by plant
cycle-unique information and analytical results. This cycle-unique information will be documented
in a separate plan cycle-unique reload licensing report for each reload.
Copies of NRC safety evaluation report and proposed licensing criteria for fuel designs and
critical power correlations (Amendment 22 of GESTAR-II) as submitted to the NRC are available
and will not be reproduced here.
Evaluation;
A set of fuel licensing acceptance criteria has been established for new fuel designs,
developing the critical power correlation and Safety Limit MCPR for these designs, and
determining the applicability of previous generic analyses to these new fuel designs. Typically, the
fuel design compliance with the fuel licensing acceptance criteria constitutes USNRC acceptance
and approval of the fuel design without specific USNRC review. In assessing this procedure for
MOX fueled designs, it is necessary to examine each criterion and evaluate the manner in which the
fl-I
use of MOX fuel as opposed to the standard Urania fuel (with or without burnable poison) changes
the criterion,either qualitatively or quantitatively.
CRITERIA
1.1.1 General Criteria
I.I.I.A NRC Approved Models
"A NRC Approved analytical models and procedures will be applied."
Table 1, taken from NEDE-31917P shows a list of NRC approved methodologies. These models
and application procedures have been approved for licensing any new GE9 fuel design. Since the
only change pertains to the use of MOX fuel (that is, urania fuel with higher Pu content), the
general methodologies employed are not affected. In particular, all phenomena external to the fuel
rods, such as the mechanical performance of the assembly and the thermal-hydraulic response of
the assembly require no changes to the methodology. Changes might be needed in the future for
specific codes and these have been identified:
Nuclear methods, TGBLA and 3-D simulator may require additional verification for use of
MOX fuel with benchmark data to be generated from the lead testing program described in Section
2.1 of the main body of this report. Since Pu is a normal component of the material mix in the
nuclear analysis of the urania fuel where Pu is bred from U238, no new cross-sections are
required. Validation of the nuclear codes for MOX fuel will however be provided through the use
of benchmark data generated from lead MOX testing and cross-correlation with Monte-Carlo
Analysis.
Fuel rod thermo-mechanical codes such as GESTR-MECHANICAL, GECAP,
SAFR/GESTR-LOCA, may also need additional NRC approval, for example in respect to such
factors as fission gas release and pellet-cladding mechanical interaction. Validation will be
provided by the use of benchmark data generated from lead MOX testing and the predictions of the
applicable codes for the test conditions.
None of the other codes are affected.
I.I.I.B Lead Use Assemblies
"New Design features will be included in lead test assemblies."
Table I NRC Approved Methodologies
I II
References(See Attachment B)
Nuclesr
OEBLA (GENESIS) 10, 11, 12, 58
BWR Simulator(GENESIS) 12, 14, 16, 58
GEMINI Physics 67, 71, 15
Void/Doppler (GENESIS) 58, 73, 74
ExtendedBurr=._p 17, 18, 19,20, 65
Thermal Hydraulic
ISCOR (methods as described in GESTAR) 58, 15
GEXL/GETAB 41, 43, 44, 45, 46, 52, 58, 71
GEXL-Plus 50, 51, 53, 54, 55, 56, 57, 71
Safety IJmlt MCYR
OLCPOW 41, 42, 43, 47, 48, 49, 52, 71, 46ii
Transient Analyses
ODYN 1, 2, 3, 8, 9
GEMINI/ODYN 4, 5, 6, 7
SCAT/'rASC 58, 89
CRNC 1,% 3
REDY (cold water injectiononly) 81, 82, 83
GEBLA (GENESIS) 10, 11, 12, 58
BWR Simulator(GENESIS) 12, 14, 16, 58
GEMINI Physics 67, 71, 15
GEXL-Plus 50, 51, 71
GEXL/GETAB 41, 52, 58, 71
Smbmty
FABLE 21, 22, 58
fl-3
Table I (Continued)
LOCA_CS Performance
CORECOOL 71, 84, 85
, SAFER/OESTR.LOCA 65,67,76,77,78,S4,S5SAFE/REFLOOD 31, 32, 67
CHASTE 31, 32, 33, 37, 67
SCAT 31, 32
LAMB 31, 32
GEGAP 86
Clad Balooning 31, 32, 34, 35, 36, 37, 65
ATWS
REDY 87,SSTASC 87, 88
STEMP ST,SSVoid/Doppler 58
RodDrop Accident
"GE RDA calculationmethodology" 24, 70, 91, 92, 93
BPWS 23, 80, 94
FuelStorage andHandling
Spent fuel storage methods 23, 64
Fuel handing accident 30
Thermal.Mechanical
GESTR-MECHANICAL 24, 65
Seismic and LOCA loads 79
Extended Burnup 17
Creep-C_llapse
Bowing 28
R-4
Compliance with this criterion has been fully met. No new design features are employed in the
proposed GE9 fuel design employing MOX fuel.
I.I.I.C Post-Irradiation Fuel Examination:
'The generic post-irradiation fuel examination program approved by the NRC will be maintained."
NRC-approved fuel examination program will be used in implementing the rod tests described in
Section 2.1 of this report on lead testing program. This will include lead detection tests, such as
sipping, visual inspections, nondestructive testing of select_ rods by ultrasonic and eddy current
techniques and dimensional measurements. In addition, axial gamma scanning and destructive
examinations for fuel metallography will also be carried out on selected rods.
1.1.1.D New Fuel-Related Licensing Issues
"New fuel-related licensing issues identified by the NRC will be evaluated to determine if the
current criteria properly address the concern; if necessary, new criteria will be proposed to the
NRC for approval."
The major new fuel-related licensing issue is the use of MOX fuel. A qualitative assessment of the
effect of up to 10% Pu addition to the standard urania fuel on fuel properties and performance has
been carded out in Section 2.1 of this report. Based on this initial assessment, it is concluded that
no new fuel-related issues hz 'e been raised by the use of MOX fuel.
While the properties of the fuel and the thermo-mechanical response are affected by the addition of
Pu, it is concluded that these effects are minor for Pu enrichments to 10%. For the levels of core-
averaged enrichment levels of Pu proposed in this study, which are nearer to 5%, no changes to
the current methodology or criteria are envisioned. The required modifications to the fuel
properties are readily incorporated based on available data, and are imbedded in the current models.
Satisfactory performance of MOX fuel in this range of Pu enrichment has already been verified. A
more detailed assessment will be carried out during the following phases of the study to confirm
this finding. The results of early MOX rod tests proposed in the lead testing program will also be
used to confirm these findings.
I.I.I.E NRC Separate Review
[I-5
"If any of the criteria in Subsection 1.1 (of Amendment 22) are not met for a new fuel design, that
aspect will be submitted for review by the NRC separately."
All the criteria of Subsection 1.1 are met, however, NRC will have to independently evaluate the
findings given in Section 1.1.1.D above.
1.1.2 Thermal.Mechanical
The subsections which apply to fuel rod thermal-mechanical design are:
i. "The fuel rod stresses, strains, and fatigue lifetime usage shall not exceed the material ultimate
stress or strain and the material fatigue capability."
vi. "Loss of fuel rod mechanical integrity will not occur due to excessive cladding pressure
loading."
ix. "Loss of fuel rod mechanical integrity will not occur due to fuel-melting."
x. "Loss of fuel rod mechanical integrity will not occur due to pellet-cladding mechanical
interaction."
All these limits, with the exception of those related to fuel-melting, will be verified using (a)
analytical models and the extensive data base already available for MOX fuel properties and
performance, and, (b) by the results from the lead MOX rods tests recommended in Section 2.1 of
the report. The principal reason for the tests, even though a significant MOX fuel data base is
already available, is to fabricate fuel to given QA requirements to produce the specific fuel
microstructural features needed to ensure satisfactory thermo-mechanical performance of the fuel
rod. Fuel-melting related issues will be addresses through a combination of already available data
base as well as new thermal arrest studies, should they be warranted.
1.1.2.B (ii)Fretting, 1.1.2.B (iii) Metal Thinning
"Mechanical testing will be performed to ensure that loss of fuel rod and assembly component
mechanical integrity will not occur due to fretting wear when operating in an environment free of
foreign material."
11-6
"The fuel rod and assembly components evaluations include consideration of metal thinning and
any associated temperature increase due to oxidation and buildup of corrosion products to the
extent that these effects influence the material properties and structural strength of components."
These criteria will be met by the GE9 design and the performance is unaffected by the use of MOXfuel.
1.1.2 B (iv) Fuel Rod Internal Hydrogen Content
"The fuel rod internal hydrogen content is controlled during manufacture of the fuel rod consistent
with ASTM standards C776-83 and C934-85 to assure that loss of fuel rod integrity will not ,.:cur
dueto internal hydrating."
Compliance of this criterion, which controls the maximum amount of hydrogen allowed to be
present in the manufactured fuel rod, will be demonstrated as part of the lead testing program.
1.1.2. B (v) Fuel Rod/Channel Bow, 1.1.2. B (vii) Control Rod Insertion, 1.1.2.
B (viii) Cladding Collapse
"The fuel rod is evaluated to ensure that fuel rod or channel bowing does not result in loss of fuel
rod mechanical integrity due to boiling transition."
"The fuel assembly (including channel box), control rod and control rod drive are evaluated to
assure control rods can be inserted when required."
"Loss of fuel rod mechanical integrity will not occur due to cladding collapse into a fuel rod
column axial gap."
These criteria are unaffected by MOX usage and will be met by the GE9 design.
1.1.3 Nuclear
1.1.3 A Doppler Reactivity Coefficient
" A negative Doppler reactivity coefficient shall be maintained for any operating conditions."
This criterion has been demonstrated using approved methodology for all operating conditions.
1.1.3 B Moderator Void Coefficient
1t-7
"A negative core moderator void reactivity coefficient resulting from boiling in the active flow
channels shall be negative for any operating conditions."
This criterion has been demonstrated for all the core design options, using NRC approved
methodology.
1.1.3. C Moderator Temperature Coefficient
"A negative moderator temperature coefficient shall be maintained for temperatures equal to greater
i than hot standby using NRC approved methodology."
This criterion has been demonstrated using NRC approved methodology.
1.1.3. D Prompt Reactivity Feedback
"For a super prompt critical reactivity accident (e.g., control rod drop accident) originating from
any operating condition, the net prompt reactivity feedback due to prompt heating of the moderator
and fuel shall be negative."
The ABWR design features preclude accidents such as control rod drop. Reactivity feedback will
be dominated by the prompt Doppler feedback and the margins for a postulated reactivity insertion
is expected to be as good, if not better than urania fueled cores. Analyses will be carried out to
quantify this margin during the detailed phase of this study.
1.1.3.E Power Coefficient
" A negative power coefficient, as determined by calculating the reactivity change due to an
incremental power change from a steady state base power level, shall be maintained for all
operating power levels above hot standby."
The power coefficient is derived based on three component feedbacks, Doppler, Modertor Void
coefficient and Moderator Temperature coefficient. Since all three are negative, no further analysis
is required.
1.1.3.F Cold Shutdown Margin
"The plant shall be calculated to meet the cold shutdown margin requirement for each plant cycle
specific analysis."
The required cold shutdown margins for the ABWR plant has been demonstrated using NRC
approved methods.
I.I.3.G Fuel Storage
" The effective multiplication factor for new fuel designs stored under normal and abnormal
conditions shall be shown to meet fuel storage limits by demonstrating that the peak uncontrolled
k-infinity calculated in a normal reactor core configurationmeetsthe limits provided in Section 3
(of GESTAR-II) for GE-designed regular or high density storage racks."
The k,,. of the proposed MOX fuel designs is very similar to the urania fueled designs (there are no
system level changes) and therefore all requirements for fuel storage will be met as easily. Detailed
analyses will be carried out in the future.
New Fuel Design Licensing Evaluation (No corresponding subsection in
Amendment 22)
"Licensing evaluations of new fuel designs will include generic analyses of the ABWR plant at
limiting points of the cycle for an equilibrium loading of the new fuel design to assure that (1)
nuclear design criteria are satisfied, and (2) safety limit MCPR values are correct. In addition,
Chapter 15 safety analyses are performed for each reload application on a cycle-specific basis for
(3) limiting anticipated operational occurrences and (4) bounding accidents. The cycle-specific
plant (5) operating limit MCPR is determined and the effect of the new fuel design on previously
evaluated accidents must be reconfirmed or reanalyzed."
Compliance with the nuclear design criteria and safety limit MCPR has been documented at limiting
points of the cycle for equilibrium loading for each of the core design options presented in the main
body of the report. Compliance with cycle-specific anticipated operational occurrences, bounding
accidents and operating limit MCPR will be performed in the more detailed phases of this study
when cycle-specific data are generated.
1.1.4 Thermal-Hydraulic
" Flow pressure drop characteristics shall be included in plant cycle specific analyses for the
calculation of the Operating Limit Minimum Critical Power Ratio."
The flow pressure drop characteristics will not change as a result of the use of MOX fuel and these
will be included in cycle-specific analyses during the detailed phases of this study when cycle-
specific data are generated.
1.1.5 Safety Limit MCPR
I.I.5.A Confirmation of Applicability
"Safety limit MCPR shall be recalculated following steps in 1.1.5.B (of Amendment 22) or
confirmed when a new fuel design or new critical power correlation is introduced."
The safety limit MCPR is influenced by the critical power correlation and by bundle design
parameters which affect the bundle R-factor distribution and the core radial power distribution.
These parameters include the spacer design, assembly dimensional geometry, enrichment level and
distribution, and fuel discharge exposure. The spacer design and assembly dimensional geometry
remain the same as for conventional urania designs. For the GE9 fuel design, the recalculation of
Safety Limit MCPR will follow the steps specified in Subsection 1.1.5.B of Amendment 22 asdefined below.
1.1.5.B Safety Limit MCPR Calculation
"Safety Limit MCPR calculation will be performed under the following conditions."
i. Analysis shall be performed for a large high power density plant.
ii. Analysis shall be performed for a bounding equilibrium core.
iii. Core radial power distributions shall be selected to maximize the numberof bundles at or near thermal limits.
iv. Local power distribution shall be selected such that the largest anticipatednumber of rods will be near boiling transition.
v. Ninety-nine and nine-tenths percent (99.9%) of the rods in the core mustbe expected to avoid boiling transition.
vi. Uncertainties used in the analysis shall be the same as documented in GESTAR I1,Section 4, except for the uncertainty associated with a new critical powercorrelation. The critical power correlation uncertainty used in the SafetyLimit MCPR determination, shall be that uncertainty associated with theoperating regions that can be obtained during normal operation or duringanticipated operational occurrences."
A-18
Safety limit MCPR is not expected to change as a result of the use of MOX fuel. However, withany new fuel design, it is customary to recalculate the safety limit MCPR and this will be carriedout for the specific MOX fuel designs (to be chosen) for disposition.
1.1.6 Operating Limit MCPR Licensing Evaluation
I.I.6.A Cycle.specific Analyses
"Plant operating limit is established by considering the limiting anticipated operational occurrences
for each operating cycle."
Cycle-specific analyses will be conducted during the detailed phases of the study when such cycle-
specific information will be considered.
1.1.6. B Generic Analyses
"For each new fuel design, the applicability of generic MCPR analyses described in Section 4 (of
GESTAR II) or in the country specific supplement to this base document shall be confirmed for
each operating cycle or plant-specific analysis will be performed."
Operating cycle/Plant-specific analyses will be performed during the detailed phase of this study.
The applicability of the MCPR analyses will be demonstrated.
1.1.7 Critical Power Correlation
I.I.7.A New Fuel Design Features
"The currently approved critical power correlation will be confirmed or a new correlation will be
established when there is a changed in wetted parameters of the flow geometry; this specifically
includes fuel and water rod diameter, channel sizing and spacer design."
