radionuclide release calculations for selected severe

XA04NO368NUREG/CR-4624

fNIS-XA-N--072 BMI-2139Vol. 2

Radionuclide Release Calculations forSelected Severe Accident Scenarios

PWR, Ice Condenser Design

Prepared by R. S. Denning, J. A. Gieseke, P. Cybulskis, K. W Lee,H. Jordan, L. A. Curtis, R. F. Kelly, V. Kogan, P. M. Schumacher

Battelle's Columbus Division

Prepared forU.S. Nuclear RegulatoryCommission

NOTICE

This report was prepared as an account of work sponsored by an agency of the United StatesGovernment. Neither the United States Government nor any agency thereof, or any of theiremployees, makes any warranty, expressed or implied, or assumes any legal liability of re-sponsibility for any third party's use, or the results of such use, of any information, apparatus,product or process disclosed in this report, or represents that its use by such third party wouldnot infringe privately owned rights.

__J

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NUREG/CR-4624BMI-2139Vol. 2

Radionuclide Release Calculations forSelected Severe Accident Scenarios

MR, Ice Condenser Design

Manuscript Completed: May 1986Date Published: July 1986

Prepared yR. S. Denning, J. A. Gieseke, P. Cybulskis, K. W. Lee,H. Jordan, L. A. Curtis, R. F. Kelly, V. Kogan, P. M. Schumacher

Battelle's olumbus Division505 King- AvenueColumbus, OH 43201

Prepared forDivision of Risk Analysis and OperationsOffice of Nuclear Regulatory ResearchU.S. Nuclear Regulatory CommissionWashington, DC 20555NRC FIN A1322

ABSTRACT

This report presents results of analyses of the environmental releases offission products (source terms) for severe accident scenarios in a pressurizedwater reactor with an ice-condenser containment. The analyses were performedto support the Severe Accident Risk Reduction/Risk Rebaselininq Proqram (SARRP)which is being undertaken for the U.S. Nuclear Regulatory Commission bySandia National Laboratories. In the SARRP nroqram, risk estimates are beinggenerated for a number of reference plant designs. The equoyah Plant hasbeen used in this study as an example of a PWR ice-condenser plant,

TABLE OF CONTENTS

ELge

1. INTRODUCTION ........................................ . ............ 1-1

.2. GENERAL APPROACH ................................................. 2-1

2.1 Source Term Code Package .................................... 2-12.2 Radionuclide Groups ......................................... 2-4

:3. DESCRIPTION OF PLANT AND ACCIDENT SCENARIOS ...................... 3-1

3.1 Accident Sequences Considered ............................... 3-13.2 Primary System Flowpaths .................................... 3-33.3 Containment Flowpaths ....................................... 3-73.4 Containment Failure Mode and Pressure Level ................. 3-19

4. BASES FOR TRANSPORT CALCULATIONS ................................. 4-1

4.1 Phenomenological Modeling Assumptions ....................... 4-14.2 Results of Thermal Hydraulic Analyses ....................... 4-2

4.2.1 S3HF Sequence ......................................... 4-24.2.2 TB Sequence ........................................... 4-40

4.3 TMLU-SGTR Sequence .......................................... 4-544.4 TBA Sequence ................................................ 4-644.5 Additional Sequences Considered ............................. 4-76

4.5.1 S3H Sequence .......................................... 4-844.5.2 S2HF Sequence ......................................... 4-89

4.6 Radionuclide Sources ........................................ 4-91

4.6.1 Sources Within Pressure Vessel ........................ 4-914.6.2 Sources Within the Containment ........................ 4-91

5. RADIONUCLIDE RELEASE AND TRANSPORT ............................... 5-1

5.1 S3HF Sequence ............................................... 5-1

5.1.1 Release and Transport in RCS ........................... 5-15.1.2 Release and Transport in Containment for

S3HF1 Scenario ......................................... 5-85.1.3 Release and Transport in Containment for

S3HF2 Scenario ......................................... 5-145.1.4 Release and Transport in Containment for

S3HF3 Scenario ......................................... 5-14

v

TABLE OF CONTENTS(Continued)

f�qe

5.2 TB Sequence ................................................. 5-21

5.2.1 Release and Transport in the RCS ...................... 5-215.2.2 Release and Transport in Containment .................. 5-21

5.3 TMLU-SGTR Sequence .......................................... 5-335.4 TBA Sequence ................................................ 5-41

5.4.1 Release and Transport in the RCS ...................... 5-415.4.2 Release and Transport in Containment .................. 5-55

5.5 Noble Gas and Energy Release to Environment ................. 5-61

5.6 Icebed Decontamination ...................................... 5-61

6. SUMMARY AND CONCLUSIONS .......................................... 6-1

7. REFERENCES ....................................................... 7-1

vi

LIST OF FIGURES

f,�2e

Figure 2.1 Source Term Code Package ................................. 2-2

Figure 31 Primary System Flowpath for Sequoyah Seal LOCA(S3) Sequences ........................................... 3-4

Figure 32 Schematic of TRAP-MERGE Control Volumes for SequoyahSeal LOCA (S3) Sequence .................................. 3-5

Figure 33 Fission Product Flowpaths for TMLU-SGTR Sequence ......... 3-6

Figure 34 Primary System Flowpaths for Steam Generator TubeRupture Accident ......................................... 3-8

Figure 35 Primary System Flowpath During Initial Core MeltPhase for Sequoyah TBA Sequence .......................... 3-9

Figure 36 Schematic of TRAP-MERGE Control Volumes DuringInitial Core Melt Phase for Sequoyah TBA Sequence ........ 3-10

Figure 37 Primary System Flowpath During Later Core Melt Phasefor Sequoyah TBA Sequence ................................ 3-11

Figure 38 Schematic of TRAP-MERGE Control Volumes During LaterCore Melt Phase for Sequoyah TBA Sequence ................ 3-12

Figure 39 Containment Fission Product Flowpaths for Sequoyah S3HF1. 3-13



Figure 312 Containment Fission Product Flowpaths for Sequoyah TB(S3B) .................................................... 3-17

Figure 313 Containment Fission Product Flowpaths for SequoyahTBA Sequence ............................................. 3-18

Figure 41 Primary System Pressure Response or the S3HF Sequence... 4-13

Figure 42 Primary System Coolant Leakage for the S3HF Sequence ..... 4-14

Figure 43 Primary System Water Inventory for the S3HF Sequence ..... 4-15

Figure 44 Maximum and Average Core Temperatures for the SUFSequence ................................................. 4-16

Figure 45 Fractions of Core Melted and Cladding Reacted forthe SU F Sequence ........................................ 4-17

Figure 46 Containment Pressure Response for S3HF1 .................. 4-18

vii

LIST OF FIGURES(Continued)

PaVeFigure 47 Containment Temperature Response for S3HF1 ............... 4 9

Figure 48 Containment Sump and Reactor Cavity WaterInventories for S3HF1 .................................... 4-21

Figure 49 Containment Sump and Reactor Cavity WaterTemperatures for S3HF1 ................................... 4-22

Figure 410 Progression of Concrete Attack for S3HF1 ................. 4-23

Figure 4.11 Ice Inventory for S3HF1 .................................. 4-24

Figure 4.12 Volume of Gases Leaked for S3HF1 ......................... 4-25

Figure 4.13 Containment Pressure Response for S3HF2 .................. 4-31

Figure 4.14 Containment Temperature Responses for S3HF2 .............. 4-32

Figure 4.15 Containment Sump and Reactor Cavity WaterInventories for S3HF2 .................................... 4-33

Figure 4.16 Containment Sump and Reactor Cavity WaterTemperatures for S3HF2 ................................... 4-34

Figure 4.17 Progression of Concrete Attack for S3HF2 ................. 4-35


Figure 4.19 Total Volume of Gases Leaked for S3HF2 ................... 4-37

Figure 4.20 Containment Pressure Response for S3HF3 .................. 4-38

Figure 4.21 Containment Temperature Responses for S3HF3 .............. 4-39

Figure 4.22 Containment Sump and Reactor Cavity WaterInventories for S3HF3 .................................... 4-41

Figure 4.23 Containment Sump and Reactor Cavity WaterTemperatures for S3HF3 ................................... 4-42

Figure 4.24 Progression of Concrete Attack for S3HF3 ................. 4-43


Figure 4.26 Total Volume of Gases Leaked for S3HF3 ................... 4-45

Figure 4.27 Primary System Pressure Response for S3B ................. 4-47

Figure 4.28 Primary System Leak Rates for SH ........................ 4-48

Figure 4.29 Primary System Water Inventory for S3B ................... 4-49

Figure 4.30 Maximum and/-Average Core Temperatures for S3B ............ 4-50

viii


Page

Figure 431 Fractions of Clad Reacted and Core Melted for S3B ........ 4-51

Figure 432 Containment Pressure Response for S3B .................... 4-52

Figure 433 Containment Temperature Responses for S3B ................ 4-53

Figure 434 Containment Sump and Reactor Cavity WaterInventories for S3B ...................................... 4-55

Figure 435 Containment Sump and Reactor Cavity, WaterTemperatures for S3B ..................................... 4-56

Figure 436 Progression of Concrete Attack for S3B ................... 4-57

Figure 4.37 Ice Inventory for S3B .................................... 4-56

Fi(ure 438 Total Volume of Gases Leaked for SB ..................... 4-59

Fi(ure 439 Steam Generator Secondary Side Water InventoryFor TMLU-SGTR ............................................ 4-61

Figure 440 Primary System Pressure Response for TMLU-SGTR ........... 4-62

Figure 441 Primary System Water Inventory for TMLU-SGTR ............. 4-63

Figure 4.42 Maximum and Average Core Temperatures for TMLU-SGTR ...... 4-65

Figure 4.43 Fractions Cladding Reacted and Core Melted for TLU-SGTR. 4-66

Figure 4.44 Leakage Through Pressurizer Relief Valve for TMLU-SGTR ... 4-67

Figure 4.45 Leakage Through Ruptured Steam Generator Tubes forTMLU-SGTR ................................................ 4-68

Figure 4.46 Steam Generator Secondary Side Water InventoryDuring TBA Sequence ...................................... 4-70

Figure 4.47 Primary System Pressure History for, TBA Sequence ......... 4-71

Figure 4.48 Primary System Water Inventory for TBA Sequence .......... 4-72

Figure 4.49 Maximum and Average Core Temperatures for TBA Sequence ... 4-73

Figure 4.50 Fractions of Cladding Reacted and Core MeltedFor TBA Sequence ......................................... 4-74

Figure 4.51 Temperatures of Gases Leaving the Core and Leakingto Containment for TBA Sequence .......................... 4-75

ix


Page

Figure 452 Containment Pressure Response for TBA Sequence ........... 4-77

Figure 453 Containment Temperature Response for TBA Sequence ........ 4-78

Figure 4.54 Ice inventory for TBA Sequence ........................... 4-79

Figure 4.55 Containment Sump and Reactor Cavity WaterInventories for TBA Sequence ............................. 4-80

Figure 4.56 Concrete Attack for TBA Sequence ......................... 4-81

Figure 4.57 Total Volume of Gases Leaked for TBA Sequence ............ 4-82

Figure 4.58 Noble Gas Distribution for TBA Sequence .................. 4-83

Figure 5.1 Mass of CsI Released from Indicated RCS Component as aFunction of Time--S3HF Sequence .......................... 5-4

Figure 5.2 Mass of CsOH Released from Indicated RCS Componentas a Function of Time--S3HF Sequence ..................... 5-5

Figure 5.3 Mass of Te Released from Indicated RCS Component as aFunction of Time--S3HF Sequence .......................... 5-6

Figure 5.4 Mass of Aerosol Released from Indicated RCS Componentas a Function of Time--S3HF Sequence ..................... 5-7

Figure 5.5 Schematic Diagram Showing Containment CalculationProcedures for the S3HF1, S3HF2, S3HF3, and TB Scenarios. 5-10

Figure 5.6 Mass of CsI Released from Indicated RCS Component asa Function of Time--TB Sequence .......................... 5-25

Figure 5.7 Mass of CsOH Released from Indicated RCS Component asa Function of Time--TB Sequence .......................... 5-26

Figure 5.8 Mass of Te Released from Indicated RCS Component as aFunction of Time--TB Sequence ............................ 5-27

Figure 5.9 Mass of Aerosol Released form Indicated RCSComponent as a Function of Time--TB Sequence ............. 5-28

Figure 5.10 CsI Behavior in Steam Generator Secondary ................ 5-37

Figure 5.11 CsOH Behavior in Steam Generator Secondary ............... 5-38

Figure 5.12 Te Behavior in Steam Generator Secondary ................. 5-39

Figure 5.13 Particulate Behavior in Steam Generator Secondary ........ 5-40

x

LIST OF FIGURES(Conti-nued)

Page

Figure 514 CsI Behavior in the Primary System During InitialPhase of TBA Sequence .............. ...................... 5-45

Figure 5.15 CsOH Behavior in the Primary System During InitialPhase of TBA Sequence ..................................... 5-46

Figure 516 Te Behavior in the Primary System uring InitialPhase of TBA Sequence ..................................... 5-47

Figure 517 Particulate Behavior in the Primary System DuringInitial Phase of TBA Sequence ............................. 5-48

Figure 5.18 CsI Behavior in the Primary System During SecondPhase of TBA Sequence ..................................... 5-51

Figure 519 CsOH Behavior in the Primary System During SecondPhase of TBA Sequence ..................................... 5-52

Figure 520 Te Behavior in the Primary System uring SecondPhase of TBA Sequence ..................................... 5-53

Figure 521 Particulate Behavior in the Primary System DuringSecond Phase of TBA Sequence .............................. 5-54

xi

LIST OF TABLES

Page

Table 2.1 Radionuclide Groups ...................................... 2-5

Table 4.1 Timing of Key Events ..................................... 4-4

Table 4.2 Core and Primary System Response ......................... 4-5

Table 4.3 Containment Response ..................................... 4-8

Table 4.4 Containment Leak Rates ................................... 4-26

Table 4.5 Inventories of Radionuclides and Structural Materials .... 4-92

Table 4.6 Inventory by Group ....................................... 4-93

Table 4.7 Inventory of Melt at the Time of Vessel Failurefor Sequoyah (kg) ........................................ 4-94

Table 4.8 Aerosol Release During Core-Concrete Attack forS3HF1/S3HF2 .............................................. 4-96

Table 4.9 Aerosol Release During Core-Concrete Attack for S3HF3 .... 4-100

Table 4.10 Aerosol Release During Core-Concrete Attack for TB ....... 4-104

Table 4.11 Aerosol Release During Core-Concrete Attack for TBA ...... 4-108

Table 5.1 Masses of Dominant Species Released from Fuel (Total)and Retained on RCS Structures (RET) as a Functionof Time--S3HF Sequence ................................... 5-2

Table 5.2 Masses of Radionuclide Released from Fuel andRetained on RCS (by Group)--S3HF Sequence ................ 5-3

Table 5.3 Summary of Release to Containment for the SUF Sequence.. 5-9

Table 5.4 Size Distribution of Aerosols in Containment--S3HFIScenario ................................................. 5-11

Table 5.5 Fraction of Core Inventory Released fromContainment--S3HF1 Scenario .............................. 5-12

Table 5.6 Distribution of Fission Products by Group--S3HF1Scenario ................................................. 5-13

Table 5.7 Size Distribution of Aerosols in Containment---S3HF2 Scenario ........................................... 5-15

Table 5.8 Fraction of Core Inventory Released fromContainment--S3HF2 Scenario .............................. 5-16

xii

LIST OF TABLES(Continue97

Page

Table 59 Distribution of Fission Products by Group-S3HF2 Scenario ............................................ 5-17

Table 5.10 Size Distribution of Aerosols in Containment--S3HF3 Scenario ............................................ 5-18

Table 5.11 Fraction of Core Inventory Released from Containment-S3HF3 Scenario ............................................ 5-19

Table 512 Distribution of Fission Products by Group-S3HF3 Scenario ............................................ 5-20

Table 513 Masses of Dominant Species Released from Fuel (Total)and Retained on RCS Structures (RET) as a Functionof Time--TB Sequence ................ ..................... 5-22

Table 514 Masses of Radionuclide Released from Fuel and Retainedon RCS (by Group)--TB Sequence ........................... 5-23

Table 5.15 Summary of Release to Containment f the TB Sequence ... 5-24

Table 516 Size Distribution of Aerosols in Containment--TB Scenario .............................................. 5-30

Table 517 Fraction of Core Inventory Released from Containment-TB Scenario .............................................. 5-31

Table 5.18 Distribution of Fission Products by Group-TB Scenario .............................................. 5-32

Table 519 Summary of Primary Coolant System Fission ProductBehavior for TMLU-SGTR ................................... 5-34

Table 520 Time Dependent Fission Product Behavior in SteamGenerator Secondary Side ................................. 5-35

Table 521 Cumulative Fission Product Deposition in SteamGenerator Secondary Side ................................. 5-36

Table 522 Environmental Releases for Tl4LU-SGTR ..................... 5-42

Table 523 Time Dependent and Fission Product Release andDeposition in the Primary System for the InitialPhase of the TBA Sequence ................................ 5-43

Table 524 Cumulative Fission Product Releases for the VariousGroups During Initial Phase of TBA equence .............. 5-44

xiii

LIST OF TABLES(Continued)

Page

Table 525 Time Dependent and Fission Product Release andDeposition in the Primary System for the SecondPhase of the TBA Sequence ................................ 5-49

Table 526 Cumulative Fission Product Releases for the VariousGroups During Second Phase of TBA Sequence ............... 5-50

Table 527 Fission Product Source Terms Released to theContainment for TBA Sequence ............................. 5-56

Table 528 Size Distribution of Aerosols in Containment-TBA Scenario ............................................. 5-57

Table 529 Fraction of Core Inventory Released from Containment-TBA Scenario ............................................. 5-58

Table 530 Distribution of Fission Products by Group-TBA Scenario ............................................. 5-59

Table 531 Distribution of Fission Products by Group-TBAI Scenario ............................................ 5-60

Table 532 Noble Gas and Energy Release to the Environment .......... 5-62

Table 5.33 Icebed Decontamination Factor ............................ 5-64

xiv

REPORT

on

RADIONUCLIDE RELEASE CALCULATIONS FORSELECTED SEVERE ACCIDENT SCENARIOS

Volume IIPWR, Ice Condenser DE-Sign

to

U.S. Nuclear Regulatory Commission*

from

BATTELLEColumbus Division

May 30, 1986

1. INTRODUCTION

This report presents results of analyses of the environmental

releases of fission products (source terms) for evere accident scenarios in a

pressurized water reactor with an ice-condenser ontainment. The analyses

were performed to support the Severe Accident Risk Reduction/Risk Rebaselining

Program (SARRP) which is being undertaken for the U.S. Nuclear Regulatory

Commission by Sandia National Laboratories. In the SARRP program, risk esti-

mates are being generated for a number of reference plant designs. The

Sequoyah Plant has been used in this study as an example of a PWR ice

condenser plant.