No new fuel design features involving the wetted perimeter or other hydraulic parameters arise as a
result of using MOX fuel.
1.1.7.B New correlation data
"A new correlation may be established if significant new data exists for a fuel design."
No new c,orrelation arises as a result of using MOX fuel with the GE9 bundle design.
1.1.7.C Critical Power Correlation Calculation
Not applicable as no new correlation is involved.
A-II
1.1.8 Stability Licensing Acceptance Criteria
"The new fuel design must meet either of the criteria described below:"
I.I.8.A Comparison with previously e Jproved designs
"The stability behavior, as indicated by core and limin'ng channel decay ratios, must be equal to or
better than a previously approved GE fuel design."
Stability analysis has been conducted for the proposed designs and the allowable flow-power
operating region during startup set to satisfy this criterion.
1.1.9 Overpressure Protection Analysis
"Adherence of the ASME overpressure protection criteria shall be demonstrated on plant cycle
specific analysis."
Bounding analysis confirming that the overpressure protection criteria has been met is reported in
the main body of the report for the equilibrium cycle conditions. Cycle-specific analyses will be
carried out when such information is considered during the detailed phase of this study.
1.1.10 Loss.of-Coolant Accident Analysis
1.1. IO.A Emergency Core Cooling System Criteria
"The criteria in IOCFR 50.46 shall be met on plant-specific or bounding analyses."
The emergency core cooling system (ECCS) criteria in 10CFR50.46 are met by the expsoure-
dependent maximum average planar linear heat generation rate limit in bounding analyses.
Compliance has been demonstrated for the proposed fuel designs. GE will continue to
demonstrate compliance with these ECCS crtieria for specific fuel designs used in the disposition
project.
I.I.10.B Plant MAPLHGR
"Plant MAPLHGR adjustment factors must be confirmed when a new fuel design is introduced."
A-12
PlantMAPLHGR issometimesadjustedfora specificoperationalconfigurationorregion.GE
will confirm the revised MAPLHGR limit before each cycle operation.
1.11 Rod Drop Accident Analysis Licensing Evaluation
Rod drop accidents are precluded by ABWR design features.
I.II.A Cycle Specific Analysis
"Plant cycle specific analysis result shall not exceed the licensing limit described in GESTAR H."
(Plant cycle specific analysis will be conducted to show that the Ganged Withdrawal Sequence for
the specific reload is in conformance with all accepted limits. However, rod drop accidents are
precluded by ABWR design features.)
1.1 I.B Bounding BPWS Analysis
"Applicability of the bounding BPWS Analysis must be confirmed."
(See comment under 1.11.A)
1.1.12 Refueling Accident
"The consequences of a refuel accident as presented in the country-specific supplement or the plant
FSAR shall be confirmed as bounding or a new analysis shall be performed (using the methods
and assumptions described in the country supplement) and documented when a new fuel design isintroduced."
The activity released to environment during a refueling accident is principally related to the
exposure and not the Pu content. Therefore, the consequences will be similar to a urania fueled
core and compliance with applicable limits will be demonstrated.
1.1.13 Anticipated Transient without Scram
"The fuel must meet either of the acceptance criteria described below:"
A-13
1.1.13.A Void Reactivity Coefficient Range
"A negative core moderator void reactivity coefficient, consistent with the analyzed range of void
coefficients shall be maintained for any operating conditions above the stratup critical condigtion."
The GE9 design has been generically analyzed and found to satisfy this criterion if the void
coefficient falls within the range of -2.5 to -11.6 cents/% voids. The maxium void coefficient for
the reference fuel design case was -13 cents/% voids, slightly more negative than the range over
which generic analyses had earlier been conducted.
1.1.13. B Plant Evaluation
"If the preceding criterion is not satisfied, the limiting events will be evaluated to demonstrate that
the plant response is within the ATWS criteria."
Background analyses for the ABWR have shown that the ATWS presuurization transients are not
the most limiting. Specific overpressure bounding transient analyses have been conducted
showing compliance will applicable peak vessel pressure limits. Such a bounding transient
involves the sudden closure of Main Steam Isolation Valve (MSIV), the normal MSIV position
scram is assumed to fail - the reactor scrams on high neutron flux. The details of this analysis are
contained in the main body of thi_ report.
R-14
APPENDIX B: REPOSITORY CONSIDERATIONS:
10 CFR PART 961, Article Vl. Criteria for Disposal
__poe_0.
A. Qmmsd J_qwtmUt. _ d. I_el may trove "5PatJedreef* mxt/armwnd "5__ Fuel" ealUlmt_v(s) _zospt u _ i_ed In t/sis
emu'_ DOu ,_ uomt hereunder ULJy a. _ _ end8uh:hue_SeeU_--su_ BIO' ud/_.i.K,W.._eh meow t&eOermx_l0_mtJl_n, for _ _ amd Oemuy_,lpael/lma_wuW u mt for_ lnaM_adlx B, anneud 1. _ Now_e_ P_stcal fHwum.liereto imd msl i_pert tus_of.
Cn) _ 8tmn ueurdely elmOrlINT and/or HLW _ to delivery In ee. i| |
L ,Pl_n_. -_ ...........
taned dmnc_e ot the 81f end/or Him Je_, _,_,_ u _ ,J_,.--I ,, J,,_o_,L_i_____ Ol8 lieaJ0ns-_ IJInctm8X0 iSSuemuten_--m¢ forthtn_ r. mmend ..... J_,L ] _,0tunto sad msde a mr_ ttmwt, Pun_sssr____Rd__ eueu,uKtn mid 81Q' and/orIRZ_Wu man u ttzey nmu:_ al i_=umleumbecome tnewn to me pm_tum_, w _ m 0_mm_Jr_-_m ._ um_j
Co)DOB'8eblilptlou lot' dbpmtn8 of Nun,
than etendud fue_ herons, tot 8m__ nen4s tn_udtns, but not ISmtt_tto, eentroli _ been__ bY the _ meiden_burnable polmmrod mmemblJm,ran.eruothertbms4mubrdfuei, uUm_teem I_l rod elemeatL Udmble _uSm, filmb dsdlned In _ _. the _ etmmoe_ ud mtmary end meoadu_ neu-lli____ trmssom_ tlmt s_ amtma_ mtl_ tim
mtULUedviae Purehmer u_bin jUt_ (04)) Intelrad _ of the _ miemtly. _mktt douo_ _eqldre _ tnendt_, rely _ Jaetud-_ _r _ _ _ _ r_queerN to tho teeeddcslfmfutllty of dis. ml u 94trtair tho spent nvrJeu'hid de/h.erul toe 6i/mml mnmmn¢qtolid_ eonlru_IPOSLn_of amebrue1m_tho ems,eull_ ,nereed IPle_z:Puml_ does,,oil,m_etUNt au_ecJ.
let eueb nr54e_ Puei--CluJ NS-L_s.tm_z E a, Ceeft_. _ ndntmumeooU_ ttme for
fuel _ live (8)Oeae_/,Ir_io_a N_ _ _ dim notmeetthis8pecUt.
es_lon 868_ be etmmifled 88 Hmwhmdardd, _ _ ldilsltfluiffos _
1. _o__ s_ usereuon. _ _o_ _ _el _ e_er _umlibleeflor_t, utUIMn8_ equlssde_rt, II_R pogey llclllUq_ IdudJIbe dMmlftedhis_. _o_ _ _ __ peru. N_ __i ]18.4. 8ueh fuel
flee. to _ dsm_ 80_t N_ _ _ be _ue and _ 8eee_! tumdlmur.
(81q_) prior to deMver_te DOE, u lol/omcaKoiwse,and l lamlllUee,
_mul_l JPaaelmmun 6N_Pthat mee_ 11.(3mup_ _ Roal_IPWlwhichhuilL1the 131_rterld_l_m __ wt beendll_aembled8_d 8Lmredw4U3LhefualIov_hIn pau'snrrsphB belew, rods In • eoeuoik_ted rammer_QI _
laOtmeet otw or unoveof th_ Oeneml _I- e. P_8_fedYueLse_ focth U_ 8ubl_rsSr_J I 8, Vlsual I_or_
threuSb 6 o! _h B bdou. gmcli __ _ bev_tadl_, _ for18lid S NU_ _el _ evidenceof shrucq_nd defo_ _ _eMH-ItJ_ou4_ _ pmmmnt to _ to timing or m_ers uldeh n_3_ rectuireB below, slmdld haindllhll.Assemblies whJei_(|l 8xe
C, ,eaffed _ me_r_ 8NF Um_ mNtt, _ 81a3_e_Undlydoformedor hive danmgedspecaflmUons ut forth tn 8ub_smemptu I et_dtr_ W _ extent tJ_st_il _QridlingII_h $ ol ps_ph B below, Jmdi _ be usqulred or Ill] for 8my _n
_ u _ _I _ _-L _ be _ed _th _tl I_I him.U515ml_F4 _t tO___ e of _ equlprnerlt_tll be ehum_fiedlUl._I_penGrnph B Ibelow. IPtx__ F-I.
B-1
is. Ptmvimss_ _ted Ammmbitee. UL _ Jut], dlatorttom, cinddtnz,/bmRblies eueupmJle_ed by _ deme_ m. o_er dmmmle to the spent,fuel,
pcJor to cli.uUkmUon hefmmdlr Jhd! k or nonfud eoatl_nenla wit.bin U'.ts 8h_pirtirdtJdtlul u _ lqLel--OuJ r-8. Put- bit W_ will rtqulre JiJectt_ hudilnt J_m.ebsewr shaU sdv_ OOB of the _mson for eedurea. (Al_,ach 84fdft0OllA|pales if needed.)the I_m' enespsudsUon of Msemblleu in go/6 ........liclen4 detail so _ DOE may plan for 8_ = .......prOl_ldm attbaequel_,hJxldlk_, IV, Amemb4y-iRlimber----
o. Relulsto_ Requlresmmta. 8hlpotn8 lot #---8peril fuel mm.mbUes _nsll be pedmaed
end nbzed In etmkn so thst, sll sppil_blereaulutory requlrememt8sre reeL. . .
1./Jlandmnf PuzL' • l_L CIm8 8-1: -I--l.--l.-.l.--l-..-b. Ctus 8-1: BWR _.8tmw iw (mwew_)1. Nmukndm_ JL_j_r: t._ em (.wcJm_ '!".... !--.-t-.--'t ..... |'--
s. el..8.-I: _lcul Dlme_,, 8._. __....J_ l 1 )b.am Ns-_,J_on_ com_ner_ 4._ _ row,,ms**_c.:..j.__.C:: "e. Cbt.u H6-3: 8holq, Coeled J, ImaJ_ eetm_tLmmmJyInm_ v_ _ qqm_Jd. Clam Ha-4: Hon.bW'R mWelaml meu_t ------- m e_0me ------e. Cims N8-5: Como.d_ted 3PuelRods. .......S./FeU_ FUEL"it,.CIm F-U VJmml PaJlurrt e_ I)sma_ Any lslSeo n_lUou8 or frsdulenL at_te-b. Ctus I:'-_ Rsd_os_We "16esk_e" men&m_¥ be plmhdltble by fine or ImDrls.
Ch._ P-3: ICncspmJst,od oemumt (U.8. Code. 'r'iUe L8. l_t, ton lOOI).
D, HWh.Lnel RadCa_t(_ Weufe B_ Pur_luumn
The DOE st_dl ecce_ tdah.levet rmdtotc. 8la'naCureWe wute. De_lled 8ece_ta,nce er_14t, and Title8enera/a, pecU'(cst,ions for tmch waste will be DateJsmsed by Che DOE no later thJm the dMe on_hh:h DOB sulbmlla 114llcetsse o4ppl_c_UontO the Nuclear Rel_at,0r_ C_Jm_sdo_ forUse fln_ dtspoeal IselllLy.
'Am,mmaz P
/)e£eUed/)esc_pfh_4t oy'PuyrTt4serV
lmrormsMon Mudl be jurm4ded byl_rcJ_=er for esch dieUn_ fuel type _t&hln• 81hlpplrurLo_ no(. later tlum 81xf,y (80)cht,ym prior Lo the _edule LmM_rUt_Joodate,.Purehme_Con_nu:&Number/I_te ----t--------Rnctor/lP_tcJllt, Y _sme--
I. D_wln_ tn,'_uded _ 8enertc doutec.mmmmm._Im
t. lPuel ,As_.mt_ DWGI ---2. ODl)e_& l_mer ewd llt,Unls DWGI --Domt_ lqumber: ---DOE _bl4ppinf:._tf: -----I AmembJtesDeecrlbed:
...---PWR
.----OUner
[I. Deskm M_'#_ Desc_Mions.Jh_dJteme_tL"
1. Elemen_ tylX ----(Js_, plal,e, etc.)I. _04sl ler_h' --/tin.)L Active lenl_ --(in.)4. C]tddlng mttertal -- (7_",8.8.. etc.)
,JuemMv Deecr(_tton."
I. Humber o! Etemen_ ----2. Overall dtm.-,t_ons (le_nl£h ---- (cross
Imct,ton)---- (in.)_I.O_crld] veill_r.. --..-
B-2
BWR SPENT FUEL DISCHARGE STATISTICS
"T"_t.._ .S:2. Histoncallydischargedas_mblicsforthe GE BWR/4-6 Aue.mbly_ brown downby AssemblyTypeandDischargeYear (reproducedfrom the LWR QuantitiesDatabase).
i i i
LWR QUANTITIT_ DATABASE
Historical Data through December 31, 1990DataBrokenDown By:.Assembly _, Dis,-J_,geYearDischargedAssembliesby Aue.mbly Class:OE BWI_4-6
AVEI_GE TOTAL AVERAGEDISCHARGE NUMBER OF DEFECTIVE BURNUP WEIGHT INITIAL
YEAR ASSEMBLIES ASSEMBLIES" (Mwd/MTIHM) (M'II}IM) ENRICH.,|, li ' i ,
Assembly Type: GE BWR/4.6 7 X 7 GE-21973 50 50 3741 9._ 2.5001974 328 328 9199 63.3 2.5001976 285 5 g718 55.8 1.0981977 304 1 9848 59.5 1.1071978 137 1 9620 26.8 1.0981979 38 10747 7.4 1.097
Assembly Type Totals 1142 585 9373 222.4 1.559
' Assembly Type: GE BWR/4-6 7 X 7 GE-3a1975 2 2 4653 0.4 2.1201976 104 19 7947 19.5 1.3891977 397 21 14032 74.4 1.9931978 575 24 20874 107.8 2.4101979 462 10 18787 86.6 23121980 1233 _ 22 23312 231.2 2.4341981 490 8 24304 91.9 2.450
1982 337 4 23143 63.1 2.2941983 12 22942 2.2 2,334
1984 64 24425 12.0 23341985 76 21396 14.2 2334
Assembly Type Totals 3752 110 21057 703.3 2325
AssemblyType: GE BWR/4.6 7 X 7 GE.3b1976 191 3 10570 36.4 2.0561977 1 19646 0.2 2.5071978 312 3 20538 59.1 2.1501979 435 14 254545 82.6 2.4141980 112 26645 21.3 2.5071981 g7 25146 18.4 2.5041982 36 26674 6.8 2306
AssemblyTypeTotals 1184 20 21943 224.9
AssemblyType: GE BWR./4.68 X 8 ANF1989 220 26708 38.8 2.7411990 512 29675 90.4 2.863
At_cmbly TypeTotals 732 28784 129.1 Z826
• As reported _ theutiliucs
B-3
T,,blea..ti iiii i __ ' i i
LWR OUANTITgES DATABASE
Historical Data throup December 31, 1990Data Broken Down By:. Assembly "I_e, Ditclutrse YearDischarged Assemblies by Assembly _ GE BWR/4.6
AVERAGE TOTAL AVERAGEDISCHARGE NUMBER OF DEFECTIVE BURNUP WEIGHT INITIAL
YEAR ASSEMBLIES ASSEMBLIES" (MWd/MITHM) (MTIHM) ENRICH.l I ' l Jl i i J i i ii
Assembly Type: GE BWR/4.,68 X 8 OE-4,a1977 112 189'24 20.6 2.19019'78 9'2 8 18801 16.9 2.2381979 158 56 20073 _.0 2.3921980 267 73 _1 49.1 2.6701981 392 19 24825 72.2 2.6501982 271 8 26187 49.9 2.7351983 300 2 27497 55.3 2.7361984 333 10 28652 61.2 2.7631985 19 24842 3.5 2.730
Assembly Type Totals 1944 176 24993 357.6 2.631
Assembly Type: GE BWR/4.6 8 X 8 GE-4b1978 3 2 13892 0.6 2.1921979 137 4 17153 25.6 2.1771980 621 19261 116.0 Z1641981 479 22602 89.5 2.4221982 262 2 22328 48.9 2.1941983 91 29507 17.0 2.6851984 146 l 25220 27.3 2.7391985 48 16416 9.0 2.114
Assembly Type Tomb 1787 9 21368 333.7 2.311
Assembly Type: GE BWR/4.6 8 X 8 GE-51980 78 2 2848 14.3 0.7621981 33 12 20093 6.0 2.6551982 220 44 22088 40.0 2.4871983 950 18 27626 173.9 2.7061984 772 27 27108 141.3 2.6781985 630 1 19046 115.5 2.0351986 48 27421 8.8 2.2631987 264 12636 48.3 1.5461988 644 23423 117.8 2.2561989 343 27862 62.9 2.5481990 216 23405 39.5 2.123
Type Totals 4198 104 23645 768.3 2.362
reported _ the utilities
B-4
Table _1. (continued)II I ii i iBij I III I II [] II
LWR QUANTITIES DATABASEH_toncal Data thmu_ December 31, 1990
Data Broken Down By:. Auernbly 1313e, Dicflarge YearDischarged Asu:mbliesby Assembly_ GE BWR/4.6
AVERAGE TOTAL AVERAGE
DISCHARGE NUMBER OF DEFECTIVE BURNUP WEIGHT INITIALYEAR ASSEMBLIES ASSEMBLIES* (MWd/MTIHM) (MTIHM) ENRICH.
....... i i ! I I I II
Assembly Type: GE BWR/4.6 8 X 8 GE-61981 26 2 23200 4.8 2.8831982 1 9565 0.2 Z8611983 198 4 7,6833 36.2 2.8061984 556 39 18516 101.7 2.8181985 1437 96 23946 262.7 2.6261986 1231 14792 225.9 1.7681987 2583 18"/08 473.0 2.362
1988 1263 _ 3 24977 231.3 2.462
1989 1329 ._j/ 20432 243.92.O25
1990 929 20432 170.9 2.037
Assembly Type Totals 9553 144 20394 1750.6 2.297
sembly Type: GE BWR/4.6 8 X 8 GE-71984 1 1 5508 0.2 2.6571986 8 16829 1.5 2.6591987 225 8910 41.7 1.2871988 322 1 15369 59.4 1.9361989 1417 17 18086 261.7 2.2271990 1329 5 24649 245.2 2.474
AssemblyType Totals 3302 24 19826 609.6 2.235
AssemblyType: GE BWR/4-6 9 X 9 ANF Prepress.1989 1 24000 0.2 3.319
,Assembly Type Totals 1 2,4000 0.2 3.319
TOTALS 27595. 972 21094 5099.6 Z309
reported by. the utilities
B-5
Table :_. Projectedquantitiesof spentfuel from GE BWRI4-6 A,u=mblyClass,brokenclownbydischargeyear andbumup bin (Reproduced from the LWR QuantitiesDatabase)
III IN NI ummlll I I I I IN I ii
LWR QUANTITIES DATABASE
Projected Data: No New Orders Casewith ExtendedBumupDam BrokenDown By:. Di_luu'ge Year and Bumup BinProjectedAssembliesfor AssemblyCam: GE BWR/4.6
NUMBER AVERAGE TOTAL AVERAGEDISCHARGE BURNUP --Oi_ BURNUP _r.IGHT INITIAL
YEAR BXN ASSEMEiLmS (MW_) (M'rmM) ENRXCHMENTm INI u I I n I n I n I I Ul I
1991 0- 5000 4 3000 0.7 0.7111991 , 15001-200(X) 132 17310 23.5 1.6501991 20001-25000 572 23007 102.1 Z3001991 25001-30000 I 100 28359 200.8 2.6951991 30001-35000 879 33251 157.2 3.0821991 35001-40000 4 36000 0.7 2.990
1992 O-5000 52 4077 9.2 0.9401992 10001.15000 96 13327 17.0 1.4161992 15001-20000 224 18438 39.5 1,7281992 20001-25000 $84 22416 103.7 2.2081992 25001-30000 610 28392 111,2 2.5991992 30001-35000 1421 32791 252.5 2.9921992 35001-4{X_ 113 36381 20.1 3.201
1993 20001-25000 176 24122 31.8 1.8681993 25001-30000 788 28792 143.4 2.6721993 30001.35000 1597 32959 286.7 3.0601993 35001-40000 96 36000 16.6 3.310
1994 15001-20000 180 17600 31.9 1.9501994 25001-30(X)0 768 28610 138.8 2.8061994 30001-35000 1449 32816 258.5 3.0641994 35001-40000 575 37059 102.5 3.167
1995 25001-30000 264 29143 48.0 2.7291995 34X)01-35000 200I 33192 360.5 3.050
1995 35001-4(XX)0 375 36128 67.2 3_,324
1996 25001-30000 656 29140 118,9 2.8341996 30001-35000 1114 32882 198.6 3.0441996 35001-40000 1268 36340 224.8 3.2131996 40001-45000 4 42000 0.7 3.140
1997 15001-20000 36 16000 6.4 3.2001997 25001.300(X) 79 28206 14.0 2.8451997 3(X)01-35000 1717 33525 309,7 3.0911997 35001-40(X)0 1322 37011 233.9 3.267
1998 25001.34XXX) 208 27629 37,7 3.0331998 30001-35000 676 33538 121.7 3.0411998 35001-4(XXX) 1666 36476 301.7 33001998 40001.45000 144 41004 25.6 3.620
B-6
Tmc _ (mntin_)II gl ii I ii I ii I II I I I I III II III II I
LWR QUANTrrIF_ DATABASE
Projected Data: No New Orders Case wills Extended BumupData Broken Down By:. Disct_'ge Yea' and Bumup BinProjected As_mbliesfor Assembly_ GE BWR/4.6
NUMBER AVERAGE TOTAL AVERAGEDISCHARGE BURNUP OF BURNUP WEIGHT INFIIAL
YEAR BIN ASSEMBLIES (MWd/MTIHM) (MTIHM) ENRICHMENTII I Ill I I I II I I I I
1999 15001-213000 179 19155 32.4 2.5101999 25001-34X}00 55 28562 9.7 3.0111999 30001-35000 1312 33702 235.1 3.1131999 35001-40000 1990 36638 354.4 3.209
2000 25001.30000 219 28024 39.4 3.0622000 30001-35000 478 33399 86.5 3.2282000 35001-40000 1835 36977 328.8 3.272
2001 25001.30(D0 198 29950 35.9 2.6112001 30001-35000 385 33067 70.2 3.1172/301 35001-40000 1616 37293 289.7 3.2822D01 40001-45000 136 42066 24.3 3.687
2002 25001.300(X) 147 27931 26.7 3.0832002 30001-35000 238 32,440 43.2 3.2642002 35001-40000 2532 37691 455.3 3.3462002 40001-45000 557 40607 99.0 3.361
2003 25001.30000 92 28539 16.5 3.1012003 30001.35000 91 33589 16.3 3.3722003 35001-40(_ 1667 38031 298.1 3.38521303 40001-45000 514 41087 91.6 3.3562003 45001-50000 129 45039 22.9 3.874
2004 25001-30000 149 28315 26.9 3.0772004 30001-35000 264 33177 48.2 3.1852004 35001-40000 1721 38103 309.4 3.3022004 40001-45000 223 40789 40.0 3.575
2005 25001-30000 73 28720 13.1 3.0992005 30001-35000 310 33272 56.2 3.23821305 35001_ 1993 37709 357.5 3.3282005 40001-45000 1221 41297 215.2 3.509
2006 25001-3(X_ 145 28271 26.3 3.1042006 30001-35000 110 34122 20.0 3.54421306 3500 I-4(X)(K) 1439 38228 262.3 3.3332006 40001-45000 283 42626 51.5 3.678
2007 25001-300(0) 73 28938 13.2 3.1142007 30001-35000 185 32627 33.3 3.2642007 3500 I.-40000 1618 37384 289.3 3.2892007 40001-45000 1058 40925 186.5 3.510
B-7
Table'S2,. (cominu_)
L_t QU_ DATABASE
ProjectedDam: No New OrdersCasewith E,s'tend_BumupData Broken Down By: DisclmrlteYear zud Bumup BlaProjected A.uanblies for Assembly Class: GE BWR/4.6
NUMBER AVERAGE TOTAL AVERAGEDISCHARGE BURNUP OF BURNUP WEIGHT INITIAL
YEAR BIN ASSEMBLIES (MWd/MI]HM) (MTIHM) ENRICHMENT
2008 25001.30000 147 28547 26.7 3.0922008 30001.35000 194 33642 33.3 3.2852008 35001.40000 2097 37967 378.9 3.3532908 4000145000 913 40985 1617 3,5732008 45001-50000 128 45,,165 22.7 3.901
2009 23001-30000 73 _ 13.1 3.1222009 30001-35000 II I _ 20.0 3.1932009 35001..40000 1303 311147 233.4 3.2972009 40001-45000 29 41310 5.5 3.359
2010 23001.30000 141 29221 25.5 3.1642010 30001-35000 202 32951 36.4 3.2812010 35001.40000 2020 38101 365.5 3.3522010 40001-45000 1427 41145 232.7 3.5122010 45001.50000 127 45800 22.6 3.922
2011 23001-30000 142 29416 23.7 3.2052011 30001-35000 44 34214 &0 3.20320 11 35001-40000 1081 38256 195.2 3.39920 11 40001-45000 315 42117 56.7 3.404
2012 15001-20000 132 16149 23.5 3.1872fi12 30001-35000 272 30958 49.0 3.2802012 3500144X_ 497 38619 89.7 3.4492012 40001-45000 1810 41647 324,,5 3.561
2013 10001-15000 228 13349 41.4 19712013 20001-25000 80 22262 14.5 3.0482013 23001-30000 148 26255 26.9 3.0822013 30001-35000 232 31593 411 3.1702013 35001-40000 759 37903 138.0 3.4032013 44)001-45000 749 42575 134.4 3.4942013 45001-50000 129 45128 22.8 3.596
2014 5001-104_ 28 9457 5.1 3.2172014 10001-15000 716 13694 128.7 3.2852014 15001-204_ 472 16632 84,.0 3.3632014 20001-25000 144 23879 23.6 3.3192014 25001-304_ 840 28250 150.8 3.4452014 30001-35080 324 32708 57.8 3.4272014 35001-404X_ 881 38378 157.3 3.4862014 40001-450430 985 42159 176.9 3.6262014 45001-50000 2/30 47572 35.2 3.903
B-8
Tree_2.. (cm_ucd)i i,i ii i ..... i i, i i iii i i i,
LWR OUANTITIES DATABASE
Projected Data: No New Orders Care with Extended BumupData Broken Down By:. DltclutrSeYear and Bumup BinProjected_blk_ forAssemblyClass:GE BWR/4.6
NUMBER AVERAGE TOTAL AVERAGEDISCHARGE BURNUP OF BURHUP WIFJGHT INITIAL
YEAR BIN ASSEMBLIES (MWd/MTIHM) (M'IIHM) ENRICHMENTii ii i i ii , ii , i Ill ,I • ,
2015 10001-15000 188 14900 35.2 3.5292015 15001-20000 176 16850 31.2 3.94,220 15 25001-30000 256 29811 47.2 3.7832015 30001-35000 151 31837 26.8 3.9012015 35001.,40000 111 37691 20.0 3.4542015 40001.45000 1691 42880 302.4 3.74620 15 45001-50080 20 45543 3.6 3.675
2016 5001-10000 16 9635 2.9 3.0852016 10001-15000 400 14660 73.6 3.2372016 15001-20000 16 19064 2.9 3.1912016 25001-30000 416 28217 76.6 3.3852016 30001-35000 67 33015 12.2 3.-9912016 3500144X_ 54,4 38708 98.8 3.4302016 40001.45000 1357 43210 243.8 3.67720 16 45001-50000 190 45180 33.5 3.915
2017 30001-35000 46 34166 8.1 3.3642017 35001-40(X_ 337 38795 61.3 3.4112017 443001.45000 918 42736 164.4 3.590
2018 15001-20000 196 15558 36.4 3,4222018 30001-35000 196 31284 36.4 3.6052018 35001..400(30 213 38821 38.1 3.5392018 443001.45000 1064 43255 192.4 3.62620 18 45001-50000 536 45218 94.2 3.843
2019 30001-35000 46 33844 8.2 3.3432019 35001-40000 233 38041 42.7 3.34920 19 40001.45000 1053 42820 186.2 3.663
2020 30001-35000 47 33104 8.4 3.2982020 35001,.40000 314 37812 56.5 3.3782020 443001.45000 1434 42271 256.6 3.5942020 45081-50000 203 45164 36.1 3.642
2021 30001-35000 3 31524 0.6 2.0452021 35001-g0(_ 304 38274 53.6 3.3432021 443001.45000 818 42233 144.6 3.6032021 45001-50000 145 45017 25.7 3.632
2022 10(30I-15000 471 13200 83.7 3.274
2022 20001-25000 116 21183 20.0 3.5552022 25001-30000 354 2804 1 63.5 3.3332022 30001-35000 164 31913 28.6 3.52.52022 35001-4(X)_ 444 37941 79.7 3.3682022 40001-45000 1496 42142 267.2 3.591
B-9
Tae4eE;2. (eontinued)i I • ii II I I II I IIII II I I I II I I III II I I I I I III
LWR OUAN'ITr'IF_ DATABASE
ProjectedData: No New Orden Casewith Era.endedBumupData Broken Down By:. D_lutrge Year and Bumup BinProjectedAssem0Uesfor Assembly Class: GE BWPd4.6
NUMBER AVERAGE TOTAL AVERAGEDISCHARGE BURNUP OF BURNUP WEIGHT INITIAL
YEAR BIN ASSEMBLIES (MWd/MTIHM) (MTIHM) ENRICHMENTI I l -- l J l l l l l l Ill Ill Ill E l
2023 X}001-35000 82 32669 14.6 3.2462023 35001-40000 821 38459 148.8 3.4682023 40001-45000 677 43486 120.9 3.742
2024 10001-15000 423 12745 76.0 2.9452024 15001.20000 656 16928 114.8 3.2632024 20001.25000 144 24913 25.4 2.8312024 25001.30000 486 27659 S6.3 3.3782024 30001.35000 7 32928 1.3 2.8572024 35001.40000 564 38382 98.5 3.2552024 40001-45000 1111 41455 196.5 3.506
2025 10001.15000 452 13275 81.9 3.3302025 15001.20000 292 17125 53.1 3.2262025 20001.25000 124 22132 22.0 3.2862025 25001.30000 553 27089 100.0 3.3252025 30001-35000 511 33895 92.5 3.2462025 35001.40000 10,,6 37165 189.2 3.3242025 40001-45f)00 136 41344 24.8 3.427
2026 10001.15000 368 13219 66.5 3.4302026 15001.20000 88 15654 16.3 3.3862026 20001.25000 252 23731 45.0 3.451:2026 25001.30000 204 27475 37.9 3.386202.6 30001.35000 263 32853 47.2 3.3932026 35001-40000 450 37803 81.8 3.4172026 40001-45000 305 41464, 53.0 3.375
2027 10001-15000 172 11160 31.8 3.6492027 15001-20000 276 17827 44.8 3.8022027 20001-25000 172 23296 31.8 3.7372027 30001.35000 287 33351 48.8 3.8982027 35001.-40000 229 36991 42.3 3.8082027 40001.45000 323 42790 55.3 3.7592027 45001-50000 40 48560 7.3 3.940
2028 30001-35000 I 19 34508 21.0 3.2852028 35001-40000 164 39737 29.1 3.299
2029 I0001-15000 332 14537 58.8 3.2102029 20001-25000 216 24280 38.3 3.2102029 30001-35000 216 32858 38.3 3.210
Grand Total 97616 _ 17503.7 3304i i
. B-10
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APPENDIXC
T2P2z
A Computer Program for Estimating
Tritium Target Performance and
Tritium Environmental Source Terms
by
Peter C. Owczarski
John R. Honekamp
January 1994
Science Applications International Corporation, Inc.Richland, WA 99352
for
Plutonium Disposition Study
under
General Electric Contract 190-PUC-9027
TABLE OF CONTENTS
1.0 INTRODUCTION AND SUMMARY ..................... I
2.0 TECHNICAL BASES .........................- 22.1 Thermal Analysis ...................... 22.2 Pressure/Stress Analysis . _ . ............... 32.3 Normal Target Permeation Leakage .............. 52.4 Tritium Leakage from Failed Target Rods ........... 5
2.5 Distribution in the Reactor System ............. 8
3.0 INPUT REQUIREMENTS ................... 9
3.1 Input File T2P2-1.DAT _ ....... •..... 93.2 Input File T2P2-2.DAT .............. 10
4.0 OUTPUT DESCRIPTION ....................... 11
5.0 REFERENCES ............................ 17
6.0 T2P2 PROGRAM LIST ......................... 18
- iii
2 • 0 I'NTRODUCTION AND SUMMARY
The purpose of this document is to descr£be both the operation and
technical bases of the T2P2 code. The code itself is a PC operated composite
collection of four codes that were written independently. The four codes were
all part of the technical procedure for estimating the performance of the
tritium target in the ABWR and estimating the tritium environmental source
terms. With the four pieces together along with additional supporting
subroutines the user can now easily optimize the target rod design while being
aware of the environmental consequences of the design or irradiation history.