All of the analyses in this report have been performed with an0.)interim version of the Source Term Code Package . These results supplement

analyses reported in BMI-2104 Volume IV (2) using essentially the same codes as

in the code package but in their stand-alone forms.

* This work was funded under subcontract to Sandia National Laboratories.

2-1

2. GENERAL APPROACH

The accident scenarios analyzed in this report were selected on the

basis of being significant potential contributors to the risk profile of the

Sequoyah plant. Based on the results of these scenarios, source term bins*

will be developed by Sandia National Laboratories which describe the timing,

quantity, and characteristics of the release of fission products to the

environment.

The methods of analysis used to predict fission product release and

transport behavior are essentially the same as those presented in NUREG-0956,

"Reassessment of the Technical Basis for Estimating Source Terms" (3) . These

computer codes have been assembled as a Source Term Code Package which is

scheduled for public release in the spring of 1986. An interim version of the

code was used in this study.

2.1 Source Term Code Package

A number of changes have been made in the process of integrating the

BMI-2104 source term codes into a Source Term Code Package. Many of these

changes merely simplify the use of the codes by streamlining and automating

the data transfer between codes. Some of the canges, however, involve actual

improvements in the models or in the coupling btween models.

Figure 21 illustrates the manner in hich the codes are grouped in(4) (5) (6)the Source Term Code Package. The MARCH 2 , CORSOR , and CORCON-Mod 2

codes are now coupled. The CORSOR-M version of the CORSOR code, which uses an

Arrhenius form for the empirical correlation, has been incorporated into

MARCH. A consistent treatment can now be made f the release of fission

products and the transport of sources of decay heat from the fuel. Based on

model improvements suggested by ORNL, the release rates of silver and indium

from control rods has been reduced substantially from those in the earlier

version of CORSOR. Similarly, CORCON-Mod 2 is now used in the code package to

Each of the accident scenarios identified by the Accident SequenceEvaluation Program (ASEP) and the Severe Accident Risk Reduction/RiskRebaselining Program (SARRP) is mapped to one of the source term binsin the process of developing the risk profile for the plant.

2 2

INPUTPlant and

Sequence Descriptior

MARCH3(MARCH2, CORCON MD2, CORSOR)

TRAP-MERGEVANESA (TRAP-MELT2, MERGE)

v

NAUA/SPA C/ICEDF

FIGURE 21. SOURCE TERM CODE PACKAGE

2 3

predict the thermal-hydraulic loads on containment due to core-concrete

interactions and as input to the VANESA (7) code to calculate fission product

release. In BMI-2104 these processes were treated in a potentially

inconsistent manner with two different models, INTER (4) and CORCON-Mod 1(8)

Potentially significant changes also resulted from the intimate

coupling of the MERGEM and TRAP-MELT (10) codes in the code package. The

most important of these are listed below and sould be kept in mind when

comparing the present results to results presented in BMI-2104 for equivalent

accident sequences:

o The decay heat contribution to the thermal hydraulicsof the RCS is now considered.

o The fission product transport calculations (TRAP) arenodalized congruently with the thermal hydraulic calcu-lations (MERGE). This includes the use of structuresin control volumes that define the boundaries of con-vective, mixing flow. Previously, distinct structureshad to be nodalized as consecutive control volumes.

o Gas properties used in TRAP are those calculated byMERGE and now account for the presence of hydrogen.

o Heat transfer coefficients used in TRAP are supplied byMERGE; mass transfer is based on those using theChilton-Colburn analogy.

o Aerosol particles are allowed to all back to upstreamvolumes if orientation and geometry permit.

o Aerosol particles settling into the melt are instantan-eously revaporized by species constituents with conden-sed vapors revolatilizing as vapors and particlesregenerating as particles with nucleation size.

o The treatment of chemisorption on walls now accountsfor gas-phase mass transport, which can be limiting forsome flows, especially for the highly reactive Tespecies.

Each of the other codes is run separately in the Code Package. In

general, the interfaces between the codes have been automated so that an

output file from one code is used as the input file for the next.

2-4

2.2 Radionuclide Groups

Initially in the BMI-2104 analyses, four groups of radionuclides

were tracked: iodine, cesium, tellurium, and gross aerosols. In order to

facilitate ex-plant consequence analyses, the groupings were subsequently

changed to the WASH-1400(ii) structure: noble gases, iodine, cesium,

tellurium, strontium, ruthenium, and lanthanum. In both cases the element

named actually represented a group of elements with similar chemical behavior.

For the current study, the NRC recommended that two of the WASH-1400 groups

(strontium and lanthanum) be further subdivided. Table 21 identifies the

radionuclide groups used in this study and the additional elements represented

by each group. Additionally, the inert aerosols generated in-vessel and those

generated ex-vessel are tracked as separate groups. A tracer has also been

used in the NAUA calculations to permit a direct heating source term to be

assessed at a later date if necessary. A massless source of strength unity is

introduced into the containment at the time of vessel failure. The fractional

release to the environment of this simulated source is determined as a

function of time in the same manner as for the different groups of

radionuclides.

2- 5

TABLE 21. RADIONUCLIDE GROUPS

Group Elements

1 Xe Kr

2 I, Br

3 Cs, Rb

4 Te, Sb, SE!

5 Sr

6 Ru, Rh, Pd, Mo, Tc

7 La, Zr, WI, Eu, Nb, Pm, Pr,sm Y

8 Ce, Pu, Np

9 Ba

10 In-vessel aerosols

11 Ex-vessel aerosols

3-1

3. DESCRIPTION OF PLANT AND ACIDENT SCENARIOS

The representation of the Sequoyah plant design in the present

analyses is substantially the same as that used in BMI-2104. A notable change

in the modeling of the plant is the updating of the volumes of the reactor

cavity and the containment floor prior to the overflow of water into the

cavity. Also, the present analyses take into ccount the fact that the bottom

of the reactor vessel would be submerged in water if the reactor cavity is

fully flooded.

3.1 Accident Sequences Considered

The accident sequences selected for ource term analyses for the

Sequoyah Ice Condenser PWR included: several ariations of a pump seal loss-

of-coolant-accident with emergency core coolini and containment spray

recirculation failure (S3HF), station blackout accompanied by a pump seal loss-

of-coolant-accident (TB or SH), an accident-induced steam generator tube

rupture with the TMLU sequence as the starting point (TMLU-SGTR), and a station

blackout accompanied by an accident-induced large break in the primary piping

(TBA). Each of these accident sequences are described more fully below.

The S3HF sequence consists of a pump seal loss-of-coolant-accident

accompanied by the failure of both the emergency core cooling and containment

spray systems in the recirculation mode. Analyses by the Accident Sequence

Evaluation Program (ASEP) have suggested that pump seal failure leak rates may

range from 50 to 500 gallons per minute; the value utilized in the present

analyses was near the top of this range. The dominant mode of failure for

both the emergency core cooling and containment spray recirculation systems has

been indicated to be failure of the valves required for this switchover; thus,

the recirculation systems would fail immediately after the injection phase of

operation. For purposes of defining the timing of refueling water storage tank

depletion, the available engineered safety systems have been assumed to

operate at their full capacities. Within the above framework three variations

of this sequence were evaluated:

- In the first variation, designated S3HF1, both the release from

the reactor vessel at the time of head failure as well as the

3-2

subsequent releases from corium-concrete interactions were

assumed to be scrubbed by the water in the reactor cavity. The

bottom head of the reactor vessel is expected to be submerged in

the cavity water for this sequence.

In the second variation, designated S3HF2, an accident induced

break in the hot leg piping is assumed prior to vessel head

failure; thus, the release from the primary system at the time

of vessel failure would not be scrubbed by the cavity water.

The releases from the corium-concrete interactions would again

be subject to scrubbing by the water in the cavity.

In the third variation of this sequence, S3HF3, only the water

discharged from the accumulators is assumed to be available for

fission product scrubbing. This variation is intended to be a

surrogate for the plugged containment drain scenarios in which

the reactor cavity would be expected to have minimal water.

In the station blackout sequence with early pump seal failure, or

essentially a small-break loss-of-coolant-accident accompanied by station

blackout (SH), the steam driven auxiliary feedwater system is the only active

safety system available. The latter would fail when the station batteries

became depleted, or the steam generators could lose their effectiveness due to

depletion of the primary system inventory through the break. In this case,

the reactor cavity would be expected to be dry except for the discharge of the

accumulators following reactor vessel failure.

The accident-induced steam generator tube rupture scenario assumed

the TMLU sequence as the starting point, with the tube rupture assumed to take

place near the end of the core melting phase of the accident but prior to

reactor vessel failure. The analyses for this sequence were focused on the

releases to the environment through the steam generator and did not address

the fate of the fission products released to the containment.

The TBA sequence is initiated by the complete loss of AC power with

the attendant loss of essentially all the active engineered safety features.

Part of the auxiliary feedwater system supplying water to the secondary side

of the steam generators is steam driven and would continue to operate as long

as DC control power is available. The latter is supplied by the station

batteries which have been estimated to last for five hours under these

3 3

circumstances. After the failure of the batteries and the auxiliary feedwater

system, the steam generators would dry out and boiloff of the primary system

coolant inventory through the pressurizer safety.-relief valves would ensue;

core uncovery and melting would follow. Some time after the start of core

melting it is postulated that a large break in te primary piping takes place;

such a break would be due to accident induced overheating of the hot leg

piping. Depressurization of the primary system hrough the large break leads

to the discharge of the accumulator water onto te partially molten core, with

recovery and quenching of the core. Boiloff of he accumulator water requires

considerable time and leads to a significant depletion of the ice. Eventually

the core remelts, fails the reactor vessel head, and falls into the reactor

cavity. The reactor cavity is expected to contain little or no water in this

scenario, thus concrete attack and associated fission product releases would

take place in a dry cavity.

In addition to the foregoing sequences for which detailed source

term analyses were performed, several other potentially important sequences

were the subject of more limited analyses; these are discussed in the next

chapter.

3.2 Primary System Flowpaths

The flowpaths for fission product transport within the reactor

coolant system for the S3 (seal failure LOCA) scenarios are illustrated

schematically in Figure 31. The TRAP-MERGE control volumes and their

connections used to model these flowpaths are illustrated in Figure 32. Upon

leaving the core region the fission products enter the upper plenum of the

reactor vessel; the latter is represented by a single well-mixed control

volume with four structures within it. The structures modeled include: the

upper core plate, the control rod guide tubes and support columns, the top

support structure, and the core barrel. From the upper plenum the fission

products flow through the hot leg piping to the steam generator, through the

steam generator tubes, and through the crossover pipe to the pump. The

fission products exit the primary system through the failed pump seal.

The primary system flowpaths for the TMLU-SGTR scenario are

illustrated schematically in Figure 33, with the TRAP-MERGE control volumes

I

w

.4

I C;L A

k-

FIGURE 31. PRIMARY SYSTEM FLOWPATH FOR SEQUOYAH SEAL LOCA (S3) SEQUENCE

3-5

4

STEAM GENERATOR

3PIPING

2UPPER PLENUM

BARREL

UPPER SUPPORT PLATE]

GUIDE TUBES

GRID PLATE

CORE

FIGURE 32. SCHEMATIC OF TRAP-MERGE CONTROL VOLUMES FORSEQUOYAH SEAL LOCA IIS3) SEQUENCE

A A

w

1%II

I

FIGURE 33. FISSION PRODUCT FLOWPATHS FOR TMLU-SGTR SEQUENCE

3-7

and their connection illustrated in Figure 34. The flowpaths in this

sequence change during the course of the accident. After leaving the core

region the fission products pass through the upper plenum of the reactor

vessel; the latter is represented as described bove. During the initial core

heatup and melting, prior to the induced steam enerator tube rupture, the

fission products flow from the hot leg piping trough the pressurizer surge

line into the pressurizer, and from there through the relief valve into the

containment. (It is assumed that by the time f fission product release the

pressurizer quench tank rupture disk has failed.) After the occurrence of the

steam generator tube rupture the fission product flowpath changes, going from

the hot leg piping to the steam generator, through the broken tubes into the

steam generator secondary side, and through the secondary side relief valves

to the environment.

The primary system fission product flowpaths during the initial core

melt phase of the TBA sequence are illustrated in Figure 35. The

corresponding TRAP-MERGE control volume breakdown is illustrated in

Figure 36. These flowpaths and their computational representation are

identical to those typically utilized for PWR, I-ML, and TMLB accident

sequences. Figure 37 illustrates the primary ystem fission product flowpath

following the accident-induced break in the primary system; the corresponding

TRAP-MERGE control volume breakdown is illustrated in Figure 38. The

accident-induced opening is assumed to be a larce hot leg opening.

3.3 Containment Flowpaths

The containment flowpaths for the three variations of the SUF

sequence are illustrated in Figures 39 310, and 311. Prior to the time of

reactor vessel failure the behavior is identical in all three variations. The

fission products released from the reactor coolant system are released to the

lower compartment of the ice condenser containment; from there they flow

through the ice condenser into the upper compartment. The air return fans

will transport some of the still airborne activity back down to the lower

compartment, with multiple passes through the ice condenser possible.

The three variations on the SUF sequence differ from one another

after the time of reactor vessel failure. Containment failure at or about the

3-8

9

CONTAINMENT

4

8 DRYERS

DISCHARGE LINE

7 5

PRESSURIZER STEAM GENERATOR 3

-y- SEPARATORS

6 4

SURGE LINE HOT LEG

2

BAFF ES73

HOT LEG 7ET7�]

2 SHELL

BARREL

TUBES

UPPER SUPPORT PLATE]

GUIDE TUBES

GRID PATE 7INPUT

CORE

RUN RUN 2

FIGURE 34. PRIMARY SYSTEM FLOWPATHS FOR STEAM GENERATOR TUBERUPTURE ACCIDENT

IUD [L

FIGURE 35. PRIMARY SYSTEM FLOWATH DURING INITIAL CORE MELT PHASEFOR SEQUOYAH TBA SEQUENCE

3-10

CONTAINMENT

3PIPING

PRESSURIZER

(2 STRUCTURES)

2

UPPER PLENUM

(4 STRUCTURES)

CORE

FIGURE 36. SCHEMATIC OF TRAP-MERGE CONTROLVOLUMES DURING INITIAL CORE MELTPHASE FOR SEQUOYAH TBA SEQUENCE

FIGURE 37. PRIMARY SYSTEM FLOWPATH DURINGLATER CORE MELT PHASE FOR SEQUOYAHTBA SEQUENCE

3-12

CONTAINMENT

12

UPPER PLENUM

(4 STRUCTURES)I I

_ I1

CORE

I

FIGURE 3.8. SCHEMATIC OF TRAP-MERGECONTROL VOLUMES DURINGLATER CORE MELT PHASEFOR SEQUOYAH TBA SEQUENCE

PRIOR TO REACTOR VESSEL FAILURE

PRIMARY LOWER ICE UPPERSYSTEM 110-COMPARTMENT CONDENSER � �bCOMPARTMENT

AIR RETURNFANS

AFTER REACTOR VESSEL AND CONTAINMENT FAILURE

PRIMARYSYSTEM

CAVITY LOWER ICE UPPERWATER COMPARTMENT CONDENSER COMPARTMENT ENVIRONMENT

CORIUM-CONCRETE

INTERACTIOh

FIGURE 39. CONTAINMENT FISSION PRODUCT FLOWPATHS FOR SEQUOYAH S3HF1

PRIOR TO REACTOR VESSEL, FAILURE

PRIMARY LOWER ICE UPPERSYSTEM COMPARTMENT CONDENSER COMPARTMENT

AIR RETURNFANS


PRIMARYSYSTEM

CAVITY LO ICE UPPERWATER 00 COMPARTM CONDENSER COMPARTMENT 10ENVIRONMENT

-T

CORTUM-CONCRETE

INTERACTION

FIGURE 3.10. CONTAINMENT FISSION PRODUCT FLOWPATHS FOR SEQUOYAH S3HF2


PRIMARY LOWER ICE UPPERSYSTEM 10 COMPARTMENT CONDENSER COMPARTMENT

AL

AIR RETURNFANS


CORIUM- LOWER ICE UPPERCONCRETE oCOMPARTMENT P, CONDENSER COMPARTMENT. .pENVIRONMENT

INTERACTION

FIGURE 311. CONTAINMENT FISSION PRODUCT FLOWPATHS FOR SEQUOYAH S3HF3

3-16

time of vessel failure was predicted for all three cases; somewhat different

assumptions were made regarding the scrubbing of fission products by the water

in the reactor cavity. In the first variation or the base case (S3HF1) it

was recognized that the bottom of the reactor vessel would be submerged at the

time of vessel failure, and the fission products still suspended in the

primary system at that time were assumed to be scrubbed by the water in the

reactor cavity. Since the reactor cavity was flooded, the releases from the

corium-concrete interaction were also scrubbed by the water pool. In the

second variation (S3HF2) the puff release at the time of reactor vessel

failure was not subjected to scrubbing by the reactor cavity water; the

subsequent releases from the corium-concrete interaction were, as before,

scrubbed by the overlaying water pool. This variation was intended to

simulate the case of an induced hot leg rupture in the primary system prior to

the occurrence of vessel meltthrough. Whether the difference between the

first and second cases is significant will depend on the quantity of suspended

activity in the primary system at the time of vessel breach. The third

variation (S3HF3) was intended to simulate the plugged drain situation where

the reactor cavity would be dry, except for the water from the accumulators,

at the time of vessel breach. In this case the fission product scrubbing

would be limited to that afforded by the accumulator water before it is boiled

away.