In T2P2 the average target rod or pin represents the whole core load of
pins. Thus the environmental impact is from the whole core. However, peak or
low irradiation pins can be studied independently of the other pins. In the
code the chosen design and irradiation parameters are used to calculate the
pin temperature profile, internal pressure and stresses, and normal permeation
leakage. The contribution of normal leakage and failed pin leakage is modeled
in a distribution of tritium in the coolant, pool, and in environmental
streams and reactor and turbine building atmospheres.
The model for leakage of failed pins is new. In this new model the
beginning-of-life failure mode was replaced by a more realistic and
mechanistic failure mode during irradiation. This new model has reduced the
estimate of one-cycle tritium operation environmental source terms by an order
of magnitude.
This document begins with a brief description of the technical bases for
the code in Section 2. Section 3 describes the input requirements for the PC
based code. Section 4 describes the seven output files of the code. Sections
5 and 6 are the references and source code listing, respectively.
To use T2P2 for target optimization, the user should refer to
constraints on the target design discussed by Weber (1992) and Lanning et al.
(1992). These constraints along with those necessary to the ABWR and
production goals should compose a fairly complete basis for initial target
design. Eventually any final pin design would require a T2P2 upgrade that can
examine pin behavior along its axis as well as its position in the reactor.
2.0 TECHNICAL BASES
This section provides brief discussions of the technical bases for each
of the major models in T2P2. The user is encouraged to refer to the source
code listing in this document to examine the calculational methods used and
the expressions used to represent the technical bases.
T2P2 is a collectlon of codes and subroutines. Four codes were written
and used independently, but they depended on each other via input/output
files. Combining them dramatically improved the efficiency of using them and
made parametric studies much easier. The original codes consisted of a
thermal analysis of the target pin at power, an analysis of the reactor
coolant/refueling pool tritium inventory, calculations of pin pressure/stress,
and computation of normal permeation leakage of tritium from the pin. The
first two were FORTRAN codes, the third was a spreadsheet program, and the
fourth was written in MATHEMATICA. The combined program is now in FORTRAN and
includes supporting subroutines. Five crucial parts to T2P2 are explained
more fully below.
T2P2 was programmed to apply to the average target pin performance.
Although the peak irradiation rod performance can be examined on an individual
basis, the overall response of the tritium in the coolant and the
environmental source terms should be interpreted only for the average pin
which represents the contribution of all the pins in the core.
Target pins are assumed to produce tritium at a constant rate during
irradiation. The leakage of tritium to the coolant from the pins also
proceeds at a constant rate. However, the concentration of the tritium in the
coolant responds to loss rates proportional to the tritium concentration with
constant proportionality coefficients (e.g., radioactive decay) during power
and with other constant coefficients during refueling. The resulting coolant
tritium concentration exhibits nonlinear transient behavior.
2.1 Thermal Analysis
The temperature profile in the tritium target is calculated by solving
the steady state equation for conduction in a solid with heat generation for a
composite cylinder of concentric cylindrical layers with heat transport out in
the radial direction only. The added heat transport across the gas gaps due
to radiation and free convection are ignored, so the simulation produces
slightly higher temperatures than reality. The heat generation per unit
volume is calculated for each layer (see Figure 2.0) from the gamma energy
C-2
absorbed and the n-u energy (LiAIO 2 only). The user must supply the average
coolant temperature. Conductivities for the various layers were taken from
Wilson (1991).
The temperature profile is used in the calculations of the gas pressure
in the pin during irradiation, the permeability of the pin to tritium gas, and
the yield stress of the cladding.
2.2 Pressure/Stress Analysis
The technical bases for subroutine GETARG are described as follows.
This subroutine first computes the free gas space in the pin from the pin
dimensions. The gas is assumed to be helium from the initial fill at one
atmosphere and 32 °F and helium from the n-a reaction in the pellet column
which does not retain any helium. The ideal gas law with free gas space
temperature, volume and free gas moles gives the interior pressure.
The interior pressure is calculated both at operating temperature andat
IO0°F (refueling condition). The exterior pin pressure is either the water
vapor pressure at the coolant temperature (at power) or 30 feet of water head
(at refueling). The interior-exterior pressure differences are used to
C-3
Center Gas Space
Zircaloy Liner
LiAIO 2 Target
Gap
Nickel Plated Zircaloy Getter
Gap
Aluminide Barrier
Stainless Steel Cladding
Figure 2.0. Reference Target Pin Radial Cross Section.(to scale with cladding o.d. = 0.483 in.)
calculate the pin cladding hoop and axial stresses at power and refueling and
compared with the 90% yield stress of the stainless steel cladding.
Other useful calculations performed by GETARG are the GVR and the total
hydrogen/getter zirconium ratio (TH/Zr). The GVR is the ratio of STP gas
volume produced to the original pellet volume. The TH/Zr is 1.1 times the
atoms of tritlum/atoms getter zirconium. The 1.1 factor (based on In-reactor
test results} reflects the likelihood that some protium from the primary
coolant diffuses into the pin.
GETARG also supplies the heat generation rates for the thermal analysis.
2.3 Normal Target Permeation Leakage
No materials are now known that can be placed in the target
configuration and under reactor conditions can totally prevent any tritium
from escaping the pin. Since all the materials of construction have some non-
zero permeability, it is necessary to be able to estimate the permeation
losses from the pin in terms of the internal tritium partial pressure and the
barrier permeabilities. The portion of the main T2P2 code called TLEAK
estimates this permeation loss as a function of irradiation time. Sherwood
(1992) gives a detailed description of the models within TLEAK; the
descriptions are not repeated here. The models are empirical representations
of physical data that include the solubility of tritium gas in the pellet and
getter and the permeation rate through the cladding. The permeation leakage
depends on the temperature of the pin components as well as the pin
dimensions. Any studies made with T2P2 would be sensitive to changes in the
permeation model parameters.
2.4 Tritium Leakage from Failed Target Rods
None of the getter-barrier target rods irradiated during the DOE Tritium
Target Development Program developed cladding leaks. Thus, there are no
directly applicable experimental data on which to base clad failure
predictions or tritium leakage from a failed target rod. However, there are
extensive data for Zircaloy clad LWR fuel elements and Stainless Steel clad
fast reactor fuel elements.
The prior model used to estimate tritium target rod failures (Apley
1992) was based on the premise that the target rod cladding was less likely to
fail than the LWR fuel rod cladding which operated under more stringent
conditions. Post-irradiation examination of the getter-barrier target rod
C _ 5
(Lanning 1992) has confirmed that its cladding did not experience any chemical
or mechanical interaction with its internal components and could be modeled as
a free-standing gas-pressurized tube. On the other hand, LWR fuel element
cladding is subject to internal corrosion, hydrldlng and fuel-clad mechanical
interaction in addition to gas pressure. Further, the 20% cold-worked 316-
stainless steel used for the target rod is a much stronger material than
zircaloy at equivalent temperatures.
The Apley model used a clad failure frequency of 10_ased on LWR fuel
element experience and assumed that 50% of the tritium inventory in the failed
rod would be released. The 50% release implies that all cladding failures
exist from beginning of life and that the capacity for retaining tritium is
determined by the solubility of tritium in the LiA102. This assumption was
considered to be very conservative, since most failures would likely occur
late in life when essentially all the tritium is tied up in the zirconium
getter and pellet. Further, throughout the target rod life very little
tritium is present in gas phase. However, because the tritium permeation from
intact target rods is so low, the release from failed target rods using the
Apley model dominates the tritium source term.
To overcome this limitation a more mechanistic and realistic target rod
failure model was developed for T2P2. First we assume a cladding failure mode
similar to stainless steel clad LMR fuel elements. These failures are
typically microscopic pin-holes that can take many hours to depressurize. We
also assume that both the LiAIO2 pellets and the zircaloy getter tubes will
retain their tritium inventory at the time of clad failure. This assumption
is based on the knowledge that both the pellets and getters are stable in
water and steam at the ABWR operating conditions.
The model assumes that there ks no steam or water ingress into a target
rod with a pin-hole cladding failure until a shutdown - restart cycle occurs.
This is based on the premise that clad failure is very unlikely until the
internal helium pressure exceeds the primary system pressure. Further, as
long as the internal helium pressure is equal to or greater than the primary
system pressure, and the hole in the cladding remains small, the getter would
continue to perform its function. However, during the first shutdown -
restart cycle following clad failure it is assumed that the target rod
depressurizes and becomes water/steam logged. After the shutdown - restart
cycle we assume that the getter is oxidized, ceases to function as an internal
tritium sink, and all tritium released from the LiAIO2 escapes to the coolant.
The amount of tritium leaving the pellet is governed by tritium
solubility. This solubility is designated GVRo (Sherwood 1992}. GVRo
decreases as temperature increases and ks about 40-60. For GVR < GVRo, only
3% of the tritium formed escapes the pellet due to recoil. For GVR > GVRo
100% of tritium formed escapes the pellet. Thus we have defined that the
availability of tritium to leave the pan depends on the degree of irradiation
and the ingress of water into the pin after a shutdown cycle followlng the
tim8 of failure.
In Subroutine FAIL we have assigned a frequency of shutdown cycles to
two (the user can recompile T2P2 with a different number} with the usual
refueling periods. That leaves three periods per refueling cycle in which the
pins can fail. Each period has a different consequence of tritium leakage.
We now explain the likelihood of failure.
Target rods failures are assumed to follow a Weibull distribution
similar to stainless steel clad LMR fuel elements with a failure frequency of
10 4 over a three year irradiation period. The Weibull distribution function,
and the details for selecting the three parameters involved, are documented in
the references listed in the source code for Subroutine FAIL. The model
quantifies the failure assumptions and produces a tritium source term to the
coolant for the exposure cycle between target pin loadings. The source term
rate is assumed to be linear over the loading cycle and is added directly to
the normal permeation leakage term.
The overall effect of the new model is to delay the failure to an
exposure determined by the Weibull distribution function and the frequency of
shutdown-restart cycles. This reduces the impact of failure because only
tritium released from the LiAIO 2 after clad failure escapes to the coolant.
For the parameters selected, this model reduced the consequences of failure by
an order of magnitude compared to the Apley model. Further, with this model
the tritium release is readily adjusted to accommodate actual operating
experience with respect to shutdown-restart cycles and clad failure
statistics.
C A 7
2.5 Distribution in the Reactor Systea
The distribution of tritium in the reactor coolant system follows a
simple first order differential equation:
Dc/dt = S - (L l + L2)C
where C = bulk coolant curies of tritium
t = time
S = source leakage rate from pins
Li = fractional rate of loss of coolant via steam leaks
L2 = fractional rate of loss of coolant via pool evaporation.
When pool liquid is isolated from the rest of the coolant during power,
a separate differential equation with only a L2 term is solved for that
liquid.
The Ls and S are held constant during power and with Ll=S=O during
refueling. The Subroutine H3CYCLE listing gives the details of the solution
to the differential equations.
The model for pool evaporation contains an evaporation mass transfer
model and assumes instantaneous mixing of the evaporation source with reactor
building air giving a bulk air concentration of tritium in the building. This
simple model has room for upgrading in two areas. The first might be to
establish local or 3-d concentration profiles, if possible, for detailed
worker exposure. The second is to provide a mechanistic mass transfer
coefficient based on local air flow velocities, rather than use a recommended
but nonmechanistic default value.
C^ 8
3.0 INPUT REQUI_4m_TS
There are two separate input files needed to run T2P2: T2P2-1.DAT and
T2P2-2.DAT. The latter file input varlables control the disposition of
tritium throughout the reactor system and control the environmental source
term. The former file variables control the target pin behavior in the
reactor. The simplest way to describe the file variables Is to use an example
run throughout this document. A description of each number used in the
example is found below. An important distinction in input values in the two
DAT files is that the duration of the refueling cycle does not have to be the
same as the tritium target irradiation cycle.
3.1 Input File T2P2-I.DAT
The example _ile consists of the following five rows of numbers:
9,538.
.5893,.6096,.9906,1.0084,1.049,1.0592,1.0744,1.226816.3,3.18,.241,16.3,.241,100.,18.22.866E+13,381.,25.4,2.15,.5,273.75,365.
10,0.,1.,30.,60.,90.,120.,150.,180.,210.,240.,273.75
9 = number of concentric regions in the pin, not counting thecenter hole.
538. = average coolant temperature, degrees F.
.5893 = inner diameter of liner, cm.
.6096 = outer diameter of liner = inner diameter of pellet, cm. etc.to 1.2268 = outer cladding diameter.
16.3---18.2 = thermal conductivity of the 7 regions, watts/m/_.
2.866E+13 = n-alpha reaction rate, rxns/s/cm pellet.
381. = pellet column length, cm.
25.4 = pin plenum length, cm.
2.15 = gamma heating rate, watts/g.
.5 = lithium enrichment, fraction of Li-6 at beginning of life.
273.75 = duration of tritium irradiaton/cycle, days.
365. = duration of cycle, including refuelling, days.
i0 = number of time steps desired for normal pin leakagecalculations.
0.---273.75 = I0 time points, in days, for leakage calculations.
The input numbers have to be placed in the five rows. If this cannot be done,
then the code READ statements will have to be changed for the too long rows.
3.2 Input File T2P2-2.DAT
The example file consists of the following eight rowsz
9,273.75,91.25,273.75,91.254.86e+6,10.52e+63488.
83000.,35000.,201000.1,1,1
9 ffi number of consecutive cycles.
273.75 = number of days at full power in one refueling cycle.
91.25 = number of days off power in one cycle.
273.75 = differential days at power for calculational points =2_3.75/I
91.25 = differential days off power for calculational points =91.25/1
4.86E+6 = reactor coolant inventory at powers ibm.
10.52E+6 = additional refueling pool inventory, ibm.
3488. = number of target rods in core.
1 = number of leak paths from coolant during power.
1 = number of leak paths from pool during power.
1 = number of leak paths from pool and coolant off power.
C- I0
4.0 OUTPUT DESCRIPTION
The output from T2P2 consists of seven files. These files provide the
user with tritium target rod performance, reactor tritium inventories, and
environmental source terms. Each of the file numbers and titles are listed
below.
List of output files:
T2P2-1.OUT TARGET THERMAL ANALYSIS OUTPUTT2P2-2.OUT NORMAL TARGET LEAKAGE ANALYSIS OUTPUT
T2P2-3.OUT GVR/PRESSURE/STRESS CALCULATIONST2P2-4.OUT REFUELING CYCLE COOLANT INVENTORYT2P2-5.OUT LEAKAGE PARAMETERS & OUTPUTT2P2-6.OUT TRITIUM Ci BALANCE CHECKS
T2P2-7.OUT TRITIUM DISPOSITION OUTSIDE REACTOR
Sample output files follow.
T2P2-1.OUT TARGET THERMAL ANALYSIS OUTPUT
Description: This output file provides the steady state temperatureprofile of the target pin. The concentric layers have outer radii,r(out), in meters, thermal conductivities in w/m/K, heat generation
rates in w/m3, outer radii temperatures, and volume average temperaturesin degrees Celcius.
layer r(out) t cond ht sen puv degrees C
1 3.048000E-03 16.300000 1.443254E+07 359.6986002 4.953000E-03 3.180000 5.172376E+07 333.7321003 5.042000E-03 2.410000E-01 0.O00000E+00 304.2784004 5.245000E-03 16.300000 1.443254E+07 303.295400
5 .296000E-03 2.410000E-01 0.000000E+00 286.6868006 5.372000E-03 100.000000 1.702800E+07 286.6278007 6.134000E-03 18.200000 1.702800E+07 283.302700
layer Tavg
1 359.7016002 349.5846003 318.9178004 303.783600
5 294.9643006 286.657400
7 284.937100
TGASR 1126.809000 des R gas temp fcr press calcs
T2P2-2.OUT NORMAL TARGET LEAKAGE ANALYSIS OUTPUT
Definitions:
tday = days irradiation time at end of periodrl = tritium permeation rate, gmoles/secondnt2 = tritium gram moles in getter
totalt2 = total gram moles tritium produced
C° ii
iperm - target permeation rate, gmoles T2/sec
relCi - g-moles T2 released in time periodsumCl - cumulative Tritium Ci released to coolant
p - tritium gas pressure, Pascalstzratlo - ratio of T atoms/Zr atomslkratlo - ratio of leaked to total tritium
totCi - T Ci inventory in target after decay
tday rl nt2 totalt2
1.000000 2.379213E-11 7.831382E-06 7.831382E-0430.000000 5.078669E-10 2.606835E-03 2.344148E-0260.000000 9.073245E-10 9.634805E-03 4.677434E-0290.000000 1.221514E-09 2.011289E-02 6.999906E-02
120.000000 1.468536E-09 3.327537E-02 9.311616E-02
150.000000 1.663008E-09 4.852151E-02 1.161261E-01180.000000 1.815889E-09 6.537999E-02 1.390295E-01210.000000 1.936136E-09 8.348110E-02 1.618267E-01240.000000 2.030716E-09 1.025348E-01 1.845182E-01273.750000 2.113211E-09 1.248276E-01 2.099206E-01
iperm relCi sumCi p
5.076373E-15 2.192993E-10 1.276174E-05 1.805498E-042.953369E-14 4.335949E-08 2.535991E-03 6.111185E-034.254178E-14 9.340981E-08 7.971810E-03 1.268005E-024.985401E-14 1.197449E-07 1.494016E-02 1.741366E-025.508119E-14 1.359960E-07 2.285420E-02 2.125673E-025.894406E-14 1.477767E-07 3.145381E-02 2.434277E-026.185364E-14 1.565538E-07 4.056418E-02 2.680528E-026.407187E-14 1.631994E-07 5.006129E-02 2.876237E-026_577690E-14 1.682840E-07 5.985428E-02 3.031354E-02
6.723746E-14 1.939349E-07 7.113998E-02 3.167469E-02
tzratio Ikratio totCi
8.807521E-06 2.779307E-07 45.9169702.931762E-03 1.845134E-06 1374.4210001.083573E-02 2.906796E-06 2742.4740002.261985E-02 3.640224E-06 4104.1860003.742297E-02 4.186067E-06 5459.5880005.456945E-02 4.619644E-06 6808.7090007.352925E-02 4.976236E-06 8151.5790009.388657E-02 5.276149E-06 9488.225000
1.153152E-01 5.532495E-06 10818.6800001.403867E-01 5.779947E-06 12308.070000
T2P2-3.OUT GVR/PRESSURE/STRESS CALCULATIONS
This output focuses on the end of the target irradiation period (eol)in the reactor. At this point the interior pan pressure is at maximumwhile still at power with the coolant at maximum temperature. Thecoolant ks then cooled to IOOF. Symbol definitions:
VVR = void volume/pellet volume90%YS = 90% yield stressTH/Zr ratio = atoms of H+T/atoms Zr
Vcladid 358.095800 Vpellet 181.525100 cm**3
. C-12
Vgetter 24.870410 Vget Ni 21.758550
Vllner 7.246315 VsprTng 3.070000Vnet 141.383900 VVR 7.788669E-01GVR 79.420030
GVRo 43.311820
Li6 depletion % 12.024320
Reactor Cycle Capacity Factor 75.000000 %
T2 molecules, no decay 1.291335E+23T2 molecules, w/ decay 1.264352E+23T2 curies 12308.070000 TH/Zr ratio 3.025226E-01Fraction of T2 in gas 4.012920E-09
heat generation watts/cm**3:cladding 17.028000getter 14.432540liner 13.760000
pellet 51.723760
internal pressure at power 2323.939000 psiinternal pressure at 100 F 1154.948000 psi
hoop stress at power 9288.527000 psiaxial stress at power 4644.264000 psi
hoop stress at 100 F 7688.197000 psiaxial stress at 100 F 3844.099000 psi
90%YS@pwr 71737.800000 90%YS100F 82296.000000 psi
netDP@pwr 1361.992000 netDPIOOF 1127.333000 psi
Puff T2 Ci released from sudden eol pin failure:2.894677E-05 at power to coolant4.821030E-05 @ IOOF in pool4.873327E-05 @ 75F in air at 1 arm
T2P2-4.OUT REFUELING CYCLE COOLANT INVENTORY
Input Parameters:
Number of cycles 9Days at full power 273.750000Days of refueling 91.250000Lbm coolant 4860000.000000
Lbm refuel pool 1.052000E+07
Days dt at power 273.750000Days dt refueling 91.250000
CYCLE NUMBER ( i)
Days Coolant H3, Ci Pool H3, Ci
273.750000 103.991000 0.000000E+00 at power
365.000000 27.674740 59.904990 refueling
CYCLE NUMBER ( 2)
Days Coolant H3, Ci Pool H3, Ci
_- 13
273.750000 104.568100 38.834530 at power365.000000 38.163210 82.608430 refueling
CYCLE NUMBER ( 3)
Days Coolant H3, Ci Pool H3, Ci273.750000 104.786800 53.552460 at power
365.000000 42.138250 91.212840 refueling
CYCLES 4-8 OMITTED BELOW.
CYCLE NUMBER ( 9)
Days Coolant H3, Ci Pool H3, Ci273.750000 104.919900 62.507940 at power365.000000 44.556960 96.448390 refueling
T2P2-5.OUT LEAKAGE PARAMETERS & OUTPUT
Discussion: This output consists of cumulative and cycle incremental tritiumairborne leaks. Here path one is the turbine building stack and path 2 is the
reactor building stack. Loop leakage at power is from the turbine buildingsteam loss. Pool leakage at power is from evaporation of refueling pool waterduring power. Pl+ip leakage during refueling is also pool evaporation, whichincludes coolant water at that time.
Leakage parameters:
No. of target rods 3488.000000Frac. Fail T2 Loss 3.794628E-06
Ci leaked/cycle 411.041700H3 Ci/day source 1.501522
loop leak rate no. 1 at power 67962.000000 ibm/daypool leak rate no. 1 at power 15038.000000 lbm/dayip+pl leak rate no. 1 refueling 26581.730000 lbm/day
CYCLE NUMBER ( 1) ************************
cycle day= 273.750000 (at power)Total loop leak from path 1 303.708000 CiTotal pool leak from path 1 0.000000E+00 Ci
cycle day= 365.000000 (refueling)Total pl+ip leak from path 1 15.069300 Ci
Cycle 1 Incremental Leak Values:
Cycle loop leak from path 1 303.708000 CiCycle pool leak from path 1 0.O00000E+00 CiCycle pl+ip leak from path 1 15.069300 Ci
CYCLE NUMBER ( 2) ************************
cycle day= 273.750000 (at power)Total loop leak from path i 634.218600 Ci
([-, 14
Total pool leak from path 1 19.022310 Ci
cycle day- 365.000000 (refueling)Total pl+Ip leak from path 1 35.849730 Ci
Cycle 2 Incremental Leak Values:
Cycle loop leak from path 1 330.510600 CiCycle pool leak from path 1 19.022310 CiCycle pl+lp leak from path 1 20.780430 Ci
CYCLE NUMBER ( 3) ************************
cycle day= 273.75C000 (at power)Total loop leak from path 1 974.887100 CiTotal pool leak from path 1 45.253910 Ci
cycle day= 365.000000 (refueling)Total pl+ip leak from path 1 58.794630 Ci
Cycle 3 Incremental Leak Values:
Cycle loop leak from path 1 340.668500 CiCycle pool leak from path 1 26.231600 CiCycle pl+_p leak from path 1 22.944900 Ci
CYCLES 4-8 OMITTED BELOW.
CYCLE NUMBER ( 9) ************************
cycle day= 273.750000 (at power)Total loop leak from path 1 3052.322000 CiTotal pool leak from path 1 226.364600 Ci
cycle day= 365.000000 (refueling)Total pl+ip leak from path 1 203.585900 Ci
Cycle 9 Incremental Leak Values:
Cycle loop leak from path 1 346.849400 CiCycle pool leak from path 1 30.618260 CiCycle pl+ip leak from path 1 24.261920 Ci
T2P2-6.OUT TRITIUM Ci BALANCE CHECKS
cycle number 1source total 411.041700
coolant inventory 27.674740
refuel pool inv y 59.904990Ci leaked 318.777300
Ci decayed 4.684668net difference -2.336502E-05
cycle number 2source total 822.083300
coolant inventory 38.163210
refuel pool inv_y 82.60B430Ci leaked 689.090600
Ci decayed 12.221070net difference -3.433228E-05
C- 15
cycle number 3source total 1233.125000
coolant inventory 42.138250
refuel pool inv_y 91.212840Ci leaked 1078.936000
Ci decayed 20.838260net difference -1.525879E-05
CYCLES 4-9 OMITTED.
T2P2-7.OUT TRITIUM DISPOSITION OUTSIDE REACTOR
The values printed below can be used to estimate offsite doses (from buildinglosses) and occupational doses (from building airborne concentrations). Dosesmust be computed for tritium as part of a water molecule (TOH or T20), not asT2 or HT. If tritium is the only contributor to occupational dose, then alimit of 2E-06 microCi/cc can be tolerated before respiratory protection isrequired.
The following Ci and concentrations are for the two periods in the365.000000 day production cycle number 9 with
273.750000 days at power and 91.250000 days refueling.
Reactor Building:
Total evap. loss during power = 30.618260 CiAir H3 conc during power = 7.837018E-08 microCi/cc
Total evap. loss refueling = 24.261920 CiAir H3 conc refueling = 1.863017E-07 microCi/cc
Turbine Building:
Total steam loss during power = 346.849400 CiAir H3 conc during power = 1.545907E-07 microCi/cc
Other Losses:
No other losses.
C- 16
5.0 REFERENCES
Apley, W.J. 1992. Tritlum Tarqet Development Pro_ect Executive Summar 7
Topical Report. PNL-8142, Pacific Northwest Laboratory, Richland, WA.
Lanning, D.D., D.L. Baldwin, and R.J. Guenther. 1992. Final Report on theWC-1Liqht-Water Reactor Tarqet Rod Irradiation Test and postirrQdiation
Examinations. PNL-8133 Volume 1: Text, Pacific Northwest Laboratory, Richland,WA.
Sherwood, D.J. 1992. Modelinq the Behavior of a Liqht-Water Production_eactor Tarqet Rod. PNL-8010, Pacific Northwest Laboratory, Richland, WA.
Weber, J.W. 1992. Topical Report: NPLWR Tritium Tarqet Desiqn. WHC-SP-0840,Westinghouse Hanford Company, Richland, WA.
Wilson, D.R. 1991. TKTARI: A Computer Code for Predictinq Tritium Tarqet RodPerformance. WHC-SP-0684, Westinghouse Hanford Company, Richland, WA.
C_17
6.0 T2P2 PROGRAM LIST
C Program T2P2 (Tritium Target Performance Program) combines TATT, TLEAK,C GETARG, H3CYCLE,and EVAP and now employs user inputs for parametricC studies of the controlling variables.
C ************************************************************************
C Program TATT (Thermal Analysis Tritium Target) performs the steady stateC temperature analysis of a tritium target rod.
REAL k(lO), A(10), d(0:ll),C(2,0:ll),Tr(0:11),Trf(O:ll),Trc(0:11),+Trav(10),r(0:11)REAL len, mzr,molzr,mpr,ka,tday(O:lO),tsec(O:lO),rl(lO),nt2(lO),
+totlt2(lO),pus(10),pe(10),p(10),iperm(10),relCi(10),sumCi(10),+ipave,tzr(10),lambdaT,n_A
COMMON/BLK1/Nt,A,d, len,plen,N_A,ntgr,TGASR,gammaen,enrichli,+FPcycle,EOcycle,pT2,TpelletCOMMON/BLKI0/GVR,GVRoCOMMON/BLK2/sumc,totCiCOMMON/BLK5/tfluid,tcladf
Open(Unit=l,File='t2p2-1.dat',status='old')Open(Unit=2,File='t2p2-1.out',status='unknown')
C SI units used in calculations.
C k(j) thermal conductivity of concentric region j.C r(j) outer radius of region j.C C(m,j) integration constants - to be determined.C A(j) heat generation puv in region j.
C Nt = no. of concentric regions less the center gas core.
C Tf = coolant temperature, K.C h = cladding-to-coolant heat transfer coefficient.