The containment fission product flowpaths for the TB (S3B) sequence

are illustrated in Figure 312. In this case the air return fans would not be

operating due to the loss of electric power. The fission products released

from the primary system would enter the lower compartment of the containment.

From there they would flow through the ice bed and into the upper compartment.

Some recirculation back to the lower compartment may be possible through the

leak paths between the two compartments. Following reactor vessel and

containment failure the airborne activity in the upper compartment would be

available for release to the environment. The reactor cavity in this sequence

would be dry at the time of vessel failure, but would receive the accumulator

discharge. Thus pool scrubbing would be limited to that provided by the

accumulator water.

The containment fission product flowpaths during the several stages

of the TBA accident sequence are illustrated schematically in Figure 3.13.


PRIMARY LOWER ICE UPPERSYSTEM P-COMPARTMENT CONDENSER COMPARTMENT


REACTOR LOWER ICE UPPERCAVITY 10 COMPARTMENT CONDENSER COMPARTME T ENVIRONMENT

FIGURE 312. CONTAINMENT FISSION PRODUCT FLOWPATHS FOR SEQUOYAH TB (S3B)

PRIOR TO REACTOR VESSEL OR CONTAINMENT FAILURE

PRIMARY LOWER ICE UPPERSYSTEM COMPARTMENT CONDENSER 010 COMPARTMENT

AFTER CONTAINMENT FAILURE, BUT PRIOR TO REACTOR VESSEL FAILURE

PRIMARY LOWER ICE UPPER lh�SYSTEM COMPARTMENT CONDENSER COMPARTMENT ENVIRONMENT

AFTER REACTOR VESSEL AND CONTAINMENT FAILURE 00

REACTOR LOWER ICE UPPER ENVIRONMENTCAVITY COMPARTMENT bb- CONDENSER COMPARTMENT mk�

FIGURE 313. CONTAINMENT FISSION PRODUCT FLOWPATHS FOR SEQUOYAH TBA SEQUENCE

3-19

3.4 Containment Failure Mode and Pressure Level

The Sequoyah Unit No. primary containment is a cylindrical steel

shell with a hemispherical dome with a flat bottom. The containment is

surrounded by a reinforced concrete shield structure. The failure pressure

for the structure was assumed to be 65 psia in tese analyses based on a SARRP

evaluation of 65 6 psia. A value of 60 psia hd been assumed in the

BMI-2104 analyses.

4-1

4. BASES FOR TRANSPORT CALCULATIONS

4.1 Phenomenological Modeling Assumptions

The phenomenological modeling assumptions utilized for the present

analyses of the ice condenser PWR design are substantially the same as those

applied in the BMI-2104 analyses. Areas in which the present analyses differ

or involve new approaches are noted below.

In the base case S3HF sequence there is a large amount of water in

the reactor cavity at the time of predicted vessel failure. For the purposes

of the present source term analyses it has been assumed that concrete attack

begins shortly after reactor vessel failure, even with the water in the

cavity. This is equivalent to assuming no substantial fragmentation and

resultant quenching of the core debris upon contact with the water. An

alternate possibility would be debris fragmentation with rapid debris

quenching and the subsequent formation of coolable debris beds. In the latter

case concrete attack could be delayed until the -time that the water in the

cavity is all boiled off. Under the assumptions used here the fission product

releases from the early corium-concrete interactions are subject to scrubbing

by the overlaying water. Under the alternate assumptions noted above corium-

concrete interaction releases would be delayed cnsiderably in time, but would

not benefit from scrubbing by the cavity water. In the latter case it is also

possible that the ice would be melted by the time of the delayed concrete

attack.

In all three variations of the S3HF sequence the air return fans

were assumed to fail at the time of predicted containment failure. While it

does not necessarily follow that containment failure will also fail the fans,

the effectiveness of the fans becomes questionable; e.g., clearly the fans

would not function against the large pressure differentials associated with

depressurization following containment failure. Thus the assumption of air

return fan failure following containment failure is believed to be a

reasonable one and consistent with other assumptions in the analysis.

The steam generator tube rupture scenario considered in the present

analyses is intended to model the situation where the failure of the steam

generator tubes is the result of accident induced thermal loadings. Such a

4-2

postulate would be particularly appropriate if the accident scenario is

characterized by a large degree of steam and hydrogen recirculation within the

primary system as the core uncovers and overheats. Such recirculation

patterns have been postulated to result in substantial redistribution of

energy within the primary system, even to the point where failures of primary

system piping due to overheating have been postulated to take place prior to

core melting. The MARCH code cannot model the effects of the above noted

recirculation flows and their effect on primary system heat transfer. In

order to approximate the consequences of such postulated accident induced

primary system failures it has been assumed that steam generator tube rupture

takes place at the time of the start of core slumping. Core slumping as

treated in the MARCH analyses is typically accompanied by large steam and

hydrogen flows and the heating of structures downstream of the core. The

induced failure was represented by an area equal to five steam generator

tubes, with the primary system allowed to depressurize into the secondary side

of the steam generator; the latter was assumed to be maintained at 1100 psia.

The principal interest in this case involved the release to the environment

through the steam generator. After vessel failure it was assumed that the

secondary side relief valves would close. No subsequent failure of the

containment was considered. Clearly, in some related scenarios the release

through the steam generators would be accompanied by additional release if the

containment were to fail.

4.2 Results of Thermal Hydraulic Analyses

4.2.1 SUF Sequence

As noted previously, three variations of the SUF sequence were

explicitly considered as part of the present analyses. The in-vessel portion

was identical in all three variations, with the differences being in the

assumed behavior subsequent to the predicted time of reactor vessel failure.

In the base case for this sequence, designated as S3HF1, the bottom

of the reactor vessel is submerged by the water in the reactor cavity at the

time of predicted vessel failure. Thus the primary system blowdown was

assumed to go through the water in the reactor cavity. The water in the

4 3

reactor cavity was also assumed to scrub the products of the subsequent

corium-concrete interaction.

The accident event times for the S3HF1 scenario are given in Table

4.1. A summary of core and primary system conditions at key times during the

sequence is given in Table 42. Containment conditions at arious times

during the sequence are summarized in Table 43.

Figure 41 illustrates the primary system pressure, Figure 42 gives

the total water and steam leakage, and Figure 43 shows the primary system

water inventory for the S3HF sequence. The primary system pressure drops

rapidly initially in response to the break, then levels off as the break and

emergency core cooling system flows equilibrate. Failure of the emergency

core cooling system upon switchover to recirculation is followed by another

abrupt decrease in the primary system pressure, with the pressure leveling off

as the system approaches saturated conditions. The abrupt drop in the total

leak rate at about 225 minutes is associated with the change from a liquid to

a steam break, and can also be seen in the slower decrease in the primary

system inventory. The increase in the primary sytem pressure at about 400

minutes is associated with the slumping of the Cre into the vessel head.

Figure 44 illustrates the maximum and average core temperatures

during the in-vessel phase of the accident. The maximum temperature is seen

to arrest at 4130 F, the input effective melting temperature, except for brief

excursions due to rapid metal-water reactions as the molten fuel is relocated

within the core region. Figure 45 illustrates 'the fractions of core melted

and active cladding reacted during the sequence. The extended slow blowdown

in this sequence provides a continuing steam supply for metal-water reactions,

with the total fraction of cladding reacted in this case being higher than is

typically predicted.

Figures 46 and 47 illustrate the containment pressure and

temperature histories for the S3HF1 scenario. The hydrogen igniters are

available in this sequence; thus the timing of -ignition is governed by the

development of a combustible mixture in any of the containment compartments.

Examination of these figures indicates that there are two substantal hydrogen

burns prior to reactor vessel failure, but the brn leading to containment

failure in this case takes place shortly after vssel failure. Contrast of

the pressure and temperature responses illustrates that the burns that are

4-4

TABLE 41. TIMING OF KEY EVENTS

Time,Event minutes

Sequoyah S3HF1

ECCS On 1.0Fan On 10.8Spray On 11.3ECCS Recirculation Failure 36.0Spray Recirculation Failure 42.3Core Uncovery 272.4Start Melt 363.7Core Slump 391.8Core Collapse 393.4Hydrogen Burn 396.4Hydrogen Burn 409.6Bottom Head Failure 410.2Accumulators Empty 410.2Start Concrete Attack 410.2Hydrogen Burn/Containment Failure 412.0Fan Off 412.0Hydrogen Burn 504.7Hydrogen Burn 513.5Hydrogen Burn 521.0Hydrogen Burn 570.1Corium Layers Invert 610.5Ice Melt Complete 991.7Hydrogen Burn 992.7End Calculation 1010.2

Sequoyah S3HF2

ECCS On 1.0Fan On 10.8Spray On 11.3ECCS Recirculation Failure 36.0Spray Recirculation Failure 42.3Core Uncovery 272.4Start Melt 363.7Core Slump 391.8Core Collapse 393.4Hydrogen Burn 396.4Hydrogen Burn 409.9Bottom Head Failure 410.2Accumulators Empty 410.2Start Concrete Attack 410.2Hydrogen Burn/Containment Failure 411.8Fan Off 411.8Hydrogen Burn 504.5Hydrogen Burn 513.0Hydrogen Burn 520.5Hydrogen Burn 574.0Corium Layers Invert 610.2Hydrogen Burn 732.7Ice 1elt Complete 987.0End Calculation 1010.2

4- 4a

TABLE 41. TIMING OF KEY EVENTS(continued)

Time,Event minutes

Sequoyah S3HF3

ECCS On 1.0Fan On 10.8Spray On 11.3ECCS Recirculation Failure 36.0Spray Recirculation Failure 42.3Core Uncovery 272.4Start Melt 363.7Core Slump 391.8Core Collapse 393.4Hydrogen Burn 396.4Hydrogen Burn 409.9Bottom Head Failure/Hydrogen Burn 410.2Accumulators Empty 410.2Start Concrete Attact 410.2Hydrogen Burn/Containment Failure 411.8Fan Off 411.8Hydrogen Burn 546.8Hydrogen Burn 563.2Corium Layers Invert 610.2End Calculation 1010.2

Sequoyah S3B

Core Uncovery 236.6Start Melt 327.1Core Slump 354.4Core Collapse 356.0Hydrogen Burn 373.2Hydrogen Burn/Bottom Head Failure 373.6Hydrogen Burn/Containment Failure 373.6Hydrogen Burn/Start Concrete Attack 373.6Hydrogen Burn 497.6Hydrogen Burn 509.6Corium Layers Invert 528.1End Calculation 973.6

4- 4b


Time,Event minutes

Sequoyah TMLU-SGTR

Fan On 58.2Spray On 58.7Steam Generator Dryout 72.4Spray Recirculation On 91.2Core Uncovery 104.0Start Melt 127.0Steam Generator Tube Rupture 153.0Start Slump 153.9Core Collapse 154.3Vessel Head Dryout 160.5Bottom Head Failure 168.9End Calculation 169.0

Sequoyah TBA

Steam Generator Dry 466.5Core Uncovery 517.8Initial Melt Start 552.5Accumulator/UHI Tank Empty 567.2Primary System Break 572.0Hydrogen Burn 576.1Containment Failure 576.1Final Melt Start 788.9Start Slump 834.9Core Collapse 835.4Ice Melt Complete 848.6Vessel Head Dryout 855.1Vessel Head Failure 985.7Reactor Cavity Dryout 985.8Start Concrete Attack 1003.4Corium Layers Invert 1205.9End Calculation 1603.4

4- 4c


Time,Event Minutes

Sequoyah S3H

Fan On 10.8Spray On 11.3ECCS Recirculation on 36.0Spray Recirculation On 43.0ECCS Recirculation Fails 158.0Core Uncovery 349.1Start Melt 487.6Start Slump 517.6Hydrogen Burn 518.6Core Collapse 519.0Hydrogen Burn 519.2Bottom Head Failure 519.2Accumulators Empty 519.2Start Concrete Attack 519.2Hydrogen Burn 520.6Hydrogen Burn 592.5Hydrogen Burn 644.7Corium Layers Invert 695.7End Calculation 1119.0

Sequoyah S2HF

Fan On 0.5ECCS and Spray On 1.0ECCS Recirculation Failure 22.9Spray Recirculation Failure 29.7Core Uncovery 58.3Start Melt 81.8Hydrogen Burn 89.8Hydrogen Burn 92.0Hydrogen Burn 92.1Start Slump 93.8Hydrogen Burn 94.4Core Collapse 94.6Hydrogen Burn 95.8Vessel Head Dryout. 102.9Accumulators Dryout 119.5Bottom Head Failure 125.0Start Concrete Attack 125.0Hydrogen Burn 145.5Hydrogen Burn 161.5Hydrogen Burn 165.5Hydrogen Burn 168.2Hydrogen Burn 170.7Corium Layers Invert 191.7Ice Melt Complete 196.7Containment Failure 560.0Fan Off 560.0End Calculation 725.0

TABLE 42. CORE AND PRIMARY SYSTEM RESPONSE

PrimaryPrimary SystemSystem Water Average Core Peak Core Fraction Fraction

Accident Time, Pressure, Inventory, Temperature, Temperature, Core CladEvent minutes psia Ibm F F Melted Reacted

Sequoyah S3HF

Core Uncovery 272.4 1200 1.24 x 105 571 574 0.0 0.0Start Melt 363.7 1190 8.87 x 104 1550 4130 0.0 0.06Core Slump 391.8 1684 7.56 x 104 4130 ---- 0.77 0.74Core Collapse 393.4 2035 6.56 x 104 3291 ---- 0.86 0.74Bottom Head Failure 410.2 1993 2.11 x 104 3274 ---- ---- 0.74

Sequoyah S3B

Core Uncovery 236.6 1205 1.24 x 105 572 575 0.0 0.0Start Melt 327.1 1194 8.87 x 104 1575 4130 0.0 0.06Core Slump 354.4 1718 7.13 x 104 4130 ---- 0.75 0.74Core Collapse 356.0 2069 6.53 x 104 3288 ---- 0.86 0.75Bottom Head Failure 373.6 1996 2.11 x 104 3270 ---- ---- 0.75

TABLE 42. CORE AND PRI14ARY SYSTEM RESPONSE(continued)

PrimaryPrimary System

Accident Time, System Water Average Core Peak Core Fraction FractionEvent Minutes Pressure, Inventory, Temperature, Temperature, Core Clad

psia 1 bin F F Melted Reacted

Sequoyah TMLU-SGTR

Core Uncovery 104.0 2373 9.96 x 104 668 681 0.0 0.0

Start Melt 127.0 2371 6.39 x 104 2012 4130 0.0 0.06

Start Slump 153.9 2357 6.06 x 104 4130 -- 0.54 0.31

Core Collapse 154.3 2360 5.80 x 104 3511 0.73 0.42

Bottom Head Dryout 160.5 2368 1.79 x 104 3313 -- 0.43

Bottom Head Failure 168.9 1900 1.58 x 104 3364 0.43

TABLE 42. CORE AND PRIMARY SYSTEM RESPONSE(continued)

PrimaryPrimary SystemSystem Water Average Core Peak Core Fraction Fraction

Time, Pressure, Inventory, Temperature, Temperature, Core CladAccident Event Minutes psia lbm F F Melted Reacted

Sequoyah TBA

Core Uncovery 517.8 2377 1.02 x 105 665 671 0.0 0.0Initial Melt Start 552.5 2377 6.39 x 104 1986 4130 0.0 0.07Final Melt Start 788.9 21 9.26 x 104 2691 4130 0.0 0.58Core Slump 834.9 17 9.08 x 104 4130 -- 0.63 0.68Core Collapse 835.4 26 9.06 x 104 3046 0.69 0.69Vessel Head Dryout 855.1 428 2.73 x 104 1891 -- 0.69Vessel Head Failure 985.7 15 2.20 x 104 3912 0.69

TABLE 43. CONTAINMENT RESPONSE

SteamContainment Condrn�. at ion

Accident Time, Pressure. 1 empe ra tu re, Reactor Cavity Rate,Event minutes psia F Ice Sump Water Wa ter I blininMass, Temp.,Mass, mass, I P111p.