Read(l,*)Nt,TfluidFTf= (TfluidF+460.)/l.8h=40000.lambdaT=l.792E-09
Read(l,*)(_i(j),j=0,Nt)do l=0,Nt
r(1)=d(1)/200.enddo
Read(l,*)(k(j),j=l,Nt)
Read(l,*)n A, len,plen,gammaen,enrichli,FPcycle,EOcycle
c n A = n-alpha rxn rate per cm pellet/secondc fen = pellet length, cmc plen = rod plenum length, cm
ntgr=lcall getarg
do m=l,Nt
A(m)=I.E+6*A(m)
C.-18
enddo
c Define c(m,n) matrix:
C(1,1)= A(1)*r(0)**2./2./k(1)DO J=l,Nt-1
C(l,J+l)=r(J}*((A(J+l)-A(J))*r(J)/2.+C(l,J)*k(J}/r(J))/k(J+l)ENDDO
c Cladding surface temperature Trn
Trn=Tf-(C(l,Nt}*k(Nt}/r(Nt}-A(Nt)*r(Nt}/2.)/h
C(2,Nt)=Trn+A(Nt)*r(Nt)**2./4./k(Nt)-C(l,Nt)*ALOG(r(Nt))
DO J=Nt-l,1,-i
C(2,J)=C(2,J+I)+(C(1,J+I)-C(I,J))*ALOG(r(J)}++ (A(J)/k(J)-A(J+l)/k(J+l})*(r(J}**2.)/4.ENDDO
C(2,0)=-A(1)*r(0)**2./4./k(1)+C(I,I)*ALOG(r(O))+C(2,1)Tr(O)=C(2,0)
write(2,*)' '
write(2,*)' 'write(2,*)' TARGET THERMAL ANALYSIS OUTPUT'write(2,*)' 'write(2,*)' '
write(2,*)'Description: This output file provides the steady stat+e temperature"
write(2,*)'profile of the target pin. The concentric layers have+outer radii,'
write(2,*)'r(out), in meters, thermal conductivities in w/m/K, hea+t generation'write(2,*)'rates in w/m3, outer radii temperatures, and volume ave
+rage temperatures'
write(2,*)'in degrees Celcius.'write(2,*)' '
write(2,*)' 'do j=l,Nt
Tr(J)=-A(J)*r(J)**2./4./k(J)+C(I,J)*ALOG(r(J))+C(2,J)tav=-A(J)*(r(J)**4.-r(J-l)**4.)/8./k(J)+(C(2,J)-C(l,J)/2.)*
+ (r(J)**2.-r(J-l)**2.)++ C(l,J)*(alog(r(j))*r(J)**2.-alog(r(J-l))*r(J-l)**2.)
Trav(J)=tav/(r(J)**2.-r(J-l)**2.) - 273.16enddo
Tpellet=Trav(2)+273.16
do m=0,Nt
Trf(m)=Tr(m)*l.8-460.Trc(m)=Tr(m)-273.16
c write(2,*) m,Trf(m),' F',Tr(m),' K',Trc(m),' C'enddo
write(2,*)' layer r(out) t cond ht gen+puv degrees C'write(2,*)' 'do m=l,Nt
write(2,*)m,r(m),k(m),A(m),Trc(m)enddo
write(2,*)' '
C" 19
write(2,*)' layer Tavg'write(2,*)' 'do n=l,Nt
write(2,*} n , Tray(n)enddo
tcladf=Trf(Nt)tfluid=tfluidf
vgsinr=r(0)**2.
vgsgapl=r(3)**2.-r(2)**2.vgsgap2=r(5)**2.-r(4)**2.vgst=vgsinr+vgsgapl+vgsgap2tnr=Tr(0)-273.16
tgsc=(vgsinr*tnr+vgsgapl*Trav(3)+vgsgap2*Trav(5))/vgstTGASR=tgsc*l.8+492.write(2,*)' '
write(2,*)' TGASR ',TGASR,' deg R gas temp for press calcs'ntgr-2
c call getarg
C ************************************************************************
C Program TLEAK computes the permeation loss of a tritium target rodC during irradiation in the manner of EP Simonen using a MathematicaC language code (see 4/9/93 transmittal).
Open(Unit=3,File='t2p2-2.out',status='unknown,)Read(1,*)mt,(tday(j),j=0,mt)
pi=3.14159rgas=8.314len=381.
rpin=r(1)*100.rpout=r(2)*100.rgin=r(3)
rgout=r(4)rci=r(5)rco=r(7)a1=2.*pi*(rgin+rgout)
vl=pi*(rpout**2.-rpin**2.)vg=pi*(rgout**2.-rgin**2.)*len*10000.rzr=6.49
mzr=rzr*vgamuzr=91.22
molzr=mzr/amuzrwclpav=8.87E+13
pavg=n_A/vl
tc=(Trav(6)*(r(6)**2.-r(5)**2.)+Trav(7),(r(7)**2.-r(6)**2.))/+(r(7)**2.-r(5)**2.)tg=Trav(4)tp=Trav(2)
do j=O,mt
tsec(j)=tday(j)*86400.enddo
C" 20
eol=tsec(mt)mpr=pavg*vl*lO0./6.023E+23
taudif=6.65E-lO*exp(l.31/(8.62E-5*(tp+273.)))tausur=66.4*wclpav/pavgreltau-taudif+tausur
qa=8156.,4.184ka-4.1*exp(-qa/(rgas*(tg+273.)))xp=-65700./(rgas*(tc+273.))
pss-1.01E+5**(-.5)*2.33E-2*exp(xp)/224.14qe=36910.,4.184ps=638758.6*exp(-qe/(rgas*(tg+273.)))*760_*133.3
c Time marching sequence 0<t<eol.
sumc=O.
DO j=l,mt
emprl=O.5*mpr*(l.-exp(-tsec(j)/(86400.*reltau)))rl(j)=O.5*0.01*mprif(rl(j).it.emprl)rl(j)=emprl
ts=tsec(j)call INGRATE(reltau,ts,rint)
nt2(j)=0.5*mpr*rint*len/100.
totlt2(j)=mpr*tsec(j)*len/200.
c Correct nt2 and totlt2 for tritium radioactive decay. Correctionc is based on constant production rate.
xdk=ts*lambdaT
cordk=(l.-exp(-xdk))/xdknt2(j)=cordk*nt2(j)totlt2(j)=cordk*totlt2(j)
zp=-9.632E+4/(rgas*(tg+273.))pus(j)=2.888*(2.*nt2(j)*123.2/molzr)**2.*760.*133.3*exp(zp)pe(j)=pus(j)if(pus(j).gt.ps)pe(j)=ps
p(j)=(rl(j)*rgas*(tg+273.)/(al*ka)+pe(j)**.875)**(l./.875)prfave=100.
iperm(j)=2.*pi*pss*(len/lOO.)*p(j)**.5/prfave/alog(rco/rci)
dt=tsec(j)-tsec(j-l)
if(j.eq.l)thenipave=iperm(j)/2.
else
ipave=(iperm(j)+iperm(j-l))/2.endif
relCi(j)=ipave*dtsumc=sumc+2.*6.023E+23*relCi(j)/2.07E+19sumCi(j)=sumctzr(j)=2.*nt2(j)/molzr
ENDDO
C-21
pT2=p(mt)
write(3,*}' '
write(3,*}' 'write(3,*)' NORMAL TARGET LEAKAGE ANALYSIS OUTPUT'
write(3,*}' 'write(3,*}' 'write(3,*}'Definitions:'
write(3,*}' tday - days irradiation time at end of period"write(3,*}' rl - tritium permeation rate, gmoles/second'write(3,*}" nt2 - tritium gram moles in getter'write(3,*}' totalt2 - total gram moles tritium produced'write(3,*}' iperm = target permeation rate, gmoles T2/sec'write(3,*)' relCi = g-moles T2 released in time period'write(3,*}' sumCi = cumulative Tritium Ci released to coolant'
write(3,*}' p = tritium gas pressure, Pascals'write(3,*)' tzratio = ratio of T atoms/Zr atoms'write(3,*)' lkratio = ratio of leaked to total tritium'
write(3,*)' totCi = T Ci inventory in target after decay'write(3,*)' 'write(3,*)' '
write(3,*}' tday rl nt2+totalt2'
write(3,*)' '
do j=l,mt
write(3,*)tday(j),rl(j),nt2(j),totlt2(j)enddo
write(3,*)' 'write(3,*)' iperm relCi sumCi
+ p'write(3,*)' '
do j=l,mt
write(3,*)iperm(j),relCi(j),sumCi(j),p(j)enddo
write(3,*)' '
write(3,*)' tzratio ikratio totCi'write(3,*)' '
c Specific Activity of T2 = 9720 Ci/gram x 6.0321gram/gmolesat2=9720.*6.0321
do j=l,mt
totCi=totlt2(j)*sat2tlkr=sumCi(j)/totlt2(j)/sat2write(3,*)tzr(j),tlkr,totCi
enddo
call getargcall RFcycle
stopend
C ************************************************************************
Subroutine RFCYCLE
C Code RFCYCLE computes losses of H3 from the coolant system of a
C - 22
C reactor for successive refueling cycles.
REAL Lr(lO),Lb(10),Cb(lO),Cr(lO},idk,Mo,Lkb,Lkr,netdif
Real Mr,Cbp(lO),Lbp(10),Lkbp,Ntarg,Edot(2),Lxtrareal dCb(lO),dCbp(10),dCr(lO)real A(lO),d(O:ll),n a
COMMON/BLK1/Nt,A,d, len,plen,N_A,ntgr,TGASR,gammaen,enrichli,+FPcycle,EOcycle,pT2,TpelletCOMMON/BLKIO/GVR,GVRoCOMMON/BLK2/sumc,totCiCOMMON/blk3/Edot,RBacfmCOMMON/BLK4/FailT2fr,rrctrc,tb
c Lr and Lb are system leak rates in lbs/day during refueling andc buildup, respectively.c Cb and Cr are the corresponding cumulative H3 losses in Curies with noc radioactive decay once leaked.
Open(Unit=ll,File='t2p2-2.dat',status='old'}Open(Unit=12,File='t2p2-4.out',status='unknown')Open(Unit=13,File='t2p2-5.out',status='unknown'}Open(Unit=14,File='t2p2-6.out',status='unknown'}Open(Unit=15,File='t2p2-7.out',status='unknown')write(12,*)' 'write(13,*)' 'write(12,*)' ***REFUELING CYCLE COOLANT INVENTORY ***'
write(13,*)' ******LEAKAGE PARAMETERS & OUTPUT******'write(14,*)' **_****TRITIUM Ci BALANCE CHECKS*******'write(13,*)' 'WRITE(12,*)' 'Write(12,*)' Input Parameters:'write(12,*)' 'Read(ll,*)Ncm,tb,tr,dtb,dtrRead(ll,*)Mo,MrRead(11,*)NtargRead(ll,*)Wfeed,RBacfm,TBacfm
c rrctrc=ratio of tritium cycle length/refueling cycle lengthrrctrc=EOcycle/(tb+tr)
call FAIL
tlpc=Ntarg*(FailT2fr*totCi+sumc)Cdot=tlpc/tb/rrctrc
write(12,*)' Number of cycles ',Ncm
write(12,*)' Days at full power',tbwr_ e(12,*)' Days of refueling ',trwr_ue(12,*}' Lbm coolant ',Mowrite(12,*)' Lbm refuel pool ',Mrwrite(12,*)' Days dt at power ',dtbwrite(12,*)' Days dt refueling ',dtr
write(13,*)'Discussion: This output consists of cumulative and cy+cle incremental tritium'
write(13,*)'airborne leaks. Here path one is the turbine building+ stack and path 2 is the'
write(13,*)'reactor building stack. Loop leakage at power is from+ the turbine building'write(13,*)'steam loss. Pool leakage at power is from evaporation
+ of refueling pool water'
write(13,*)'during power. Pl+ip leakage during refueling is also+pool evaporation, which'
C _ 23
write(13,*)'includes coolant water at that time.'write(13,*}' 'write(13,*)' 'write(13,*)' Leakage parameters_'write(13,*)' '
write(13,*)' No. of target rods ',Ntargwrite(13,*)' Frac. Fail T2 Loss ',FailT2frwrite(13,*)' Ci leaked/cycle ",tlpcwrite(13,*}' H3 Ci/day source ',Cdotwrite(13,*}' '
Read(11,*)nlpb,nlpbp,nlpr
if(nlpb.gt.1}thendo k=2,nlpbRead(ll,*)Lb(k)
enddoendif
if(nlpbp.gt.1)thendo k=2,nlpbpRead(ll,*)Lbp(k)
enddo
endif
if(nlpr.gt.l}then
do k=2,nlprRead(11,*)Lr(k)
enddoendif
c The following acounts for Wfeed losses to other than turbinec building and reactor building losses.
Lxtra=O.
if(nlpb.gt.l)thendo k=2,nlpbLxtra=Lxtra+Lb(k)
enddo
endif
if(nlpbp.gt.l)then
do k=2,nlpbpLxtra=Lxtra+Lbp(k)
enddoendif
if(nlpr.gt.l)thendo k=2,nlpr
Lxtra=Lxtra+Lr(k)enddoendif
Call EVAP
Lb(1)=Wfeed-Edot(1)-LxtraLbp(1)=Edot(1)Lr(1)=Edot(2)
do k=l,nlpb
write(13,*)' loop leak rate no.',k, ' at power',Lb(k),' lbm/day+,
enddo
do k=l,nlpbpwrite(13,*)' pool leak rate no.',k,' at power',Lbp(k),' ibm/day
+,
enddo
C_24
do k=l,nlpr
write(13,*)' Ip+pl leak rate no.',k,' refueling°,Lr(k), '+ ibm/day'enddo
c tb=buildup period, daysc tr-refueling period, daysc Ncm=no. of cycles
c dtb=time step during buildup, daysc dtr-time step during refueling, daysc Cdot=H3 source rate to coolant, Ci/dayc Mo=coolant water inventory, pounds.c Mr=refuel pool inventory, poundsc decay lamda in 1/day for H3. Idk=l.53912E-04
Idk=l.53912E-4
c leak rate sumstlrb=0.tlrr=0.
tlrbp=0.
do ik=l,nlpbtlrb=tlrb+Lb(ik)
enddo
do ik=l,nlpbptlrbp=tlrbp+Lbp(Ik)
enddo
do Ik=l,nlprtlrr=tlrr+Lr(lk)
enddo
c Start time marching sequence.
nc=0
ncalcb=INT(tb/dtb)ncalcr=INT(tr/dtr)blr=Idk+tlrb/Mo
rlr=idk+tlrr/(Mo+Mr)blrp=Idk+tlrbp/Mr
DO jt=l,Ncm
do k=l,nlpbdCb(k)=Cb(k)
enddo
do k=l,nlpbpdCbp(k)=Cbp(k)
enddo
do k=l,nlprdCr(k)=Cr(k)
enddo
nc=nc+l
t=0.
write(12,*)' 'write(12,*)' 'write(12,*)'CYCLE NUMBER (',NC,')'WRITE(12,*)' 'write(13,*)' 'write(13,*)' 'write(13,*)'CYCLE NUMBER (',NC,') ************************
write(12,*)' Days Coolant H3, Ci Pool H3, ci'
_.25
Do mb=l,ncalcb Ot=t+dtb
q=Cdot/blrp=(-blr*CM3+Cdot)/(-blr)aa-blr
Lkb-q*dtb+p*(1.-exp(-aa*dtb))/aaLkbp=(CH3p/blrp)*(1.-exp(-blrp*dtb))write(13,*)' '
write(13,*)' cycle day= ',t,' (at power)'c loop leak
do J-l,nlpbCb(J}-Cb(j}+Lkb*Lb(j}/Mowrite(13,*)' Total loop leak from path',j,' ",Cb(J),' CA'
enddo
c pool leak
do j=l,nlpbpCbp(j)=Cbp(j)+Lkbp*Lbp(j)/Mr
' ' Cbp(j),' Ci'write(13,*)' Total pool leak from path',j, ,enddo
sumdk=sumdk+Lkb*Idk+Lkbp*idk
CH3=(-Cdot+(-blr*CH3+Cdot)*exp(-blr*dtb))/(-blr)CH3p=CH3p*exp(-blrp*dtb)Write(12,*)t,CH3,CH3p, ° at power'
Enddo
do k=l,nlpbsumlk=sumlk+Cb(k)
enddo
do k=l,nlpbpsumlk=sumlk+Cbp(k)
enddo
CiTOT=CH3+CH3pCH3=Mo*CiTOT/(Mo+Mr)CH3p=Mr*CiTOT/(Mo+Mr)
Do mp=l,ncalcrt=t+dtr
Lkr=((CH3+CH3p)/rlr)*(l.-exp(-rlr*dtr))write(13,*)' 'write(13,*)' cycle day= ',t,' (refueling)'do j=l,nlprCr(j)=Cr(j)+Lkr*Lr(j)/(Mo+Mr)
' ' ' Ci'' Cr(j),write(13,*) Total pl+lp leak from path',j, ,enddo
sumdk=sumdk+Lkr*ldk
CH3=CH3*exp(-rlr*dtr)CH3p=CH3p*exp(-rlr*dtr)
Write(12,*)t,CH3,CH3p,' refueling'Enddo
do k=l,nlprsumlk=sumlk+Cr(k)
enddo
source=float(nc)*Cdot*tb
ii{
netdif=source-CH3-CH3p-sumlk-sumdkwrite(14,*)' 'write(14,*)' cycle number ',ncwrite(14,*}' source total ',sourcewrite(14,*}' coolant inventory ',CH3
write(14,*)' refuel pool inv_y ',CH3pwrlte(14,*)' Cl leaked ',sumlk
write(14,*}' Ci decayed ',sumdkwrite(14,*}' net difference ",netdifsumlk-0.
write(13,*}' 'write(13,*)'Cycle ',jr,' Incremental Leak Values:'write(13,*)" 'do k=l,nlpbdCb(k}-Cb(k)-dCb(k)
' ' dCb(k} ' Ci'write(13,*)' Cycle loop leak from path',k, ,enddo
do k=l,nlpbpdCbp(k)=Cbp(k)-dCbp(k)
' ' dCbp(k) ' Ci'write(13,*}' Cycle pool leak from path',k, ,enddo
do k=l,nlprdCr(k}=Cr(k)-dCr(k)write(13,*)' Cycle pl+Ip leak from path',k,' ',dCr(k),' Ci'
enddo
ENDDO
C The final output is a compilation of Ci leaked to theC environment or directed out of the coolant-pool water inventory.
C The final cycle losses to the turbine building during power isc dCb(1), to the reactor building during power is dCbp(1), and toc the reactor building during refueling is dCr(1). All other dC'sc are "other" streams. The f_rmer terms enter the buildings' hvac.
RBconcap=dCbp(1)/(tb_24.*60.)/l.E-O6/RBacfm/28316.85RBconcrf=dCr(1)/(tr*1440.}/l.E-06/RBacfm/28316.85TBconcap=dCb(1)/(tb*1440.)/l.E-O6/TBacfm/28316.85tt=tb+tr
write(15,*)' 'write(15,*)' 'write(15,*)' TRITIUM DISPOSITION OUTSIDE REACTOR'
write(15,*)' 'write(15,*)' 'write(15,*)'The vaAues printed below can be used to estimate offsi
+re doses (from building'
write(15,*}'losses) and occupational doses (from building airborne+ concentrati_.ns}. Doses'
write(15,*)'must be computed for tritium as part of a water molecu+le (TOH or T20), not as'
write(15,*)'T2 or HT. If tritium is the only contributor to occup+ational dose, then a'
write(15,*)'limit of 2E-06 microCi/cc can be tolerated before resp+iratory protection is'write(15,*)'required.'write(15,*)' 'write(15,*)' '
write(15,*)' The following Ci and concentrations are for the tw+o periods in the'
write(15,*)tt,' day production cycle number',Ncm,' with '
C- 27
write(15,*)tb, ° days at power and',tr,' days refueling. 'write(15,*)" 'write(15,*)' '
write(15,*)' Reactor Building:'write(15,*}' '
write(15,*)' Total evap. loss during power = ',dCbp(1},' Ci'write(15,*)' Air H3 conc during power - ',RBconcap,' microCi
+/co'write(15,*)' '
write(15,*)' Total evap. loss refueling - ',dCr(1),' Ci'write(15,*}' Air H3 conc refueling - ',RBconcrf,' microCi/cc
+,
write(15,*)' '
write(15,*}' Turbine Building:'write(15,*)' •
write(15,*}' Total steam loss during power = ',dCb(1),' Ci'write(15,*}' Air H3 conc during power = ',TBconcap,' microCi
+/cc'write(15,*)' 'write(15,*)' Other Losses:'write(15,*}' 'if(Lxtra.eq.O.)thenwrite(15,*)' No other losses'
endif
if(nlpb.gt.l)thendo k=2,nlpb
' =' dCb(k) ' Ci'write(15,*)' Loop leak',k, , ,enddo
endif
if(nlpbp.gt.l)thendo k=2,nlpbp
' =' dCbp(k) ' Ci'write(15,*)' Pool leak',k, , ,enddo
endif
if(nlpr.gt.1}thendo k=2,nlpr
write(15,*)' Refueling leak',k,' =',dCr(k),' Ci'enddo
endif
creturnEND
C ************************************************************************
Subroutine INGRATE(rtau,ts,rint)
Tl=rtau*86400.t =.0100503,T1
i_(ts.le.t )rint=.01*ts
if(ts gt t_)thenal=.01*t
a2=(ts-t_)+Tl*(exp(-ts/Tl)-exp(-t_/Tl))rint=al+a2
endif
RETURNEND
C ************************************************************************
C-28
SUBROUTINE GETARG
REAL lambdaT, len,N_Anrg,N_Arate,mwllalo2,1t,mgbol,+mgeol,Li6depl,d(0_11),rcm(0_11},A(lO),N A
COMMON/BLK1/Nt,A,d,len,plen,N_A,ntgr,TGASR,gammaen,enrichli,+FPcycle,EOcycle,pT2,TpelletCOMMON/BLK5/tfluid,tcladfCOMMON/BLK6/YSfluid,YS10OF,PfluldCOMMON/BLKIO/GVR,GVRo
open(Unlt=4,File-'t2p2-3.out',status='unknown'}
c Data d/.5893,.6096,.9906,1.0084,1.049,1.0592,1.0744,1.2268,4,0./c nt=7
do l=0,Nt
rcm(1)=d(1)/2.enddo
teol=S6400.*FPcycle
rcladod=rcm(7)rcladid=rcm(6)ralumid=rcm(5)rgetod=rcm(4)rgetid=rcm(3)rpelod=rcm(2)rpelid=rcm(1)rlinrid=rcm(0)
pi=3.14159
fen=381.
plen=25.4lambdaT=l.792E-9
N_Anrg=4.8*l.6E-13N Arate=N A*len
c FPcycle=284.c EOCycle=378.
TGr=460.+600.
if(ntgr.eq.2)TGr=TGASRVcladid=pi*ralumid**2.*(len+plen)ccor=ii.94/12.
Vpellet=pi*(rpelod**2.-rpelid**2.)*len*ccorVgetter=pi*(rgetod**2.-rgetid**2.)*len*ccor
Vget Ni=pi*((rgetod-.00127)**2.-(rgetid+.00127)**2.)*len*ccorVliner=pi*(rpelid**2 -rlinrid**2.)*len*ccor
Vspring= 3.07Vnet=Vcladid-(Vliner+Vgetter+Vpellet+Vspring)
VVR=Vnet/VpelletTotrxns=N Arate*teol
T2molec=Totrxns/2.Heatoms=Totrxns
totmoles=(T2molec+Heatoms)/6.023E+23GVR=totmoles*22414./VpelletGVRo=exp(6.44-Tpellet*4.29E-03)sat2=9720.* 6.0321
c
if(ntgr.eq.l} goto 10write(4,*}' 'write(4,*)' 'write(4,*}' ********GVR/PRESSURE/STRESS CALCULATIONS********'write(4,*}' '
wrlte(4,*}'Thls output focuses on the end of the target irradiatlo+n period (eol)'write(4,*}'in the reactor. At this point the Anterior pin pressur
+e is at maximum'
write(4,*)'while still at power with the coolant at maximum temper+ature. The'
write(4,*}'coolant is then cooled to 100F. Symbol definitions:'wrlte(4,*)' 'write(4,*)' VVR - void volume/pellet volume'write(4,*}' 90%YS = 90% yield stress'write(4,*)' TH/Zr ratio = atoms of H+T/atoms Zr'write(4,*)' 'write(4,*)' '
e , t , #write(4,*)' Vcladid ',Vcladid, Vpellet ,Vpellet, cm* 3write(4,*)' Vgetter ',Vgetter,' Vget Ni ',Vget Niwrite(4,*)' Vliner ',Vliner ,' Vsprlng ,Vspr_ngwrite(4,*)' Vnet ',Vnet ,' VVR ',vvrwrite(4,*)' GVR ',GVRwrite(4,*)' GVRo ',GVRo
I0 continue
rhliaio2=2.57
mwlialo2=enrichli*6.02+(l.-enrichli)*7.02+26.89+32.gmlialo2=Vpellet*rhlialo2/mwlialo2atomsLi6=gmlialo2*6.023E+23*enrichli
Li6depl=100.*Heatoms/atomsLi6ccf=lO0.*FPcycle/EOcycleif(ntgr.eq.l)go to iiwrite(4,*)' '
write(4,*)' Li6 depletion % ',Li6deplwrite(4,*)' '
write(4,*)' Reactor Cycle Capacity Factor ,ccf,ii continue
totHe=Totrxns
totT2ndk=TotHe/2.It=l_,bdaT*teol
totT2wdk=totT2ndk*(l.-exp(-it))/itgmoleT2=totT2wdk/6.023e+23gmoleH2=0.1*gmoleT2
gmoleZr=Vget_Ni*6.4/91.22TH Zrrat=2.*(gmoleT2+gmoleH2)/gmoleZrm
CiT2=sat2*gmoleT2
c Fraction of T2 in gas space, frgasT2pT2=pT2/I.01325E+5
gasT2=pT2*Vnet/82.06/(TGr/l.8)totT2=totT2ndk/6.023E+23frgasT2=gasT2/totT2
if(ntgr.eq.1)go to 20write(4,*)' '
write(4,*)' T2 molecules, no decay ',totT2ndkwrite(4,*)' T2 molecules, w/ decay ',totT2wdk
' ' TH/Zr ratio ',TH Zrratwrite(4,*)' T2 curies ,CiT2, m
write(4,*)' Fraction of T2 in gas ',frgasT2
C'-- 30
20 continue
Aclad=7.92*gammaenAliner=6.4*gammaen
Agetter-gammaen*((Vgetter-Vget_ni}*8.9+Vget Ni*6o4)/VgetterAlialo2=gammaen*2.57 + N_Arate*NAnrg/Vpelletif(ntgr.eq.1)go to 30write(4,*}' 'write(4,*}' heat generation watts/cm**3: 'write(4,*)" cladding ',Aclad
write(4,*}' getter ',Agetterwrite(4,*)' liner ',Alinerwrite(4,*)' pellet ',Alialo2
30 continue
A(1)=AgetterA(2}-Alialo2A(3}=O.A(4}=AgetterA(5)=0.A(6)=AcladA(7)=A(6)
pressI=(GVR*2./3./VVR+I.)*TGr/492.mgbol=Vnet/22414.mgeol=mgbol+totHe/6.023E+23press2=mgeol*82.06*(TGr/l.8)/Vnetpress3=14.7*mgeol*82.06*560./l.8/Vnet
pressl=14.7*presslpress2=14.7*press2rd=(rcm(7)+rcm(5))/2.
CALL STRESS
ysg_l=.9*YSfluid*1000.ys9_2=.9*YSIOOF*I000.presnetl=-Pfluid+press2-14.7Pfi00=12.9159
presnet2=press3-PflO0-14.7hoopstl=presnetl*rd/(rcm(7)-rcm(5))Axialstl=hoopstl/2.hoopst2=presnet2*rd/(rcm(7)-rcm(5))Axialst2=hoopst2/2.
C Calculation of puff releases of gas phase tritium with sudden failureC of pin cladding, press4 ks at 75F.
press4=press3*535./560.
gasCi=CiT2*frgasT2puffl=(press2-Pfluid-14.7)*gasCi/press2puff2=(press3-Pfl00-14.7)*gasCi/press3puff3=(press4-14.7)*gasCi/press4
if(ntgr.eq.l)go to 40write(4,*)' '
write(4,*)' internal pressure at power',press2,' psi'
write(4,*)' internal pressure at 100 F',press3,' psi'write(4,*)' '
write(4,*)' hoop stress at power',hoopstl,' psi'write(4,*)' axial stress at power',axialstl,' psi'write(4,*)' '
' ' psi'write(4,*)' hoop stress at 100 F ,hoopst2,writ (4,*)' axial stress at 100 F',axialst2,' psi'
(- 31
write(4,*)' '
write(4,*}' 90%YS@pwr ',ys9 1,' 90%¥S100F ',y89_2,' psi'' netDP100F ',presnet2,write(4,*}" netDP@pwr ',presnetl, ' psi'
write(4,*}' '
write(4,*} ° Puff T2 Ci released from sudden eol pin failures'write(4,*}' ',puff1,' at power to coolant'
' ' ' @ 100F in pool'write(4,*) ,puff2,
write(4,*}' ',puff3,' @ 75F in air at 1 atm'40 continue
returnend
C ************************************************************************
Subroutine EVAP
C EVAP calculates pool water loss and water concentration of the ABWRC reactor building airspace.
Real Edot(2)COMMON/blk3/edot,rbacfm
Ci=3.19E-5
Cp=l.58E-4
Ak=3531.47,3.c Ak An cfm
c Ci = inlet air water conc., #moles/ft3. (50% rh at 70 F)c Cp= pool surface vapor conc (100% rh at 100 F)c Ak = mt coeff * pool area, cfmc A=600 m2
c k=30OOcm/hr (Fig.15-5, Lyman, W.J., et al. 1982. "Handbook of Chemical
c Property Estimation Methods." McGraw-Hill, NY.)c Co= reactor outlet vapor conc.,Ibmole/ft**3c Edot= evaporation rate, ibm/day
do j=2,1,-1acfm=rbacfm
Co=(Ci*acfm+Ak*Cp)/(acfm+Ak)
c Co_=Ak*Cp*lS./(acfm+Ak)/28316.85c Co_=pool added vapor conc, ibm/cc
Edot(j)=Ak*(Cp-Co)*18.*1440.Co =Edot(j)/1440./acfm/28316.85
c write(2,*)acfm,Co,Co_,Edot,AkAk=Ak/2.
enddo
return
end
C ******************************************************************
subroutine FAIL
c Program FAIL calculates the target pin-to-coolant source term asc fraction of tritium produced.
REAL Loss(20,20),Tbp(21},F(20),Fi(20),FR(20,20),npinsreal A(lO),d(O:11),n aCOMMON/BLK4/FailT2fr?rrctrc,tb
COMMON/BLK1/Nt,A,d,len,plen,N_A,ntgr,TGASR,gammaen,enrichli_+FPcycle,EOcycle,pT2,Tpellet
d 32
COMMON/BLKI0/GVR,GVRo
c nsdpc - number of undesired shutdowns per cyclec tap = days at full power in target cyclec npp - number of periods of power/cycle - nsdpc+l
nsdpc=3*nint(rrctrc)npp=nsdpc+nint(rrctrc)tap-FPcycleif(rrctrc.lt.l.01)thennsdpc=2npp=nsdpc+l
endif
npp=nsdpc+lnpins=3488.
c dos = day of pellet saturation with tritiumdos=tap*GVRo/GVR
c nps = period of power where saturation occursnps=npp
tpp=Tap/float(npp)c tpp = days of power period between shutdowns
Tbp(1)=O.do j=2,npp+lTbp(j)=Tbp(j-l)+tppif(Tbp(j).gt.dos.and.Tbp(j-l).le.dos)nps=j-i
enddo
c Define X, the fraction of period nps that is unsaturated.split=dos-float(nps-l)*tppX=split/tpp
c Define Loss(J,K) = fraction of tritium produced in K period due toc failure in J period.