Lower Upper Ibm Ibm F I bm F Lower/Ice/Upper

Sequoyah S311F1

Core Uncovery 272.4 16.9 136 105 1.60 x 106 3.27 x 106 117 4.25 x 105 116 25/437/0Start Melt 363.7 16.9 141 105 1.50 x 106 3.27 x 106 118 6.51 x 105 116 17138110Core Slump 391.8 17.9 174 105 1.44 x 106 3.27 x 106 118 7.18 x 105 116 0/366/0Core Collapse 393.4 18.0 179 105 1.44 x 106 3.27 x 106 118 7.23 x 105 116 0137310Hydrogen Burn 396.4 40.4 1308 828 1.42 x 106 0/0/0Hydrogen Burn 409.6 54.2 1099 7 44 1.34 x 106 0/(/OBottom ead Failure 410.2 38.6 250 589 1.22 x 106 3.27 x 106 119 7.97 x 105 117 7.859/79,370/0Start Concrete Attack 410.2 37.6 247 573 1.22 x 106 3.66 x 106 121 9.22 x 105 130 7,07712,65210Hydrogen/Containment 00Burn / Failure 412.0 62.0 1418 1506 1 16 x 106 3.75 x 106 122 9.21 x 105 140 0/0/0Hydrogen Burn 504.7 16.9 1211 261 1:12 x 106 0/45.590/0Hydrogen Burn 513.5 17.9 1857 253 1 1 x 106 0/29,300/0Hydrogen Burn 521.0 18.2 1813 247 1:09 x 106 0/26,�r0/0Hydrogen Burn 570.1 56.7 2671 785 1.06 x 106 0/0/(Ice Melt Complete 991.7 14.8 206 202 0.0 4.93 x 106 164 8.97 x 105 212 0/1,539/0Hydrogen Burn 992.7 46.8 745 2196 0.0 0/0/0End Calculation 1010.2 14.9 199 144 0.0 4.93 x 106 164 8.97 x 105 209 15/0/161

TABLE 43. CONTAINNENT RESPONSE(continued)

SteamContainment Condpii- at ion

Accident Time, Pressure, Temperature, Reactor Cavity Ra t e,Event minutes psia F Ice Sump Water Water Ib/111-in

Mass, Mass, Temp. Mas lemp.,Lower Upper Ibm Ibm F Ibm F Lowe r/ I c e /Upper

Sequoyah S310`2

Core Uncovery 272.4 16.9 136 105 1.68 x 106 3.27 x 106 117 4.25 x105 116 25/437/0Start Helt 363.7 1619 141 105 1.50 x 106 3.27 x 106 118 6.51 x105 116 17/381/0Core Slump 391.8 17.9 7 4 105 1.44 x 106 3.27 x 106 118 7.18 x105 116 0/366/0Core Collapse 393.4 18.0 179 105 1.44 x 106 3.27 x 106 118 7.23 xIU5 116 0137310Hydrogen Burn 396.4 40.4 1308 828 1.42 x 106 0/0/0Hydrogen urn 409.9 54.2 200 1451 1.27 x 1(6 0/(/OBottom ead Failure 410.2 49.7 260 991 1.17 x 106 3.27 x 106 119 7.97 x105 117 10.020/92,170/0Start Concrete Attack 410.2 47.2 255 922 1.16 x 106 3.72 x 106 120 9.25 x105 121 9,09fl/6.477/0Hydrogen/ContainmentBurn / Failure 411.8 62.0 1252 1551 1.09 x 106 3.82 x 106 121 9.25 x105 123 0/0/0Hydrogen Burn S04.S 17.0 1216 275 1.05 x 1(6 0/35.060/0Hydrogen Burn 513.0 18.0 1953 266 1.03 x 106 0/16,980/0ify'rogen Burn 520.5 N1.3 Iqrh 259 1.02 x 106 0123.21010Hydrogen Burn 574.0 46.0 760 2630 1.00 x 106 0/0/0Hydrogen Burn 732.7 41.7 722 2302 6.26 x 105 0/0/0Ice Melt Complete 987.0 14.8 205 241 0.0 4.93 x 106 163 8.97 105 212 85/108/0End Calculation 1010.2 14.9 208 157 0.0 4.93 x 106 164 8.97 x 105 213 150/0/73

TABLE 43. CONTAINMENT RESPONSE(continued)

stpallContainment Colld(,ii,,at ion

Accident Time. Pressure, Temperature, Reactor Cavity Ra I P,Event minutes psla F Ice Sump Water Water lb/min

Mass, Mass, Temp., Mass, Temp.,Lower Upper Ibm Ibm F Ibm F Lower/Ice/Upper

Sequoyah S311F3

Core Uncovery 272.4 16.9 136 105 1.60 x106 3.70 x 106 117 0.0 ------ 25/4 37 /0Start Melt 363.7 16.9 141 105 1.50 x106 3.92 x 106 118 0.0 ------ 171311110Core Slump 391.8 17.9 17 4 105 1.44 x1(6 3.99 x 1(6 lie 0.0 ------ 0/366/0Core Collapse 393.4 18.0 179 105 1.44 x106 3.99 x 106 118 0.0 ------ 0/311/0Hydrogen Durn 396.4 40.4 1308 828 1.42 x106 0.0 ------ 0/0/0Hydrogen Burii 409.9 54.2 280 1451 1.27 x106 0.0 ------ 0/0/0Bottom Head Failure 410.2 49.7 260 991 1.17 x106 4.07 x 106 119 0.0 ------ 10,020/9?,230/0Start Concrete Attack 410.2 47.2 255 922 1.16 x106 4.33 x 106 120 3.08 x 15 129 9,069/5,937/0 C)

Ilydroge YContainment 1551 1.09 x106 6 120 3.0 x 15 134 0/0/0Burn Failure 411.8 62.0 1253 4.43 x 10Hydrogeti lhoro 546.8 43.2 7 31 2488 1.01 x106 0/(/OIlydrogen Durn 563.2 39.3 717 2428 9.33 x105 0/0/0End Calculation 1010.2 14.8 234 237 8.73 x105 5.75 x 106 158 0 ------ 0/46/0

TABLE 43. r TATNUNENT RESPONSE(continued)

Containment Condmi" t ionAccident Time, Pressure, Tempera tu re, Reactor Cavity Ra te,

Event minutes psla F Ice - -- Sump Water - _ Water lb/minMass, Mass, emp. Mass, I Culp.

Lower Upper Ibm Ibm F I bm F Lower/ice/Upper

Inuoyah S3B

Core Uncovery 236.6 20.3 222 105 2.30 x106 5.29 x105 183 0.0 ------ 301112312Start et 327.1 21.1 233 105 2.26 x106 6.13 x105 Ing 0.0 ------ 265/126/1Core Slump 354.4 21.6 256 105 2.22 x106 6.61 x105 177 0.0 ------ 0/560/1Core Collapse 356.0 21.7 260 105 2.22 x106 6.66 x105 177 0.0 ------ 0/421/1Ilydroqen Burn 373.2 25.3 233 130 2.06 x106 0.0 ------ 1.586/99.070/0Ilydrogpn/BottomBurn / lead Failure 373.6 27.5 293 143 1.89 x106 7.12 x105 175 0.0 ------ 1.891/105,000/0llydroge ContainmentBurn YF&iiure 373.6 65.0 390 1699 1.89 x106 1.07 x106 160 3.1 x 105 115 0/0/0llydroge Start ConcreteBurn YAttack 373.6 $4.4 1239 2196 I.g3 x106 1.14 x106 158 3.1 x 105 115 0/508.(O(/(Hydrogen Burn 497.6 45.3 793 21112 1.64 x106Hydroqen Burn 509.6 38.9 710 2448 1.56 x106 tN/;End Calculation 973.6 14.8 235 226 3.88 x105 2.90 x106 196 ------ 212 01?7810

TABLE 43. CONTAINMENT RESPONSE(continued)

SteamContainment Reactor Cavity Condensation

Temperature, Ice Sump Water Water Rate,Time, Pressure, F Mass, Mass, Temp., Mass, Temp., lb/min

Accident Event minutes Psia Uo-wer Upper lbm Ibm F lbm F Lower Ice Upper

Sequoyah TBA

Steam Generator Dryout 466.5 18.0 179 114 2.45 x106 5.67 x 104 175 0.0 828 0 0

Core Uncovery 517.8 22.0 233 109 1.85 x106 9.50 x 105 150 0.0 717 2008 0

initial Melt Start 552.5 21.1 226 104 1.78 x106 1.06 x 106 150 0.0 406 0 0

Accumulator/UHI Tank 567.2 23.2 273 104 1.38 x106 1.54 x 106 144 0.0 0 0 0Empty

Hydrogen Burn in 576.1 47.1 538 937 1.36 x106 -- -- 0 180,000 0Upper Compartment 4-

Containment Failure 576.1 65.4 1090 1484 1.36 x106 1.56 x 106 145 0.0 0 0 0r-j

Hydrogen Burn in 576.1 78.5 1991 1853 1.34 x106 -- -- 0 200,000 0Lower Compartment

Final Melt Start 788.9 14.7 215 227 9.25 x105 3.12 x106 151 0.0 49 25 0

Start Slump 834.9 14.7 207 228 8.97 x105 3.13 x106 154 0.0 I 0 0

Core Collapse 835.4 14.7 207 228 8.96 x105 3.13 x106 154 0.0 0 0 0

Ice Melt Complete 848.6 14.4 231 227 0.0 3.23 x106 154 0.0 0 0 0

Vessel Head Dyout 855.1 15.8 245 156 0.0 3.23 x106 164 -- 169 0 2401

Vessel Head Failure 985.7 14.8 191 167 0.0 3.23 x106 164 4.3 x 104 157 116 0 0

Reactor Cavity Dryout 985.8 32.8 253 230 0.0 3.23 x106 164 0.0 -- 16,540 0 20,740

Start Concrete Attack 1003.4 14.6 203 199 0.0 3.23 x1(6 166 2.4 x 1(4 165 13 0 0

Corium Layers Invert 1205.9 14.9 271 215 0.0 3.21 x106 180 1.7 x 102 216 0 0 0

End Calculation 1603.4 14.9 251 203 0.0 3.15 x106 189 1.6 x 102 216 0 0 0

SEQUOYAH S3HFI2=-o

-'e4

CL4 2=0

Iwo-o-

low-0-

=.O-

0.0

6.0 50.0 100.0 150.0 200.0 250.0 300.0 3W.0 400.0 450.0 5W.0 5W.0 .0

TI ME (M I NUTE)

FIGURE 41. PRIMARY SYSTEM PRESSURE RESPONSE FOR THE S3HF SEQUENCE

SEQUOYAH S3HF13=-0

3=02

am-0

WW-o

15W.0

E--4 low-0

5W.00

0.0

6.0 MO 00.0 1W.0 W0 00 W0 M0 4W.0 450.0 500.0 5W.0 60TI ME (M I NUTE)

FIGURE 42. PRIMARY SYSTEM COOLANT LEAKAGE FOR THE S3HF SEQUENCE

lb SEQUOYAH S3HF16.0

5.0-

4.0-

z0-4 3.0

zo

0

Ea1.0

0.0W W-0 I&O V60 260.0 2�0.0 ;&.0 W6 460D 4,0.0 560.0 5,0.0 6&.O

TIME (MINUTE)

FIGURE 43. PRIMARY SYSTEM WATER INVENTORY FOR THE S3HF SEQUENCE

SEQUOYAH S3HF15000.0

MAXIMUM........... AVERAGE

4000.0rZ4

3000.0

Ab2000.0

0

1000.0

0.00.0 50.0 1&-0 190.0 260.0 :�0-0 360.0 350.0 400.0 4,40.0 560.0 5W.0 6&.O

TI ME (M I NUTE)

FIGURE 44. MAXIMUM AND AVERAGE CORE TEMPERATURES FOR THE S3HF SEQUENCE

SEQUOYAH S3HF11.0

CLAD UCTED........... CDR. MELTED

Oa

z 0.6

E-4

Pk 0.4-

02

0.06.0 500 100 0 150 20 0 250 300 350 400.0 450 0 500 0 550 0 600 0

TI ME (M I NUTE)

FIGURE 45. FRACTIONS OF CORE MELTED AND CLADDING REACTED FOR THE S3HF SEQUENCE

SEQUOYAH S3HF170.0

60-0

50.0

40.0

PL4

30.0

00

20.0

PL4

�8 ii I III

010.0

0.0

0.0 100.0 260.0 Z&O 400.0 500.0 600.0 700.0 8W.0 900.0 1000.0 U00-0 1M.0TI ME (M I NUTE)

FIGURE 46. CONTAINMENT PRESSURE RESPONSE FOR S3HF1

SEQUOYAH S3HF1ww-o

LOWER........... UPPER

2500.0

P4

2000.0-

1500.0

E-4

1000.0

0 500.0u .............................

.............................. ..... .......

0.0 I I I I I I

0.0 100.0 200.0 300.0 400.0 500.0 600.0 700.0 8W.0 960.0 000.0 UOO. w0TIME (MINUTE)

FIGURE 47. CONTAINMENT TEMPERATURE RESPONSE FOR S3HF1

4-20

largely confined to the lower compartment of the containment lead to

significantly lower overall pressure rises than those that extend into the

upper compartment. It is interesting to note that several burns are predicted

to take place after the containment has failed.

Figures 48 and 49 present the containment sump and reactor cavity

water inventories and temperatures, respectively. From Figure 48 it can be

seen that overflow of sump water into the reactor cavity starts at about 180

minutes. The discharge of the accumulators following reactor vessel failure

results in the filling of the reactor cavity, as well as an increase in the

sump water inventory. The long term increase in the sump water inventory is

associated with continuing melting of the ice. From Figure 49 it can be seen

that both the containment sump and the reactor cavity water are substantially

subcooled at the time of reactor vessel failure. The former remains subcooled

throughout the time considered in the analysis, whereas the latter is heated

to saturation by heating from the core debris and the products of concrete

decomposition.

Figure 410 illustrates the progression of concrete attack as

predicted by CORCON. Initially vertical and radial concrete attack proceed at

about the same rate, but after the debris layers invert with the metal phase

going to the bottom, the predicted attack is primarily vertical.

The ice inventory as a function of time is illustrated in Figure

4.11. The ice is predicted to be depleted more or less at a steady rate

except at the time of reactor vessel failure and the related large combustion

events. Due to the large quantity of water in the reactor cavity which cools

off the products of concrete decomposition, the rate of ice depletion is

relatively constant even after vessel failure.

Figure 412 illustrates the total volume of gases leaked from the

containment as a function of time for the S3HF1 scenario. The initial large

leakage is, of course, associated with containment failure. The increases in

leakage from about 500 to 600 minutes are due to hydrogen burning. The abrupt

increase in total leakage at 610 minutes is the result of the inversion of the

debris layers and an increase in the rate of concrete attack. The rapid

increase near the end of the calculation is a consequence of completion of ice

melting. The time dependent containment leak rates used as input to the

containment fission product transport analyses are summarized in Table 44.

00 SEQUOYAH S3HFl5.0

SUMP........... REACTOR CAVITY

4.0

3,0-

zo

1.0

0.0

0.0 100.0 200.0 30D.0 400.0 500.0 600.0 700.0 800.0 900.0 1000.0 UDO.0 .�O.o

TI ME (M I NUTE)

FIGURE 48. CONTAINMENT SUMP AND REACTOR CAVITY WATER INVENTORIES FOR S3HF1

SEQUOYAH S3HF1250.0

IN SUMP........... IN REACTOR CAVITY

........................

200.0-r=4

150.0

100.0 . ......................

E--4

50.0-

0.06.0 I&O 260.0 300.0 400.0 500.0 600.0 700.0 800.0 900.0 000.0 H.0 m.0

TI ME (M I NUTE)

FIGURE 49. CONTAINMENT SUMP AND REACTOR CAVITY WATER TEMPERATURES FOR S3HF1

SEQUOYAH S3HFI60.0

VEMCAL........... RADIAL

50.0

0 40.0-

049

30.0

20.0

z0

10.0

0.06.0 100.0 200.0 300.0 400.0 560.0 6W.O 760.0 8&o 960.0 060 U;0-0 mm-0

TI ME (M I NUTE)

FIGURE 4.10. PROGRESSION OF CONCRETE ATTACK FOR S3HF1


20.0

u 15.0

z

C-) 10.0 T,P-A P13

0

5.0

0.06.0 I&O 2&.0 360.0 460.0 5(6-0 W60 700-0 800.0 900.0 1000.0 1100.0 1200.0

TIME (MINUTE)

FIGURE 411. ICE INVENTORY FOR S3HF1


c1l)E-4rft 20.0-

pqPA 15.0

010.0

r=40

5.0

0.06.0 100.0 200.0 300.0 460.0 5(6.0 6&0 760.0 80'0-0 960.0 1060 U60.0 1260

TIME (MINUTE)

FIGURE 412. VOLUME OF GASES LEAKED FOR S3HFl

TABLE 44. CONTAINMENT LEAK RATES

Time Leak Lo!�r �qmp�rtment LeAk Uppqr CompaLt!�ent

Interval, Rate.(a) Pressure Temp. Rate, (b) Pressure Temp.Subsequence min Y/hr M a ps'a F C v/hr MPa psla F C Rema rk s

S38 236.6 ----- 0.14 20 222 106 -------- 0.14 20 105 41 Core ncovers236.6 327.1 0.4 0.14 21 228 109 1 x10-4 0.14 21 105 41 Core eats327.1 354.4 1.0 0. 14 21 236 114 1 x10-4 0.14 21 105 41 Core elts354.4 356.2 2.0 0.15 22 259 126 1 x10-4 0.15 22 1 0 5 41 Core siumps and collapses356.2 373.2 5.6 0.15 22 257 125 1 x10-4 0.15 22 109 43 Reactor vessel eats373.2 373.56 203.0 0.10 26 237 114 2 x10-4 0.19 27 137 58 llydrogen brns

373.56 ----- 0.19 20 293 145 2 x10-4 0.19 28 143 62 Bottom head fails373.56-373.58 0.0 0.32 47 273 134 2 x10-4 0.32 47 092 478 Hydrogen burns

373.58 ----- 0.45 65 390 199 2 x10-4 0.45 65 1699 926 Containment fails N3373.50-373.60 954.4 0.57 02 7613 409 25.9 0.57 82 2358 1292 Hydrogen burns ON373.60-433.60 1.3 0.10 15 224 107 0.7 0.10 15 27 135 Concrete decomposition433.60-497.60 2.0 0.10 15 206 97 0.2 0.10 15 245 119 Coiicrete decomposition497.60-497.63 0.0 0.22 32 510 270 19.8 0.22 32 1558 848 Ilydrogen burns497.63-509.60 8.3 0.10 15 214 101 6.4 0.10 15 400 204 Concrete decomposition509.60-509.63 0.0 0.19 28 410 243 18.4 0.19 28 1406 763 hydrogen burns509.63-528.1 6.2 0.10 15 212 100 3.2 0.10 15 421 261 Concrete decomposition520.1 553.6 3.7 0.10 15 209 98 0.2 0.10 15 360 102 Concrete decomposition553.6 673.6 3.8 0.10 15 212 100 0.2 0.10 15 317 158 Concrete decomposition673.6 83.6 3.2 0.10 15 213 101 0.2 0.10 is 263 128 Concrete decomposition853.6 973.6 1.6 0.10 15 230 110 0.2 0.10 15 235 113 Coocrete decomposition

(a) Normalized to a lower compartment-free volume of 3877 x 105 ft3. Units are volume fractions/hour. Leakage is from lower to upper compartment.