DO J=l,npp
Do K=l,nppif(K. le.J)Loss(J,K)=0.if(K.gt.J)Loss(J,K)=0.03
if(K.gt.J.and.K.eq.nps)Loss(J,K)=.03*X+l.-Xif(K.gt.J.and.K.gt.nps)Loss(J,K)=l.
Enddo
ENDDO
c Weibull Statistics for failure frequency:
c Define F(J) as cumulative probability of failure over interval J.
c (see Mitchell, R.A..1967. Introduction to Weibull Analysis. PWA3001.c Pratt & Whitney Aircraft, E. Hartford, CT.)
c Weibull parametrs beta, eta, to.beta=l.045
c beta from Trans ANS 18, p125 (1974).to=Tbp(2)
eta=(3.*tb-to)/exp(-9.21029/beta)
c above eqn assumes failure rate of 1/10000/3 cycles(same as fuel pin).
F(1)=0.
C- 33
Fi(1)-O.do j=2,nppFi(J)=l.-exp(-((Tbp(J+l)-to)/eta)**beta)F(J}=Fi(J)-Fi(j-I)
enddo
FT=0.
do j=l,nppFT=FT+F(J)
enddo
T21oss=O.
Do j=l,nppdo k=l,nppFR(j,k}=Loss(j, k}*F(j )T21oss=T21oss+FR(j,k)
enddoEnddo
T21oss=T21oss/float(npp)FailT2fr=T21oss
c T21oss = fraction of tritium lost/cycle due to pin failure.
returnend
***********************************************************************
SUBROUTINE STRESS
COMMON/BLK5/tfluid,tcladfCOMMON/BLK6/YSfluid,YSIOOF,Pfluid
c Vapor pressure of water at temperature.
AAP=3.2437814BP=5.868263E-3CP=1.17023793E-08DP=2.1878462E-3a0=96.11825al=-3.02965E-2
tf=tfluid
tk=(TF-32.)/l.8 + 273.16
xp=647.27-tk
pvatm=218.167*lO.**(-(xp/tk)*(aap+bp*xp+cp*xp**3.)/(l.+dp*xp))
pfluid=14.7*pvatm
ysfluid=a0+al*tcladfyslOOF=91.44
returnend
C" 34
APPENDIX DRADIOLOGICAL SAFETY REQUIREMENTS AND CRITERIA
I. GENERAL
This section presents the safety standards, regulations and criteria agaimt which SMP will be comparedin order to demomtrate safety adequacy. In summary, the design, construction, commissioning andoperation of SMP will be carried out in accordance with the requirements of the THORP Division SiteLicense Regulations (SLRs). In addition the following specific regulations and criteria are to be appliedto SMP'
a) Radiological Protection Regulations (RPRs) - THORP Division; these will apply to normaloperational exposure.
b) Environmental Protection Regulations (EPRs) - THORP Division; these will apply to environmentaldischarges.
c) Radiological accident risk criteria for Sellafield reprocessing divisions and Drigg; these will applyto accidents.
The standards, regulations and criteria applicable to normal operations are summarized in Sectiom 2, 3and 4. The criteria for accidents are summarized in Section 5.
II. OCCUPATIONAL RADIATION EXPOSURE CONTROL
The operational radiation exposure in SMP will t_e in accordance with the requirements of the THORPDivision RPRs. The main requirements are summarized below:
* All the operations which lead to whole body exposure must be shown to be ALARP.
• The average radiation exposure, of the group of workers associated with the plant, should notexceed 5 mSv whole body dose (the sum of effective dose equivalent (EDE), from external radiationand committed effective dose equivalent (CEDE), from ingestion of radioactive material) per
calendar year, unless a special justification on ALARP grounds is authorized by the appropriateDivisional Director.
• The maximum whole body dose (EDE plus CEDE) for any individual worker should not exceed
15 mSv in any one year.
• The maximum extremity dose for any individual should not exceed 300 mSv per year.
III. ROUTINE AERIAL EFFLUENT DISCHARGES
The routine aerial effluent discharges from SMP will be in accordance with the requirements of theTHORP Division EPRs. This requires that all radioactive aerial discharges must comply with theconditions of the Discharge Authorizations. Specific requirements include:
• All plant will be designed and operated using Best Practicable Means to ensure that doses toindividual and collective doses are ALARP.
Appendix D- 1
• The quantities of radionuclides discharged to atmosphere from specified discharge points in anyperiod of 24 hours will be limited.
• For the identified discharge points the quantities of radionuclides discharged to atmosphere in anyquarter above specified levels are notifiable.
• The quantities of radionuclides discharged to atmosphere in any calendar year are limited.
• Sampling and measurement arrangements for discharge accountancy are specified.
• Formal records of discharges are maintained, then summarized, and forwarded to the AuthorizingDepartments within specified time periods,
At this stage of the project, SMP has not been allocated a discharge target. However, SMP will need
to be designed to take account of these requirements and will ultimately receive a discharge allocationin line with the requirements: this allocation constituting a percentage of the THORP division dischargeallocation.
IV. ROUTINE LIQUID EFFLUENT DISCHARGES
The routine liquid effluent discharges from SMP will be in accordance with the requirements of theTHORP Division EPRs. This requires that all discharges to sea of low active liquid waste from theSellafield site must comply with the conditions of the Discharge Authorizations. Specific requirementsare that:
• All plant will be designed and operated using Best Practicable Means to ensure that doses toindividual and collective doses are ALARP.
• The quantities of radionuclides discharged to sea in any period of 24 hours are limited.
• The quantities of radionuclides discharged to sea in any quarter are limited.
• The quantities of radionuclides discharged to sea in any calendar year are limited.
• Sampling and measurement arrangements for discharge accountancy are specified.
• Formal records of discharges are maintained, then summarized, and forwarded to the AuthorizingDepartments within agreed time periods.
At this stage of the project, SMP has not been allocated a liquid effluent discharge target. However,SMP will ultimately receive a discharge allocation in line with the above requirements; this allocation willcontribute to the overall THORP division discharge allocation.
V. ACCIDENT CONDITIONS
The risk criteria against which accidents associated with SMP will be compared are summarizedbelow. In terms of these criteria SMP is considered as 'one plant'.
Appendix D -2
V.A Accidental Risk Criteria for the Workforce (Internally Initiated Events)
a) The summed frequency of accidents (including criticalities) in a building, which could give dosesto any member of the workforce working in that building, from direct radiation, inhalation andingestion should be less than the values given below:
Effective Dose (mSv) Summed Frequency(per year)
.... 50"- 10()0 ' 10 .3
> I000 10.5L.....
b) The summed frequency of Building Evacuations due to abnormal airborne contamination (exceeding100 DAC) and high gamma dose rates (exceeding 0.2 mSv hr j) in large parts of an operating areaof a plant should be less than 10.2 y_.
c) The summed frequency of unplanned criticality events should be less than 104 yJ.
Interpretation, For the purpose of these criteria, an internally initiated event is defined as an event withan initiator on the plant or any other plant on the site where the effect on the plant being assessed is via
process or service routes. Consequently, events initiated by local road and rail vehicles, cranes andmobile platforms should be included. External events can be naturally occurring (e.g., earthquakes),man-made but outside the Company's control (e.g., aircraft crashes) and catastrophic failure of otherplant on the site (e.g., a turbine failure) where the initiator is not linked to the plant being assessed viaa process or service route.
These criteria apply to buildings, as opposed to the 'plant' concepts used fc:" accidental releases affectingthe public. However, a building should include subsidiary and ancillary buildings which would beregarded as part of a main process building as far as the process operators would be concerned. (Thiswill usually coincide with the scope section of a safety case.)
The criterion in V.A (b) above, applies only within buildings. It is not applicable to restricted accessareas, where higher airborne activity or radiation levels are to be expected and appropriate precautionstaken. The other criteria apply to all areas.
10_sfissions should be used for the reference criticality event in estimating consequences - V.A(a), unlessanother more appropriate value is agreed with the Site Nuclear Safety Officer.
Doses from contaminated wounds are not included in these criteria: qualitative assessment will suffice.
V.B Criteria for Accidental Aerial Discharges
a) The time averaged critical group dose from accidental aerial discharges from a plant should notexceed 4 #Sv per year.
Appendix D-3
b) The summed frequency of accidents on a plant, which would give doses to a member of the criticalgroup, should be less than the values given below:
--
Effective Dose (mSv) Summed Frequency(per year)
0.01 - 1 10"z
1 - 10 10.3
10- 100 104
100- 1000 10.5
> 1000 10.6......
c) The summed frequency of very large accidents on a plant, i.e., those with the potential to give arelease to the environment with societal consequences equivalent to:
- 100 deaths, e.g., a release 10,000 TBq of Iodine 131, 5 TBq Pu.or
- Land contamination effects equivalent to that from 200 TBq of Caesium 137.
should be less than 10 .6 pa.
Interpretation. For the purpose of these criteria, an internally initiated event is definec as an event with
an initiator on the plant or any other plant on the site where the effect on the plant being assessed is viaprocess or service routes. Consequently, events initiated by local road and rail vehicles, cranes andmobile platforms should be included. External events can be naturally occurring (eg earthquakes), man-made but outside the Company's control (eg aircraft crashes) and catastrophic failure of other plant on
the site (eg a turbine failure) where the initiator is not linked to the plant being assessed via a process orservice route.
In setting the criterion in V.B(a) above, allowance has already been made for the variations which occurin wind direction (wind direction is not relevant to the other criteria - Paras (b) and (c).
Events having consequences less than 10 _Sv (all pathways) should be ignored. For effective doses upto 10 mSv, all pathways dose should be used. For effective doses greater than 10 mSv, inhalation anddirect radiation doses only should be used - Paras (a) and (b).
V.C Criteria for Accidental Liquid Effluent Discharges
a) The time averaged critical group dose from accidental marine discharges from a plant should notexceed 0.4 pSv per year.
Appendix i") .4
b) The summed frequency of accidents on a plant, which could give doses to a member of the publicoutside the site arising from discharges to sea or waterways, directly or via engineered routes,should be less than the values given in the following table:
Effective Dose (mSV) Summed Frequency(per year)
.... 1o"'-loo 1or
> I00 10"4
Interpretation. For the purpose of these criteria, an internally initiated event is defined as an event withan initiator on the plant or any other plant on the site where the effect on the plant being assessed is via
process or service route. Consequently, events initiated by local road and rail vehicles, cranes and mobileplatforms should be included. External events can be naturally occurring (eg earthquakes), man-madebut outside the Company's control (eg aircraft crashes) and catastrophic failure of other plant on the site(eg a turbine failure) where the initiator is not linked to the plant being assessed via a process or serviceroute.
The criteria apply to all accidental liquid discharges, not just those via the sea lines. However, eventshaving consequences less than I0 _tSv should be ignored,
Assessment of consequences and frequencies should include consideration of clean-up plants outside thesubject plant, their reliability, and the possibility of detection and associated diversion of liquors outsidethe plant on Site where appropriate.
V.D Accident Risk Criteria for Externally Initiated Events
Seismic Events
a) Plant design shall be examined without seismic provisions to determine the expected dose to amember of the public from inhalation and direct radiation following a design basis seismic event(such an event shall be defined as having an annual probability of exceedence of 104).
- Where the dose is greater than 5 roSy, then the plant will be seismically qualified to show that
the expected dose to a member of the public off site is less than 5 roSy following the designbasis event.
- Where the expected dose to a member of the public off site is greater than I mSv but less than5 roSy, then the plant should be assessed to show that it will withstand an event with annual
probability of exceedence of 103.
- Where the dose is less than I mSv, no special seismic provisions are required beyond the
general requirement to satisfy ALARP.
b) Seismically qualified plants designed to Paragraph (a) should be examined to show there are nocliff-edge effects on release for events beyond the design events specified. This can be
demonstrated by considering a more severe event. If this cannot be demonstrated, appropriateaction should be taken where this is shown to be reasonably practicable.
Appendix'D-5
c) Where it is considered feasible that acute health effects (doses > I Sv) to workers may occurfollowing a seismic event, then the plant should be assessed to show that following a seismic eventwith annual probability of exceedence of I03, personnel escape routes from areas with highoccupancy are likely to be available.
Extreme Wind
a) It should be shown by deterministic argumentsthat containment barriers remain and any safetysystems which are required during or after the event can survive an extreme wind with an annualprobability of exceedence of 104, for those plants with potential to give a dose to a member of thepublic off site more than 5 mSv in the event of such a wind.
Notes:
- This may entail considerationof containment structures, operatingarea structures(eg buildingshell), ventilation, cooling services etc.
- This may be achieved by design and assuring that vulnerable inventories or plant areprotected, or can be protected given advanced notice of the event, from damage.
b) Plant with radioactive material directly exposed to the wind (such as, ttdoor ponds), should beexamined against extreme wind with an annual probability of exceedence of 10'*yt and it be shownthat the consequences so far as is reasonably practicable to a member of public off site are less than5 mSv from inhalation and direct radiation.
Precipitation
a) The direct effects of rainfall, local accumulation of rain water and rain falling on or around abuilding or site should be examined against site-specific data, taking account of interactions withabnormal tidal effects as appropriate. It should be shown that radioactive material is protected fromrainfall by a structure and is out of reach of flood water, or that facilities exist to containcontaminated run-off/overflow.
b) The impact of snow and its accumulation should be addressed deterministically, taking account ofcounter-measures where appropriate.
Temperatures and Drought
a) Extreme temperatures and drought do not occur unexpectedly. Where special provisions forpredicted extremes are not incorporated in design, it should be shown that contingency arrangementsto assure safety in extremes occurring within periods of abnormal temperatures and drought can beintroduced in a timely manner.
Aircraft Crashes
a) Where it can be demonstrated that the frequency of aircraft crashes with the potential to lead to arelease of radioactive material off site, taking account of pilot action and exclusion zones asappropriate, is not more than 10.7y.i per plant, then no further action to deal with aircraft crashesis required.
Appendix _I_-6
b) If the above criterion cannot be met then it should be shown that the conditions of paragraph (c)below are met by regarding an air crash as another ma" made external hazard.
Other Man Made External Events
a) Detailed consideration need only be given to the possibilities of off site explosion or gas cloudsaffecting the safety of plant if the site is within:
- 2km of a NIHHS notifiable site
The consultation planning approval distance of a NIHHS notifiable site
- 2km of a regularly used transport route for the movements of materials with the potential foraffecting the site
- Or 2km of a pipeline with a potential for affecting the site.
b) The effects of a catastrophic failure of the plant being assessed on other plants on the site (wherethe initiator is not linked to the plant via a service or process route) should meet the criteria in
Paragraph (c) or (d) below.
c) No action should be taken if it can be demonstrated that the frequency of any event affecting the
plant likely to lead to the release of radioactive material is less than I0 7 y.i, or the potentialexpected dose to a member of the public following the event is less than I mSv from a plant (takingaccount of the potential amount and type of external substance involved, the distance from the plant,
topography, wind direction and other relevant factors).
d) Where Paragraph (c) cannot be met it should be shown as far as reasonably practicable for eachexternal hazard that the best estimate consequences to a member of the public off site from a plantare less than the values in the table below:
Predicted frequency, f, per Effective Doseyear per event
per plant
10.7 < f < 10.6 I SvI0 6 < f < I0 5 100mSv10.5 < t < I0 4 10mSv
I0-4 < f ImSv
e) Where application of this proves difficult, the advice of the Director of Health Safety andEnvironmental Protection should be sought so that criteria can be developed to suit the needs of the
specific case.
Appendix D -7
III