(b) Normalized to an upper compartment-free volume of 8979 x 105 ft3. Units are volume fractions/hour. Leakage is from upper compartment to the environment.

TABLE 44. CONTAINMENT LEAK RATES(continued)

Lower Compartment Upper CompartmentTime Leak (a) Leak (b)Interval, Rate, Pressure Tem Rate, Pressurei��. - Temp.Y i� p s i�a - - Ka psia F C RemarksSubsequence min v/hr F C v/hr

S3111`1 272.4 ------ 0.12 17 137 58 -------- 0.12 17 105 40 Corp ucovers272.4 -363.7 13.4 0.12 17 136 50 3 x 10-5 0.12 17 IC15 40 Core 11pats363.7 -391.8 13.9 0.12 17 152 67 3 x 1(-5 0.12 17 105 40 Core meits391.8 -393.5 15.6 0.12 10 170 81 5 x 10-5 0.12 IR 105 40 Core sips aiid collapse%393.5 -396.4 14.6 0.12 18 174 79 5 x 10-5 0.12 le 105 40 Vesel ead hpats396.4 -396.5 353.5 0.20 29 905 485 2 x 10-4 0.20 29 420 215 Hydrogen burns396.5 -409.64 14.4 0.14 21 228 109 1 x 10-4 0.14 21 256 124 Vpssel head eats409.64 -409.69 475.6 0.21 30 662 350 2 x 10-4 0.21 30 401 205 hydrogen brns409.69 -410.2 88.5 0.27 39 309 154 2 x 10-4 0.27 39 662 350 Ve�sel head heats

410.2 ------ 0.27 39 250 121 -------- 0.27 39 589 309 Bottom had fils410.2 -412.0 11.5 0.23 33 230 110 2 x 10-4 0.23 33 474 246 Concrete decomposition412.0 -412.04 604.1 0.31 45 1105 596 2 x 14 0.31 45 865 463 Ilydroqpn burns

412.04 ------ 0.43 62 1418 770 -------- 0.43 62 1506 819 Containment fails412.04 -412.05 0.0 0.45 65 1374 745 21.7 0.45 65 1692 922 Ilydrogen burns412.05 -451.5 1.0 0.10 is 193 89 1.0 0.10 15 298 148 Concrete decomposition41S.5 -504.75 0.0 0.10 15 166 74 0.0 0.10 15 256 125 Concrete decompos i t i nii504.75 -504.78 1175.4 0.11 16 711 377 6.6 0.11 16 255 124 Hydrogen burns$04.78 -513.50 0.4 0.10 15 193 89 0.8 0.10 15 233 112 Cniirrete decomposition513.50 -513.52 2198.5 0.12 17 983 528 8.5 0.12 17 246 119 hydrogen burtis513.52 -521.0 0.3 0.10 15 204 95 0.6 0.10 15 223 106 Concrete decomposition521.0 -521.02 2178.7 0.12 37 964 51S 8.9 0.12 17 239 115 Hydrogen burns521.02 -570.12 1.6 0.10 15 180 02 0.5 0.10 15 225 107 Cotirrptp derniujin% i f i on570.12 -570.16 0.0 0.23 33 515 268 19.2 0.23 33 1497 814 hydrogen burns570.16 610.5 4.8 0.10 is 202 94 1.0 0.10 15 335 168 Concrete decomposition610.5 -662.7 3.5 0.10 15 203 95 0.2 0.10 15 291 138 Concrete dvcnmpnsiHoii662.7 -752.7 3.3 0.10 15 202 94 0.2 0.10 15 254 123 Concrete decompoOtion752.7 -842.2 2.9 0.10 15 205 96 0.1 0.10 is 232 HI Coiicrete decomposition842.2 -936.0 3.6 0.10 15 206 97 0.1 0.10 15 216 102 Concrete dpcompos i t i on936.0 -991.7 3.8 0.10 15 206 97 0.1 0.10 15 204 95 Concrete decomposition991.7 -992.681 23.1 0.10 15 208 98 1.8 0.10 15 123 51 Concrete decomposition992.6RI- (92.684 0.0 0.21 31 467 424 17.4 I 1j,'i 616 Hydrogen burns992.684-1010.2 3.9 0.10 15 200 98 2.1 0. 10 15 144 62 Concrete decomposition

(a) Normalized to a compartment-free volume of 3877 x oS ft3. Units are volume fractions/hour. Leakage is from lower to pper rompartment.

W Normalized to a compartment-free volume of 8979 x 105 ft3. Units are vume fractions/hour. Leakage is from pper rompa r I moi to t rnv I rnivir I


Lower Compartment ljpq2_r qmpartme�tT i rite Leak Leak (b)

Interval, Rate.(a) Pressure Terlip. Rate, Pressure Temp.Subsequence min V/hr Pit ___F a F C v/hr 14fla -- p-SIT V_ C Remarks

S*1172 272.4 ------ 0.12 17 137 58 -------- 0.12 17 105 40 Core uncovers272.4 - 363.7 13.4 0.12 17 136 so 3 x jo-5 0.12 17 105 40 Core heats363.7 - 391.8 13.9 0.12 17 152 67 3 x 10-5 0.12 17 105 40 Corp melts391.8 - 393.5 15.6 0.12 is 178 81 5 x 10-5 0.12 la 105 41 Core smp ad collapse.-;393.5 - 396.4 14.6 0.12 in 17 4 79 5 x 10-5 0.12 18 105 41 Vr,,;spl head heats396.4 - 396.5 353.5 0.20 29 905 485 2 x 1(-4 O.?O 29 420 215 Hydrogen burns396.5 - 409.90 19.4' 0.14 21 227 100 I x 10-4 0.14 21 254 124 VP,;wl hd eats409.90- 409.94 0.0 O.Z8 41 244 Ha 2 x 10-4 0.28 41 859 459 Ilydrogen burns409.94- 410.22 133.0 0.36 52 259 126 2 x 10-4 0.36 52 1176 635 Vespl ead heats

410.22 ------ 0.32 47 255 124 -------- 0.32 47 922 494 Bottom ead fails410.22- 411.69 14.3 0.27 39 238 114 2 x 10-4 0.27 39 702 372 Concrete decomposition411.69- 411.75 481.7 0.32 46 1024 551 2 x 10-4 0.32 46 919 493 Hydrogen brns

411.75 ------ 0.43 62 1252 678 -------- 0.43 62 1551 844 Containment fails 421.0.9 158 Concrete decomposition411.75- 451.25 0.10 15 183 84 1.0 0.10 15 316451.25- 501.25 0.0 0.10 15 161 72 0.0 0.10 15 275 135 Cniicrete decomposition 00501.25- 504.50 0.1 0.10 15 156 69 0.0 0.10 15 269 132 Concrete decomposition504.50- 504.53 1142.3 0.11 16 683 362 6.7 0.10 16 270 132 Ilydrogen burns504.53- 513.0 0.3 0.10 15 190 88 0.8 0.10 15 24 7 119 Covicrpte decomposition513.0 - 513.02 2120.0 0.12 17 1017 54 7 8.8 0.12 17 260 126 Ilydrogen burns513.02- 520.50 0.3 0.10 15 201 94 0.4 0.10 is 235 113 Cotic re te decompos i t I on520.50- 520.52 2098.4 0.12 17 981 528 9.0 0.12 17 252 122 Hydrogen burns520.52- 574.0 1.0 0.10 15 171 7 7 0.4 0.10 15 239 115 Concrete decomposition574.0 - 574.04 0.0 0.22 32 494 256 19.0 0.22 32 1466 796 Hydroqen burns574.04- 610.25 4.1 0.10 is 198 92 0.7 0.10 15 346 17 4 Concrete decomposition610.25- 732.74 3.3 0.10 15 202 94 0.2 0.10 15 274 134 Concrete decompo s I t i n734.74- 732.75 0.0 0.20 29 473 245 18.0 0.20 29 1300 704 Hydrogen burns732.75- 804.25 3.1 0.10 15 203 95 0.7 0.10 15 323 162 Concrete decompo I tioii804.25- 894.6 2.9 0.10 15 205 96 0.1 0.10 15 275 135 Cciiicrete decompo q f t f on894.6 - 987.0 3.2 0.10 15 209 98 0.1 0.10 15 251 122 Concrete dec oinpo s i t i on987.0 1010.2 4.5 0.10 15 206 97 2.0 0.10 15 159 71 Concrete decomposition

(a) Normalized to a compartment-free volume of 3877 x 105 ft3. Units are volume fractions/hour. Leakage is from lower to topper compartment.

W Normalized to a compartment-free volume of 8979 x 105 rt3. Units are volume fractions/hour. Leakage is from tipper compartment to the envfromn�nf.


Lower Compartment Upper CompartmentTime Leak Leak (b)

Interval. Rate, (a) Pressure Temp. Rate, -- Pressure Temp.Subsequence m1n v/hr MN psia F C v/hr mpa psia F C Remarks

S3111`3 272.4 ----- 0.12 17 1 37 58 -------- 0.12 17 105 40 Core wicovprs272.4 - 363.7 13.4 0.12 17 136 50 3 x 10-5 0.12 17 105 40 Core ets363.7 - 391.8 13.9 0.12 17 152 67 3 x 10-5 0.12 17 105 40 Core elts391.8 - 393.5 15.6 0.12 18 178 el 5 x 10-5 0.12 is 105 40 Core slumps and collapses393.5 - 396.4 14.6 0.12 le 174 79 5 x 10-5 0.12 18 105 40 Ves-el head hts396.4 - 396.5 353.7 0.20 29 905 485 2 x 10-4 0.20 29 420 215 Hydrotiell hIII-lis396.5 - 409.90 19.4 0.14 21 227 IU0 I x 10-4 0.14 21 254 124 VP-,%Pi hoad eats409.90- 409.94 0.0 0.28 41 244 118 2 x 10-4 0.28 41 859 459 Ilydro9pn brns409.94- 410.22 133.0 0.36 52 259 126 2 x 10-4 0.36 52 1176 635 Vessel ead heats

410.22 ------ 0.34 50 260 127 -------- 0.34 50 991 533 11pad fails/11ydroqrn hiorfis410.22- 411.69 14.4 0.77 39 230 114 2 x 10-4 0.27 39 702 37 2 rolirretp decomposition411.69- 411.75 403.0 0.32 46 1024 551 2 x 10-4 0.32 46 918 492 Ilydrogell brlis

411.75 ------ 0.43 62 1253 678 -------- 0.43 62 1551 844 Contaimnetit fails411.75- 451.25 0.8 0.10 15 183 84 1.0 0.10 15 316 158 Concrete decomposition451.25- 501.25 0.4 0.10 15 176 80 0.1 0.10 15 274 134 Concrete decomposition501.25- 546.75 1.4 0.10 15 189 87 0.3 0.10 15 260 127 Concrele derompos i t i nii546.75- 546.78 0.0 0.21 31 491 255 19.1 0.21 31 1451 780 Hydroqvii but It,546.70- 563.25 6.9 0.10 15 204 96 2.2 0.10 15 394 201 Coti r I derninpos f L i oti563.25- 563.28 0.0 0.20 29 492 256 18.7 0.20 29 1466 7 97 Ilydrnrinii hurw;563.10- 610.25 4.9 n t. 1. �umr, pi deroo.por I t r..2007 97 00.5 0.10 All 4 0 'IOS610.25- 696.95 4.6 0.10 '5 20� 96 0.2 0.10 15 330 166 Coiict-(,I.(- dernmpos i t It.696.95- 789.5 3.6 0.10 15 208 98 0.1 0.10 15 293 14 Concrete derompos I in709.5 - 891.8 2.2 0.10 15 214 101 0.1 0.10 15 267 131 Concrete decninpo% i t iciii891.8 1010.2 0.5 0.10 15 230 110 0.1 0.10 15 240 120 Coiicrete decompos i L ioii

(a) Normalized to a compartment-free volume of 31077 15 ft3. Units are volume fractions/hour. Leakage is front ower to tipper comparlinont.

(h) Normalized to a compartment-free volume of 8979 x 105 ft3. Uflts'are volume ractions per hour. Leakage is fom tipper compartment Lo the environvinnt.


Lower Compartment Upp(��- C2�!pLyjt �L� tTime Leak Leak

Interval Rate,(a) Pressure Temp. Rate,(b) Pressure ___Iemp -W - F C RemarksSubs eqtience min v/hr MPa sia F C v/hr a psia

TBA 0.0- 466.5 0.1 0.10 15 104 40 6x 10-5 0.10 15 101 38 Dryout of Steam Generators466.5- 517.8 7.1 0.14 21 223 106 1x 10-4 0.14 21 117 47 Core Heats517.8- 552.5 1.2 0.15 22 231 III Ix 10-4 0.15 22 106 41 Core Uncovers552.5- 567.2 19.8 0.15 22 279 138 1x 10-4 0.15 22 103 40 Initial Melting567.2- 576.1 0.7 0.17 24 273 134 1x 10-4 0.17 24 109 43 Accumulators Empty, Core Quenched576.1- 576.13 200.6 0.30 43 537 280 2x 10-4 0.30 43 784 418 Hydrogen Burns

576.13 -- 0.45 65 1090 588 -- 0.45 65 1484 807 Containment Fails576.13-576.15 614.2 0.52 76 1630 888 22.7 0.52 76 1818 992 Hydrogen Burns576.15-672.9 8.4 0.10 15 252 122 0.4 0.10 15 245 118 Core Reheats LO672.9- 788.9 2.6 0.10 15 224 106 0.0 0.10 15 227 108 Core Reheats co788.9- 834.9 0.1 0.10 15 210 99 0.0 0.10 15 227 108 Core Melting Resumes834.9- 835.7 1.1 0.10 15 209 98 0.0 0.10 15 228 109 Core Slumps and Collapses835.7- 848.6 7.7 0.10 15 235 113 0.4 0.10 15 227 108 Ice Melt Complete848.6- 855.1 17.1 0.10 15 236 113 3.5 0.10 15 136 58 Dryout at Vessel Head855.1- 985.7 0.7 0.10 15 208 98 0.2 0.10 15 182 83 Vessel ead [feats

985.7 -- 0.10 15 191 88 -- 0.10 15 167 75 Vessel Head Fails985.7- 985.8 1654.2 0.17 24 231 110 11.8 0.17 24 205 96 Dryout of Reactor Cavity985.8-1063.4 3.4 0.10 15 240 116 1.7 0.10 15 205 96 Concrete Decomposition

1063.4-1183.4 4.5 0.10 15 292 145 1.2 0.10 15 211 99 Concrete Decomposition1183.4-1363.4 1.3 0.10 15 271 132 0.5 0.10 15 213 101 Concrete Decomposition1363.4-1543.4 1.1 0.10 15 253 123 0.4 0.10 15 207 97 Concrete Decomposition1543.4-1603.4 1.1 0.10 15 252 122 0.4 0.10 15 203 95 Concrete Decomposition

(a) Normalized to a compartment free volume of 3887 x 105 ft3. Units are volume fractions/hour. Leakage is from lower to uppercompartment.

(b) Normalized to a compartment free volume of 8979 x 105 ft3. Units are volume fractions/hour. Leakage is from uppercompartment to the environment.

4 30

In the S3HF2 scenario the blowdown from the primary system at the

predicted time of reactor vessel failure was not assumed to go through the

water in the reactor cavity. This can be taken as a simulation of the case of

an induced hot leg rupture prior to vessel head failure. The analysis of the

subsequent concrete attack did take into account the water in the reactor

cavity.


temperature histories for the S3HF2 scenario. As in the preceding case, there

are two hydrogen burns prior to reactor vessel failure but containment failure

is predicted shortly after reactor vessel failure. Because of the different

treatment of primary system blowdown, the details of the hydrogen combustion

events in this case differ from those in the other variations of this

sequence.

Figures 415 and 416 illustrate the containment sump and reactor

cavity water inventories and temperatures, respectively. The behavior in

these is substantially the same as in the preceding case.

Figure 417 illustrates the progression of concrete attack for the

S3HF2 scenario. The mass of ice in the ice condenser is illustrated in

Figure 418. The behavior in these is substantially the same as in the

preceding scenario. The total volume of gases leaked from the containment as

a function of time for this scenario is shown in Figure 419. The differences

between this and the preceding variation of this sequence are due to

differences in the detailed timing of hydrogen burn events. The time

dependent containment leak rates used as the basis for containment fission

product transport analyses are again given in Table 44.

In the variation of the S3HF sequence designated as S3HF3 the

reactor cavity was not allowed to fill with water; this is intended as a

surrogate for sequences in which the plugging of the drains between the upper

and lower compartments results in all the sump water being pumped into the

upper compartment, thus leading to the failure of the recirculation systems.


temperature histories for the S3HF3 scenario. As in the previous variations,

containment failure is predicted shortly after the time of reactor vessel

failure. The predicted course of hydrogen burn events is somewhat different

here from the other variations, as would be expected.

SEQUOYAH S3HF270.0

60.0

50.0

40.0

30.0

20.0

0 10.0

0.0 6.0 i00.0 200.0 300.0 400.0 500.0 600.0 700.0 &6 G&O 10000 U;oo 12�00

TI ME (M I NUTE)


SEQUOYAH S3HF23=- -


r=4

2500.0

2000.0W04

E-4 1500.0

E-4z

1000.0

laL4

�40 500.0-

........... I................. ...............................

.... ......................................

0.0

0.0 100.0 260.0 300.0 4�0-0 560.0 6W.O 7W.0 8M.0 900.0 1000.0 U M0TIME - (MINUTE)

FIGURE 414. CONTAINMENT TEMPERATURE RESPONSES FOR S3HF2


SUMP........... REACTOR CNI TY

4.0

3,0

zoE-4

1.0

0.0 --6.0 ido-0 260.0 300 400 500 0 600 7(� 0 1� 0 G&O 106-0 U; 0 1".0

TI ME (M I NUTE)


SEQUOYAH S3HF2250.0


.....................

2W.0

150.0

E-q 100.0 ......................

50.0

0.0 6.0 100.0 200.0 300.0 400.0 500.0 6W.0 7M.0 W0.0 9W.0 1000.0 U00- W0

TI ME (M I NUTE)


SEQUOYAH S3HF260.0

VERMAL........... PADIAL

50.0

Z.01--4 40.0E-4

a

ra30.0

W-0................................................

z0

10.0

0.00.0 100.0 2oo-o 3oo-o 4ooo 5ooo emo 7(6 86o.o 90'0.0 00 U00.0 M0

TIME (MINUTE)

FIGURE 417. PROGRESSION OF CONCRETE ATTACK FOR S3HF2

SEQUOYAH S3HF2'b 25.0

20.0-

WU 15.00-4

z1-4

Da

IL) 10.0

N40

5.0

0.00.0 100.0 260.0 360.0 4(6 580-0 6(6 760.0 f&o 900.0 060 ffoo.o two.0

TIME (MINUTE)



ce)E-4 25.0

20.0

cn 15.0

010.0

pa

5.0

0.00.0 100.0 200.0 300.0 40D.0 500.0 6(6-0 760.0 1�0-0 900.0 000.0 Hoo-0 "0

TIME (MINUTE)

FIGURE 419. TOTAL VOLUME OF GASES LEAKED FOR S3HF2

SEQUOYAH S3HF370.0

--194 60.01-4

En04

A4 50.0P4

40.0

30.0

co

20.0

PLO

010.0

0.06.0 I&O 260.0 :360.0 4W.o 5W.0 6W.0 7W.0 800.0 9W.0 1000.0 UOO-O IM-0

TIME (MINUTE)


SEQUOYAH S3HF325M.0-


rX4

P4 2000.0-

1500.0

E-41000.0

E-4

500.0-

................. . .........................................

......................................

0.0 I I i0.0 100.0 300.0 4&.0 tr�oo ew,.o -,�Oo 860.0 960.0 060 U;0.0 1260

TI ME (M I NUTE)

FIGURE 421. CONTAINMENT TEMPERATURE RESPONSES FOR S3HF3

4-40

Figures 422 and 423 give the containment sump and reactor cavity

water inventories and temperatures. The water in the reactor cavity in this

case is limited to that due to accumulator discharge, and is substantially

less than in the other variations. The water in the reactor cavity is

predicted to be boiled off during the course of the sequence. As in the other

variations, no fragmentation of the core debris was assumed in the analysis.

More rapid depletion of the water in the reactor cavity would be predicted if

debris fragmentation were assumed; debris quenching associated with such

fragmentation could delay the onset of concrete attack.

Figure 424 illustrates the progression of concrete attack for the

S3HF3 scenario. The behavior through the inversion of the debris layers is

very similar to that predicted for the other variations. It is interesting to

note, however, that the progression of radial attack is resumed in this case

after the water in the reactor cavity has been boiled off.

Figure 425 gives the ice condenser inventory for this scenario.

The behavior is similar to that of the other cases, except that the rate of

ice depletion is seen to decrease after the evaporation of cavity water.

The total volume of gases leaked is illustrated in Figure 426. The

overall behavior is similar to the other variations of this sequence, with

differences attributable to variations in the details of the hydrogen burn

events. The time dependent containment leak rates used as the basis for

containment fission product transport analyses are given in Table 44.

4.2.2 TB Sequence

The TB sequence consists of station blackout as the initiating event

accompanied by pump seal failure. Thus none of the active engineered safety

features, with the exception of the steam turbine driven auxiliary feedwater

systems, are available. It is essentially a small-small break loss of coolant

accident accompanied by loss of all electric power; it can also be designated

as S3B.

The predicted accident event times for the TB (S3B) sequence are

given in Table 41. The core and primary system conditions at key times

during the sequence are summarized in Table 42. Containment conditions at

various times during the sequence are summarized in Table 43.

00 SEQUOYAH S3HF36.0

SUMP........... MCTOR cnri

5.0

4.0

ao-

zo

1.0

.............. .................................

0.0

1&.O 20.0 0�0-0 40.0 5 W1.0 mo -&o a&o 6oo io6.o ;o.o 126-oTI ME - M I NUTE)


SEQUOYAH S3HF3250.0


........... ........ .......... ................... . .....................

200.0-

E-4 150.0

E-4 100.0 . .......................................................

50.0

0.00.0 100.0 200.0 300.0 400.0 500.0 600.0 760.0 E&O 900.0 1000.0 U0O.O 1200.0

TI ME (M I NUTE)


SEQUOYAH S3HF360.0

VEMCAL........... MIAL

50.0

040.0

30.0

20.0 ................................

10.0

0.00.0 100.0 200.0 300.0 4W.0 5W.0 6&0 760.0 1960.0 G&O 1060 U;0-0 mo-o

TIME (MINUTE)

FIGURE 424. PROGRESSION OF CONCRETE ATTACK FOR S3HF3


20.0-

0 15.0-0-4

z

Al.10.0-

0

5.0

0.06.0 1&.0 20.0 00 400.0 9ZO-0 01 0.0 700.0 800.0 900.0 1000.0 U M0

TI ME (M I NUTE)



Cl)

V 20.0

El

15.0

10.0

0

5.00

0.00.0 100.0 200.0 300.0 400.0 500.0 600.0 700.0 OW-0 9W.0 1000.0 U00- L-0

TI ME (M I NUTE)

FIGURE 426. TOTAL VOLUME OF GASES LEAKED FOR S3HF3

4-46

Primary system pressures, leak rates, and water inventory for the TB

sequence are illustrated in Figures 427 428, and 429, respectively. The

primary system pressure decreases rapidly until saturated conditions are

reached. Subsequently, the pressure stays essentially constant as the primary

system water inventory steadily declines. When the break flow changes from

liquid to steam there is an abrupt change in the leak rate, which declines

further as the core becomes uncovered. Primary system repressurization near

the end of the in-vessel phase of the accident is associated with the slumping

of the core into the vessel bottom head.

Figure 430 illustrates the maximum and average core temperatures

for the TB sequence. As long as the core is covered it is well cooled and its

temperatures differ only slightly from that of the water coolant. After core

uncovery the temperatures increase in response to continued decay heating and

the energy input from cladding oxidation. The maximum core temperature is

seen to arrest at the input effective melting temperature, except for some

brief excursions due to rapid metal-water reactions. The average core

temperature is seen to increase monotonicaly after core uncovery up to the

time of core collapse into the vessel bottom head.

The fractions of core melted and active cladding reacted are

illustrated in Figure 431. The extended slow primary system depressurization

associated with pump seal failure provides a continuing supply of steam for

cladding oxidation; thus the extent of cladding reacted in this case is

somewhat higher than typically predicted.

The containment pressure and temperature responses for the TB

sequence are illustrated in Figures 432 and 433. Containment failure was

predicted shortly after vessel head failure and was due to the pressure rise

from hydrogen burning. In this sequence the hydrogen igniters were not

available due to the loss of electric power, and hydrogen ignition was assumed

to take place after vessel head failure with the hot debris acting as the

ignition source. From Figure 433 it can be seen that hydrogen burning is

predicted to take place largely in the upper compartment of the containment.

The same is also true for the burns that are predicted to take place after

containment failure.

SEQUOYAH S3B

P4 2000.0

P4P4

1500.0-

1000.0

T,1>4

5W.0

P4P-4

0.0

0.0 56 IC6.0 1!6.0 200.0 360.0 350.0 4W.0

TI ME (M I NUTE)

FIGURE 427, PRIMARY SYSTEM PRESSURE RESPONSE FOR S3B

SEQUOYAH S3B3500.0

3000.0

2500.0-

W 2000.014

::4

1500.0

4-00

P4E-4 1000.0

500.0-0E--4

0.0 6.0 56.0 160.0 190.0 2�0-0 2W.0 :3;0.0 390.0 4&0

TI ME (M I NUTE)

FIGURE 428. PRIMARY SYSTEM LEAK RATES FOR B

00 SEQUOYAH S3B6.0

$4 5.0

4.0

ao

zo-

cri1.0

0.00.0 16 160.0 150.0 250.0 -&o M6 4W.0

TI ME (M I NUTE)

FIGURE 429. PRIMARY SYSTEM WATER INVENTORY FOR S3B

SEQUOYAH S3B5000.0

m"imm........... AVERAGE

4000.0-

3000.0

E-4 2000.0 4b-

1000.0

0.0 6-0 56.0 100.0 150.0 200.0 250.0 300.0 350.0 400.0TI ME (M I NUTE)

FIGURE 430. MAXIMUM AND AVERAGE CORE TEMPERATURES FOR S3B

SEQUOYAH S3B1.0

CLAD REACTED..... ODRE MELTED

O's

z 0.60

44 0.4

0.0

56.0 100.0 150.0 200.0 250.0 3&.0 &W-0 460.0

TI ME (M I NUTE)

FIGURE 431. FRACTIONS OF CLAD REACTED AND CORE MELTED FOR S3B

SEQUOYAH S3B90.0

-94 80.0

70.0

60.0-

50.0-

40.0-

30.0-

20.0-

10.00.0 100.0 200.0 360.0 400.0 500.0 6W.0 700.0 800.0 900.0 1000.0

TIME (M I NUTE)

FIGURE 432. CONTAINMENT PRESSURE RESPONSE FOR H

SEQUOYAH S3B3000.0-,


r=4

2500.0-

2000.0-

E--4 1500.0-

E-4

1000.0-

P-4

�40

0 500-0-

.............. ..........................

..........

................................................

0.0 I I I

0.0 100.0 200.0 300.0 400.0 500.0 6W.0 700.0 800.0 qw.0 1060

TI ME - M I NUTE)

FIGURE 433. CONTAINMENT TEMPERATURE RESPONSES FOR S3B

4 54

Figures 434 and 435 illustrate the containment sump and reactor

cavity water inventories and temperatures, respectively. In this sequence the

water in the reactor cavity is limited to that from the accumulator discharge.

The water in the containment sump comes from the primary system inventory and

the melting of the ice. As can be seen from Figure 434, it takes a long time

for the water in the cavity to be boiled off since in this, as in the other

cases considered, no fragmentation of the core debris was assumed. With

fragmentation of the core debris assumed, the water in the cavity would have

been evaporated much more rapidly, but the start of concrete attack would have

been delayed.

The predicted progression of concrete attack is illustrated in

Figure 436. Initially approximately equal rates of concrete attack in the

vertical and radial directions are predicted. After the debris layers are

predicted to invert, with the metal phase going to the bottom, the attack is

predominantly in the vertical direction. The radial attack is predicted to

resume after the dryout of the reactor cavity.

The inventory of ice is illustrated in Figure 437. Ice depletion

is relatively slow in this case up to the time of reactor vessel failure, and

is all due to primary system blowdown. The rapid decrease in the ice

inventory at the time of vessel failure is due to the release of high pressure

steam from the primary system as well as due to containment depressurization

following failure.

The total volume of gases leaked from the containment is illustrated

in Figure 438. The initial large leakage is due to containment failure; the

subsequent abrupt increases are due to hydrogen burns. The longer gradual

increase in the total volume of gases leaked is due to corium-concrete

interactions. The time dependent containment leak rates used as input for the

containment fission product transport analyses are given in Table 44.

4.3 TMLU-SGTR Sequence

The TMLU sequence was selected as the basis for illustrating the

accident source terms that may be associated with accident induced steam

generator tube ruptures. The interest in this analysis is in the behavior of

the reactor primary and steam generator secondary systems; the containment

lb SEQUOYAH S3B30.0

SUMP........... REACTOR CRITY

25.0

20.0

15.0

10.0 qLn

5.0

..................

0.0 ----0.0 1(6.0 2(6.0 3(60 4&0 5(6.0 6(6.0 7&0 860.0 900.0 1060

TI ME (M I NUTE)

FIGURE 434. CONTAINMENT SUMP AND REACTOR CAVITY WATER INVENTORIES FOR S3B

SEQUOYAH S3Bmo-0


........... .......................................... ....... ...............................

200.0rZ4

150.0

100.0 ...........................................................

50.0

0.0 6.0 160.0 2&0 360.0 460.0 5&O M.0 7�0-0 1�0-0 G&O 1060TI ME (M I NUTE)

FIGURE 435. CONTAINMENT SUMP AND REACTOR CAVITY WATER TEMPERATURES FOR S3B

SEQUOYAH S3B70.0

VERMAL,........... RADIAL

60.0

50.0

40.0

30.0tb

20.0

z0

10.0 .. ................................................

0.00.0 100.0 200.0 300.0 400.0 5&0 6C� 0 7(6 8600 9000 lowo

TI ME (M I NUTE)

FIGURE 436. PROGRESSION OF CONCRETE ATTACK FOR H

"b SEQUOYAH SH25.0

20.0-

15.0-

z0-4

WI-) 10.0 4�b

00

0

5.0

0.06.0 1&.0 260.0 3(6 4(6 560D 600.0 700.0 800.0 900.0 1000.0

TI ME (M I NUTE)

FIGURE 437. ICE INVENTORY FOR S3B

lb SEQUOYAH S3B25.0

E--4r=4 2D.0

ciDa

15.0

10.0rX4 OLn

0 9.0

9

�l 5.00

0.06.0 160.0 1260.0 360.0 4&.0 560.0 660.0 700.0 8000 900.0 1000.0

TI ME (M I NUTE)

FIGURE 438. TOTAL VOLUME OF GASES LEAKED FOR S3B

4-60

behavior will not be addressed here. In a TMLU type of sequence all makeup to

the primary and secondary systems is lost and continued decay heating boils

off first the steam generator secondary, and then, the primary coolant

inventory. The primary coolant boiloff takes place through the cycling

pressurizer relief valve; thus core overheating and melting take place at an

elevated pressure. As noted previously, for the present purposes it was

assumed that the events associated with core slumping would lead to the

rupture of the steam generator tubes, with the release of steam, hydrogen, as

well as fission products to the now dry secondary side of the steam generator.

The secondary side of the steam generator was assumed to be maintained at

1100 psia by the operation of the atmospheric dump valves. With the failure

of the reactor vessel head the flow through the steam generators was assumed

to cease and the present analysis was terminated at that point. It can, of

course, be postulated that the steam generator relief valves could stick open

with the continued release of radioactivity even after vessel head failure.

While such a course of events cannot be dismissed, it is believed to be of

lower probability.

Table 41 gives the timing of the principal accident events for the

TMLU-SGTR scenario; Table 42 summarizes the primary system conditions during

the time period of interest.

Figure 439 illustrates the steam generator secondary side water

inventory for this sequence. Initially there is very good thermal coupling

between the primary and secondary sides of the steam generators, with rapid

depletion of the secondary water inventory. This is further illustrated in

Figure 440 which shows the primary system pressure as a function of time.

The initially effective heat transfer to the steam generators cools off the

primary system water and results in a decrease in primary system pressure

below the normal operating conditions. As the steam generator effectiveness

decreases, due to the depletion of their water inventory, the primary system

pressure rises to the pressurizer relief valve setpoint. The primary system

water inventory is given in Figure 441. It can be seen that the pressurizer

relief valve first starts to discharge primary coolant at about 50 minutes; at

this point the primary system water is still somewhat subcooled and the water

continues to heat up even as some of it is expelled from the system. At about

80 minutes the primary system is essentially saturated, as reflected by the

:21

SEQUOYAH TML W/S.G. TUBE RUPTURE3W

,3

no

0

M-0

9 -E-4 W-0

laoz

5.0

E-4

oo L6.0 26.0 46.0 60.0 �0-0 160.0 1�0-0 liao 160.0 1800 2000

TI ME (M I NUTE)

FIGURE 439. STEAM GENERATOR SECONDARY SIDE WATER INVENTORY FOR TMLU-SGTR

SEQUOYAH TML W/S.G. TUBE RUPTUREIM

0-4

2 2000.0-

P404

1500.0-

1000.0 T,m

>7

500.0

WPL4

0.0

0.0 26.o 46.0 66.o �0.0 I&O 1�0-0 1�0.0 lko lino 1.0TI ME (M I NUTE)

FIGURE 440. PRIMARY SYSTEM PRESSURE RESPONSE FOR TMLU-SGTR

00 SEQUOYAH TML W/S.G. TUBE RUPTURE6.0

5.0

4.0

z0-4 ao

Pk zo0

m

LO

0.00.0 26.0 46 66.0 ko 100.0 tw-o 1�0.0 Im.0 1�0-0 L-0

TI ME (M I NUTE)

FIGURE 441. PRIMARY SYSTEM WATER INVENTORY FOR TMLU-SGTR

4-64

faster loss of inventory. With the switching from liquid to steam discharge

through the relief valve at about 85 minutes, due to the uncovering of the

pressurizer surge line, the rate of loss of inventory decreases. A further

slowdown in inventory loss takes place as the core begins to uncover at about

104 minutes. Core slumping starts at about 154 minutes, with a rapid boiloff

of the water in the vessel head from the interaction with the core debris.

Figure 442 illustrates the maximum and average core temperatures for this

sequence. The fraction of cladding reacted and core melted are shown in

Figure 443.

The steam generator tubes were assumed to fail at 153 minutes into

the accident, with head failure predicted at 168.9 minutes. Figures 444 and

4.45 illustrate the leak rates through the pressurizer relief valve and the

ruptured steam generator tubes, respectively. It is interesting to note that

immediately after core slumping there are considerable flows through both the

relief valve and the ruptured tubes; as the primary system pressure falls,

however, the relief valve closes and only the flow through the ruptured tubes

persists. The flow to the steam generator secondary is terminated at the time

of reactor vessel failure.

It may be noted that the flow split between the relief valve and

ruptured tubes would be sensitive to the number of tubes assumed to have

failed. For smaller number of tubes than considered here more flow would be

forced out of the relief line. For a larger number of failed tubes more of

the flow would tend to go out the break, but the primary system pressure would

drop faster, with an earlier termination of the flow through the secondary

side of the steam generator. The particular scenario chosen is believed to be

a representative illustration, but is by no means a unique description of the

possible outcome of steam generator tube rupture accidents.

4.4 TBA Sequence

The accident event times for the Sequoyah TBA sequence as calculated

by MARCH 3 are summarized in Table 4.1. Note that the time of the induced

primary system break was specified by input and was chosen to take place prior

to core slumping. The core and primary system conditions at key times during

SEQUOYAH TML W/S.G. TUBE RUPTURE5000.0

MAXIMUM........... AVERAGE

4000.0

3000.0

2000.0E--4

0

1000.0

0.0

0.0 26.0 '16 k-0 80.0 100.0 120.0 140.0 160.0 1�0-0 260.0

TI ME (M I NUTE)

FIGURE 442. MAXIMUM AND AVERAGE CORE TEMPERATURES FOR TMLU-SGTR

SEQUOYAH TML W/S.G. TUBE RUPTURE1.0-

CLAD REACTED.......... CORE MELTED

0.8

............

z 0.6 -0

Cm

02-

0.00.0 26.0 46.0 60.0 86.0 160.0 M-0 140.0 160.0 IkO 200.0

TI ME (M I NUTE)

FIGURE 443. FRACTIONS CLADDING REACTED AND CORE MELTED FOR TMLU-SGTR

Ob SEQUOYAH TML W/S.G. TUBE RUPTURE30.

WA=..........

25.0

20.0-

W 15.0-

4-

14

10.0

�4 5.0

0.00.0 26.0 46-o i56-o 80.0 100.0 120.0 140.0 1�0.0 180.0 200.0

TI ME (M I NUTE)

FIGURE 444. LEAKAGE THROUGH PRESSURIZER RELIEF VALVE FOR TMLU-SGTR

SEQUOYAH TML W/S.G. TUBE RUPTURE

sum...........3m-o

am-0

WW4

OW-0

co

low-0

5W.0

0.00.0 26.0 46.0 66.0 80.0 100.0 M-0 1�0.0 1W.0 100.0 260-0

TI ME (M I NUTE)

FIGURE 445. LEAKAGE THROUGH RUPTURED STEAM GENERATOR TUBES FOR TMLU-SGTR

4-69

the accident sequence are given in Table 42, with containment conditions

summarized in Table 43.

Figure 446 illustrates the steam generator secondary side water

inventory for this sequence. After an initial -transient the steam generator

inventory is maintained at the normal level by the operation of the steam

driven auxiliary feedwater pumps. Failure of te auxiliary feedwater system

due to loss of DC control power at five hours lads to the steady boiloff of

the steam generator water. As long as the steam generators serve as effective

heat sinks, the primary coolant system pressure is maintained below normal

operating levels; as the steam generators dry ot, the primary system pressure

rises to the safety/relief valve setpoint. The primary system pressure

history is illustrated in Figure 447. The abrupt decrease in primary system

pressure at 572 minutes is due to the assumed accident induced break in the

primary system. The primary system water inventory is illustrated in

Figure 448. There is essentially no loss of pimary system inventory until

after the time of steam generator dryout. Boiloff of the primary inventory

through the pressurizer relief/safety valve leads to core uncovery and

melting. The rapid drop in primary system pressure following the induced

break leads to upper head injection as well as ccumulator discharge, with

complete core recovery. The boiloff of the injected water is seen to require

quite some time. The abrupt decrease in the primary system water inventory at

about 850 minutes is associated with core slumping; the last decrease is due

to head failure. The maximum and average core emperatures during this

sequence are illustrated in Figure 449; the fractions of cladding reacted and

core melted are shown in Figure 450. Initial ore uncovery takes place at

about 520 minutes, with start of melting predicted at about 550 minutes.

Upper head injection and accumulator discharge ollowing the induced primary

system break result in complete core recovery ad quenching, as is clearly

illustrated in these figures. After the boiloff of the injected water, the

core remelts. Figure 451 illustrates the temperatures of the gases leaving

the core and those leaking to the containment. It is interesting t observe

that the depressurization of the primary system immediately following the

assumed break leads to enhanced cladding oxidation before the core is

quenched. This enhanced oxidation is reflected in very high core exit

temperatures as well as in high temperatures of the gases released to the

containment.

SEQUOYAH TBA40.0

>-4 35.004

Z 30.0

cn 25.0

W40

2D.0P4P4 14

z Cl

15.0

E-q 10.0cnz

5.0

0.0-0.0 100.0 2OD-o 300.0 400.0 5DO.0 6(6 760.0 E&O i&0 10DO-0

TI ME (M I NUTE)

FIGURE 446. STEAM GENERATOR SECONDARY SIDE WATER INVENTORY DURING TBA SEQUENCE

SEQUOYAH TBA

2000.0

1500.0

E-4C/) 1000.0

500.0

0.0

0.0 160.0 260.0 3&.0 4&-0 5&O 6&.0 760.0 960.0 1000-0TI ME (M I NUTE)

FIGURE 447. PRIMARY SYSTEM PRESSURE HISTORY FOR TBA SEQUENCE

SEQUOYAH TBA6.0

0.4 5.0

>-4

4.0

PL4

z 3.0

04E-4

rX4 zo

0

1.0

0.0

0.0 I&O 2�0.0 360.0 4�0.0 560.0 6(60 7�0-0 660.0 9w.0 1000.0

TI ME (M I NUTE)

FIGURE 448. PRIMARY SYSTEM WATER INVENTORY FOR TBA SEQUENCE

SEQUOYAH TBA4500.0

- MAXIMUM........... AVERAGE

4000.0

3500.0

3000.0

2500.0

2000.0paE-4

1500.0

1000.0

500.0

0.0'6.0 l&.0 2&.0 360D 4(6 560.0 600.0 700.0 800.0 900.0 1000.0

TI ME (M I NUTE)

FIGURE 449. MAXIMUM AND AVERAGE CORE TEMPERATURES FOR TBA SEQUENCE

SEQUOYAH TBA1.0

CLAD REACTED........... CORE MTED

0.8-

Z 0.6-0E--4

T,

0.4-

02-

0.0 4:0.0 100.0 200.0 300.0 400.0 500.0 6W.0 700.0 800.0 900.0 1000.0

TI ME (M I NUTE)

FIGURE 450. FRACTIONS OF CLADDING REACTED AND CORE MELTED FOR TBA SEQUENCE

SEQUOYAH TBA4=.O

UFAM TO COMM........... CORE EXIT

4000.0

3500.0

3000.0

2WO-0

04 2000.0

E-415M.0

1000.0

500.0-

..................

0.0-10.0 100D 200.0 360.0 460.0 5(6 6W.0 7�0-0 860.0 S&O 1000.0

TI ME (M I NUTE)

FIGURE 451. TEMPERATURES OF GASES LEAVING THE CORE AND LEAKING TO CONTAINMENTFOR TBA SEQUENCE

4-76

Figures 452 and 453 illustrate the predicted containment pressure

and temperature histories for this sequence. Containment failure due to

hydrogen combustion was predicted to take place imediately after the

occurrence of the primary system break.

The mass of ice in the ice condenser is illustrated in Figure 454.

It is interesting to note that the ice is substantially melted by the time

that the core is predicted to undergo remelting, and is completely gone prior

to the predicted time of reactor vessel failure.

The containment sump and reactor cavity water inventories are

illustrated in Figure 455. The key point here is that the cavity is

essentially dry throughout this sequence, particularly during the corium-

concrete interaction. The predicted progression of concrete attack is

illustrated in Figure 456.

The total volume of gases leaked from the containment during this

sequence is illustrated in Figure 457. The MARCH 3 calculated distribution

of the noble gases during this sequence is given in Figure 458. The

containment leak rates used as input to the fission product transport analyses

in the containment are given in Table 44.

4.5 Additional Sequences Considered

In addition to the foregoing scenarios for which the analyses were

carried out through the release of fission products to the environment, a

number of other sequences were considered to a more limited extent.

Specifically, several sequences and variations of them were treated by MARCH 3

to determine their thermal hydraulic response. Since the thermal hydraulic

analyses indicated limited challenges to containment integrity, or limited

potential for fission product releases, these sequences were not extended

through the fission product transport analyses. The thermal hydraulic results

for the sequences considered are summarized below.

SEQUOYAH TBA80.0

Lom........... upm

".0

W-0

50.0

40.0

30.0

MD

o10.(

0.00.0 0�O-o 460.0 &6 E&O IC&O 126-0 14�0-0 16;0.0 1860

TI ME - M I NUTE)

FIGURE 452. CONTAINMENT PRESSURE RESPONSE FOR TBA SEQUENCE

SEQUOYAH TBAmm.0

LOWER,........... UPPER

r=4

P$

1500.0

1000.0

Al-

co

500.0

0

...................................

0.0

0.0 200.0 4W.0 6(6 a6o.o 1060 U&O 1460 I�V-0 la;0.0TIME - (MINUTE)

FIGURE 453. CONTAINMENT TEMPERATURE RESPONSE FOR TBA SEQUENCE

SEQUOYAH TBA25.0

20.0

0 15.00-4

z

L) 10.00--4

rZ4

0

5.0

0.06.0 260.0 460.0 660.0 800.0 1000.0 12;0.0 1460 1660 1860

TI ME - M I NUTE)

FIGURE 454. ICE INVENTORY FOR TBA SEQUENCE

SEQUOYAH TBAlbv-4 35.0

SUMP........... REACTOR CAVITY

30.0

25.0

�4

20.0-

15.0-E-4

10.0

5.0

0.00.1) 200.0 4W.0 600.0 800.0 1000.0 t-,;O.O 140.0 1�)O.O 18;0.0

TI ME - M I NUTE)

FIGURE 455. CONTAINMENT SUMP AND REACTOR CAVITY WATER INVENTORIESFOR TBA SEQUENCE

SEQUOYAH TBA70.0

VERTICAL........... RADIAL

60.0

Z 50.00

E--4

z

30.000

WE--4

20.0%-Jz0

10.0

0-00.0 260.0 4&0 60.0 E&O 1060 12�0-0 14�0-0 1660 18;0.0

TI ME - (MINUTE)

FIGURE 456. CONCRETE ATTACK FOR TBA SEQUENCE

SEQUOYAH TBAlb

8.0

CO 7.0E-4fm4

6.0-

5.0

cl)

4.0

3.00

w

zo

0> 1.0

0.00.0 200.0 400.0 600.0 aw.0 1000.0 1200.0 1400.0 16W.0 18M.0

TIME - (MINUTE)

FIGURE 457. TOTAL VOLUME OF GASES LEAKED FOR TBA SEQUENCE

SEQUOYAH TBA1.0

>408-

0.6 -

----------W

04 0.4

z0

viP-4 02

0.0

0.0 260.0 460.0 6ko aw.0 1000.0 12W.0 1400.0 16W.0 um-0

�� MM= IN CORE TI ME - (MINUTE).1 ......... FRAMON IN VESSEL----- FRAcTioN IN NTmNT

FRAC17ON TO ENVRUNT

FIGURE 458. NOBLE GAS DISTRIBUTION FOR TBA SEQUENCE

4-84

The selection of specific accident sequences for source term

analyses was predicated in part on the assumption that at least some of these

sequences would involve core melting with the ice substantially melted. This

was part of the basis for selecting the S3HF and S3H sequences. It was also

anticipated that in these sequences there would be significant likelihood of

containment failure at the time of reactor vessel breach. ASEP analyses,

which became available after the initial sequence selections had been made,

however, indicated that by far the most probable time for the combined failure

of the emergency core cooling and containment spray recirculation systems (HF)

was upon initial switchover from injection. MARCH analyses for the

initiator and such early failure of the recirculation systems show that there

would be large quantities of ice still remaining during the period of core

melting. In the case of the S3H sequence the most probable cause of emergency

core cooling recirculation system failure was found to be lack of room

cooling, with failure at about two hours after switchover to recirculation.

In this case also, it was found that significant quantities of ice would be

present during core melting. In the S3HF sequence hydrogen burning at the

time of reactor vessel failure was found to lead to significant containment

pressurization. In the SH sequence, on the other hand, it was found that

hydrogen burning at the time of reactor vessel breach did not represent a

significant challenge to containment integrity. With the availability of the

igniters and with the containment sprays and air return fans operating at

their full capacities in the S3H sequence, sufficient hydrogen was predicted

to be burned off prior to vessel failure to limit the containment pressures at

head failure well below the nominal failure pressure as specified by SARRP.

In view of the foregoing findings, a number of analytical variations

on the S3H sequence were explored in order to examine the sensitivity to the

above conclusion to variations in modeling assumptions. These analyses and

the principal findings are discussed below.

4.5.1 S3H Sequence

The initiating event for the S3H sequence was assumed to be a pump

seal failure with a characteristic size equivalent to 075 inches in diameter.

The auxiliary feedwater, emergency core cooling, containment spray, and air

4-85

return systems were assumed to operate at their full capacities. The

emergency core cooling system was assumed to fail two hours after switchover

to recirculation due to lack of pump room cooling. The hydrogen igniters were

operable and were assumed to ignite hydrogen-air mixtures when the hydrogen

concentration reached eight volume percent, subject to the availability of

oxygen and consideration of inerting by diluents.

The general timing of events for the SM sequence is summarized in

Table 41. It may be observed that the timing of the switchover to

recirculation, and hence the subsequent timing of emergency core cooling

system failure, were governed by the depletion of the refueling water storage

tank by the containment sprays. With the assumed operation of the sprays at

their full capacity, i.e., two trains, the time to the switchover is rather

short. In the present analyses it was assumed that the containment sprays and

the air return fans were started when the containment pressure reached 2 psig

(16.7 psia). The predicted times of vessel failure varied in the individual

cases in accordance with the failure modes assumed. The containment responses

varied considerably from case to case and the ey observations for each will

be summarized later.

Primary System Response. With the relatively small break in the

primary system combined with the operation of -the auxiliary feedwater and

emergency core cooling systems, the primay system was found to depressurize

very slowly. Depressurization of the secondary side of the steam generators

was not assumed in the present analyses. With both the auxiliary feedwater

and emergency core cooling systems operating it was found that all the decay

heat could be removed from the primary system ith no net steam generation.

This time period was characterized by liquid fows out of the failed pump seal

of about 320 gallons per minute. Just prior to the time of the assumed

emergency core cooling system failure about 40 percent of the decay heat was

being removed by the steam generators. After emergency core cooling system

failure, the primary coolant temperature increased and the steam generators

became even more effective; just prior to core uncovery about 81 percent of

the decay heat was being removed by the steam generators. With only a

fraction of the decay heat being released to te containment, and the

4-86

entire sump inventory being available to absorb the released energy,

relatively slow ice depletion was predicted. At the time of the reactor

vessel failure about 50 percent of the initial ice inventory was still

remaining.

Containment Response. The predicted responses of the containment

pressure are summarized below for a number of variations in sequence

assumptions.

1. In the first case considered the failure of the vessel bottom

head was based on the development of a localized failure after

heatup of the head to a depth of 2 inches. The steam and

hydrogen released from the primary system were assumed to pass

through the water in the reactor cavity. A single hydrogen

burn with a peak containment pressure of 23 psia was predicted

prior to the time of reactor vessel failure for this case. A

burn at the time of vessel failure produced a peak pressure of

about 25 psia, and shortly thereafter another burn with a peak

containment pressure of 32 psia was predicted. These burns

consumed 279, 394, and 729 lb of hydrogen, respectively.. The

largest of the three events was confined to the lower

compartment. In the longer term, a number of additional

smaller burns were predicted to take place.

2. The second case considered was identical to the first except

that the primary system blowdown following vessel failure was

not assumed to go through the reactor cavity water. The

initial burn prior to vessel failure was identical to that of

the previous case. The burn at the time of vessel failure

involved 465 lb of hydrogen to yield a peak pressure of

27 psia. The subsequent burn shortly therafter consumed 745 lb

of hydrogen and produced a containment pressure of 34 psia.

3. The third case assumed a small hole in the reactor vessel

bottom head immediately after collapse of the core and no

cooling of the primary system effluent by the reactor cavity

water. The initial burn prior to reactor vessel failure was

again identical to those in the preceding cases. With the

4-87

earlier vessel breach a peak pressure of 43 psia from the

combustion of 832 lb of hydrogen was predicted. A number of

smaller combustion events were predicted later in time.

4. In the fourth case evaluated an induced primary system breach

was assumed at the time of core slumping, with head failure

coincident with core collapse following shortly thereafter. In

this case two closely spaced burns involving 285 and 138 lb of

hydrogen and producing pressures of 23 and 24 psia,

respectively, were predicted prior to vessel breach.

Immediately following vessel breach a burn of 735 lb of

hydrogen resulted in a pressure of 39 psia.

5. The foregoing cases were based on nominal spray height

characteristics. Since the water droplets suspended in the

atmosphere are included in the combustion energy balance and

appeared to be influencing the results of the calculations, the

factors affecting spray drop residence time in the atmosphere

were refined in the next case. This had the effect of reducing

the amount of water in the atmosphere at any point in time.

With the revised spray input and again assuming vessel failure

upon core collapse, an initial burn of 226 lb of hydrogen with

a pressure of 23 psia was predicted. At the time of vessel

failure the combustion of 859 lb of hydrogen with a peak

pressure of 53 psia was predicted. The latter burn was

confined to the upper compartment of the containment. A number

of smaller burns were predicted subsequently.

6. The next case was similar to the foregoing, except for turning

off the radiation heat transfer odel in the containment. For

this case the initial burn involved 278 lb of hydrogen and

produced a containment pressure f 24 psia. The burn following

reactor vessel breach consumed 807 lb of hydrogen and produced

a pressure of 53 psia. The latter burn was confined to the

upper compartment.

7. Since the suspended water droplets in the atmosphere appeared

to have a significant influence 'on the predicted results,

another case was selected in which the fallout of the blowdown

4-88

water was enhanced by increasing a user specified drop

diameter. This change should primarily affect the predicted

events in the lower compartment, since the sprays would

dominate the behavior in the upper compartment. With this

change the initial burn prior to vessel failure was predicted

to consume 385 lb of hydrogen and yield a pressure of 26 psia.

The burn after vessel failure consumed 707 lb of hydrogen and

produced a pressure of 49 psia.

8. In this case it was assumed that the collapse of the core into

the bottom head induced a break in the primary system which

allowed the release of steam and hydrogen to the containment,

but the core debris were released later, upon head failure.

With the depressurization of the primary system through the

induced break the failure of the vessel head was considerably

delayed. For this case four burns were predicted prior to the

delayed vessel failure. They involved 278, 117, 541, and 444

lb of hydrogen and produced containment pressures of 24, 23,

34, and 26 psia, respectively.

9. The next case utilized the revised spray height parameters, but

assumed an ignition threshold of 6 volume percent hydrogen

rather than the volume percent in the other cases. With the

lower ignition threshold two burns were predicted prior to

vessel breach. They involved the combustion of 62 and 179 lb

of hydrogen and produced containment pressures of 19 and 21

psia. The burn at the time of vessel breach consumed 747 lb of

hydrogen and produced a pressure of 37 psia. The latter was

initiated in the lower compartment and propagated into the

upper compartment, but apparently due to the relatively long

duration did not produce high pressures. The lower compartment

burn had a duration of about 4 seconds, while the upper

compartment burn lasted about 11 seconds.

All of the above cases produced pressures below the 65 6 psia

failure pressure adopted by SARRP. Thus early containment failure was not

indicated to be of significant probability for any of the cases considered

4-89

here. Further, with the timing of emergency core cooling system failure

defined by ASEP, there were substantial quantities of ice predicted to be

available at the time of reactor vessel breach to absorb combustion energy as

we'll as to retain airborne radioactivity. Thu,, even if the containment were

to fail, the consequences of this sequence would not be severe.

The containment pressurization for thE' S3H sequence as indicated

here is considerably less severe than would have been inferred from earlier

analyses. The smaller break size considered leads to a more protracted core

meltdown scenario with more opportunity to burn off the hydrogen early. The

relative effectiveness of the steam generators, also due to the small break

size, leads to low rates of ice depletion. The use of full spray capacity

rather than minimum as in earlier analyses increases the suspended water that

is available to absorb combustion energy. It may be noted that the models

used to describe spray behavior during hydrogen combustion have been revised

from those in earlier versions of MARCH. The use of maximum air return fan

flow rather than minimum levels leads to more mixing of the hydrogen within

the containment and reduces the likelihood of building up the high

concentrations required to produce high pressures. All of these factors

contribute to the somewhat different results of the present analyses in

comparison with earlier work.

4.5.2 S2HF Sequence

Since the S3H sequence did not appear to involve significant

likelihood of early containment failure, some attention was given to the

examination of another of the more probable core melt scenarios, namely S2HF.

The S2HF scenario is initiated by a small break in the primary

system; for purposes of the analysis the break was taken to be 2 inches in

diameter. The combined failure of the emergency, core cooling and containment

spy-ay recirculation systems (HF) was again specified to take place immediately

upon switchover from the injection mode. The air return fans and the hydrogen

igniters were operable.

The analyses of the S2HF scenario in BMI-2104 were based on the

plugged drains as being the cause for recirculation system failure. In that

event failure of the recirculation systems follows ice depletion. For the

4 90

present analyses the principal interest was in situations that also could lead

to core melting with little or no ice available to mitigate the consequences.

The first S2HF scenario considered was based on the break being located high

in the system so that only steam could leak out. A steam versus liquid or

two-phase leak tends to prolong the blowdown process and leads to enhanced ice

depletion. The timing of the accident events for this case is summarized in

Table 41. The time of the switchover to recirculation, and hence the timing

of core cooling failure, was determined by the depletion of the refueling

water storage tank by the containment sprays. With the relatively slow steam

blowdown, considerable time was required for core uncovery. The upper head

injection tanks were predicted to empty shortly after core uncovery. The

accumulators were emptied shortly after the start of core melting. At the

start of core melt only about percent of the initial ice inventory remained.

The ice was completely melted prior to vessel failure. Four hydrogen

combustion events, producing containment pressures of 21, 30, 23, and 33 psia

were predicted to take place prior to vessel failure. Since the primary

system was essentially depressurized by the time of vessel failure, there was

no serious challenge to the containment at the time of vessel breach. Thus,

while this steam break scenario did lead to major depletion of the ice prior

to core melt, it did not present any serious early challenges to containment

integrity.

Additional analyses were performed for the S2HF sequence with the

break elevation at the primary piping level, but assuming only 80 percent of

the initial ice inventory. The latter assumption is equivalent to saying that

the ice may not be fully effective at all times, e.g., nonuniform melting in

the ice beds may permit flow breakthrough before all the ice has melted. The

timing of the accident events is summarized in Table 41. As in the other

cases considered, the timing of the switch to recirculation and hence the

timing of the core cooling failure, is controlled by depletion of the

refueling water storage tank by the spray system. With the initial liquid

break the core is seen to uncover much earlier than in the preceding case. At

the start of core melting about 27 percent of the nominal, or 34 percent of

the reduced inventory of the ice was still remaining. Complete ice melting

was predicted about an hour after the start of concrete attack. Four hydrogen

burns were predicted prior to the time of vessel failure, with peak

4 91

containment pressures ranging from 21 to 31 psia. There was no burning

predicted immediately after vessel failure, but three large burns were

predicted some time later. These produced containment pressures of 45, 40,

and 52 psia. The last two burns involved combustion of carbon monoxide as

well as hydrogen. At the time of these three burn events there was still some

ice remaining. Containment failure by long-term overpressurization was

predicted about seven hours after vessel breach.

Thus, for the most probable variations of the S2HF sequence as

defined by ASEP and the SARRP specified failure ressure of 65 psia, early

containment failure does not appear likely based on the present analyses.

4.6 Radionuclide Sources

4.6.1 Sources Within Pressure Vessel

The inventory of fission products used in these analyses is the same

as in Volume IV of BMI-2104. Table 45 provides the inventories for each of

the key fission product, actinide and structural elements. The values for the

radionuclides are based on an ORIGEN analysis for end-of-cycle conditions in a

three-region model with burnups of 11,000/22,000,133,000 MW days/tonne. The

structural masses are based on values provided i the FSAR. In Table 46 the

elements are collected into the elemental groups used in this study.

4.6.2 Sources Within the Containment

The VANESA code was used to predict the release of fission products

and inert materials to the containment during core-concrete attack. The

composition of the core materials entering the cvity is provided in

Table 47. The total release rates and compositions of released materials are

given in Tables 48 through 411 for the different scenarios.

4-92

TABLE 45. INVENTORIES OF RADIONUCLIDES AND STRUCTURAL MATERIALS

Fission Products Actinides/StructuralElement Mass (Fg) Element Mass (kg)

Kr 17.0 u 89,000

Rb 18.7 Pu 596

Sr 60.9 Cr 0

y 29.1 Mn 0

Zr 227 Fe 8,690

Nb 3.5 Ni 0

MO 197 Zr 23,100

Tc 47.2 Sn 332

Ru 132 Gd 0

Rh 26.6 Ag 2,290

Pd 66.8 Cd 144

Te 31.7 In 421

I 15.2

Xe 330

Cs 166

Ba 77.7

La 79.2

ce 167

Pr 64.5

Nd 217

Sm 43.2

Eu 11.3

Np 33.0

PM 9.2

4-93

TABLE 46. INVENTORY BY GROUP

Total MassGroup Elements (kg)

I Xe, Kr 347

2 I, Br 15.2

3 Cs, Rb 185

4 Te, Sb, Se 31.7

5 Sr 60.9

6 Ru, Rh, Pd, Mo, Tc 470

7 La, Zr, Nd, Eu, Nb, Pm, Sm, Y 684

8 Ce, Pu, Np 796

9 Ba 77.7

4 94

TABLE 47. INVENTORY OF MELT AT THE TIME OFVESSEL FAILURE FOR SEQUOYAH (kg)

S3HF1/S3HF2/S3HF3 TB TBA

Cs 4.8 4.94 0.15

I 0.454 0.468 0.014

Xe 9.73 10.0 0.30

Kr 0.501 0.516 0.015

Te 4.98 5.05 6.20

Ag (FP) 0 0 0

Sb 0 0 0

Ba 76.8 76.8 74.6

Sn 309 309 252

Tc 47.2 47.2 47.2

U02 10100 10100 101000

Zr (struct) 5930 5830 6770

Zr (FP) 58.3 57.3 66.5

Fe 47500 47500 57000

MO 197 197 197

Sr 60.9 60.9 60.8

Cr 9320 9320 9320

Ni 5180 5180 5180

Mn 0 0 0

La 79.2 79.2 79.2

Ag (struct) 2290 2290 2290

Cd 144 144 144

In 421 421 421

Ce 167 167 167

Rb 0.925 0.941 0

Br 0 0 0

Ru 132 132 132

Rh 26.6 26.6 26.6

Pd 66.8 66.8 66.8

4 95

TABLE 47. INVENTORY OF MELT T THE TIME OFVESSEL FAILURE FOR SEQUOYAH (kg)(continued)

S3HF1/S3HF2/S3HF3 TB TBA

Nd 217 217 217

Eu 11.3 11.3 11.3

Gd 0 0 0

Nb 3.5 3.5 3.5

PM 9.2 9.2 9.2

Pr 64.5 64.5 64.5

Sm 43.2 43.2 43.2

y 29.1 29.1 29.1

Np 33.0 33.0 33.0

PU 596 596 596

Se 0 0 0

FeO 0 0 1430

ZrO2 23100 23200 22000

TABLE 48. AEROSOL RELEASE DURING CORE-CONCRETE ATTACK FOR S3HF1/S3HF2

SPECIES TIME 0 t200 0 2 400 3600.0 4000.0 6000.0 7200.0 8400

rEO .2RORE-t7 .37 16 3.307 9.043 B. 70 to. 1 12. 35 1394

CA203 .3431 2 fgogf-os .2802E-03 .798JE-02 5733E-01 '99ROE-01 .7 1241 -01 .2205

NI . 1603E-05 .4St4E-02 . 590 1.003 1 t5o 44 15 2409 .4420

Mn .1139E-15 5904E-10 .3370E-07 .103SE-05 .1197E-05 .2267E-06 .743BE-07 .5553E-04

Rif .9559E-15 .4483C-09 .2424E-06 .7232E-09 .8341E-05 .1003E-05 .53IOE-08 .66GRE-06

SN .1699E-02 27451-01 IS30 .5005 .5257 2736 .1872 1.214

SR 0. 0. 0. 0. 0. 0. 0. 0.

TE .80SIE-02 .711SE-01 .1169 .1461 .1436 tl37 .999CE-01 .2838

AG 96241-02 1.701 13.67 17.63 17.30 20.10 16.37 30.21

MN 0. 0 0 0. 0. 0. 0. 0.

CAO 0. .654tE-Of 15.91 f9 as 17.95 20.79 24.89 41.48

AL203 0. .12201-04 .1232 .9146 1.517 .8709 . 272 .142RE-02

NAZO 0. 1.258 1.766 2.036 2.888 1.942 .2935 t9S9

K20 0. 21.76 t2.51 7.784 7.787 9.314 10 94 3.993

S102 0. .3031 12.91 10.21 10.76 12.02 t3.61 .2519

U02 .5428E-04 .1725E-02 .984SE-01 .8681 .9258 .2928 .1308 1.439

ZR02 t241E-04 .1302E-04 .7122E-03 .119SE-01 .1301E-Ot .29t7E-02 .7587E-03 .39SIE-04

CS20 3 761 2.701 1.149 .6796 .040S .6972 .6635 .5311

SAO .1340E-01 7164 2.3SO 2.050 t.913 1.529 .8798 .4237E-01

SRO .1065 1.328 2.571 3.453 3.260 2.078 .9923 .4471E-02

LA203 28SOE-00 .870SE-03 .8547E-01 .9949 f.076 .2931 .9883E-01 .3572E-02

CEOZ 4932E-06 .7184E-02 .3051 2 227 2.297 .7081 .2082 .17S7E-03

H13205 2r90E-08 2820E-08 1.505 4.986 4.058 .3150 0 0.

CST �2503E-Ol .6804 1.990 2.792 .7121 IOSIE-02 .391SE-19 .1173E-14

co 96.07 69.00 29.33 17.36 16.36 t7.81 ts.95 13.SS

oxinc MELT TEMPIK) 13112 11118 2tRG 2459 2472. 2336. 2247. 2184.

SOURCE RATE(GM/S) .2350 1 279 7.102 29.23 119.5 277.8 249.S SOA7

AlpnsnL DENSITY(GM/CM3) 3 7 4 3 174 3.303 3.726 3.852 3.581 3.426 4.058

AFRnsnL SZEIMICRnN) rl�llfl 7145 99ni I 0113 1 097 f 051 1 004 .61"t

TABLE 48. AEROSOL RELEASE DURING CRE-CONCRETE ATTACK FOR S3HFlf'S3HF2(continued)

SPECIES TIME 9600. 0 10800.0 12000.0 13200.0 t4400.0 15600.0 ISROO.0 I OOO.

FEO .4278 Stag .5559 .4428 .4056 .3320 2RS7 360S

CR203 .41S3 .2643 .3000 .1570 .1052 .5510F-01 .2593E-01 ISIOE-01

NI .6017 .5031 .4374 .2043 .2414 .2174 .20RS .2002

Mc .7139E-04 .6172E-04 .613GE-04 .3723E-04 .6660E-04 .200SE-03 t239E-02 .2159E-02

RU .66stE-06 .4286E-09 .299RE-06 tooaf-os .7491c-ol SJ33F-07 .4164E-07 .3679E-07

SM 1.942 t.897 1.925 1.601 1.902 2,5ql 4 030 4.710

so 0. 0. 0. 0. 0. 0. 0. 0.

.4877 .9278 .5092 .6440 .7253 .8319 gjqo 1.029

AO S7.F2 S5.34 53.89 44.55 4S.43 46.34 48.30 49.75

mm 0. 0. 0. 0. 0. 0. 0. 0.

CAO .5650 .5884 .6100 .5069 S562 .6266 .7569 .823A

AL203 igotE-03 .1641E-03 .147SE-03 .8537E-04 .8553E-04 .8831E-04 .100SE-03 .1001-03

MA20 .5713 .7355 .0617 f.00S t.092 1.103 1.022 1.062

X20 10.29 13.37 16.13 21.38 24.49 26.94 27.34 29.99

S102 .3003 .3037 .2494 .1254 .90RU-01 .5424[-Ol .300RE-01 .237ME-01

W2 1.87i i.535 .92". .89!" 1.1m 1 729 1,870

ZRO2 .423SE-04 .4993E-04 .6670t-04 .7913E-04 .932BE-04 .1133E-03 t320E-03 .1470E-03

C320 .92S2 .9131 .8690 i.077 .9049 .74RO .5694 .3770

BAO .685se-ol .6382E-01 .6345E-01 .599IF-01 .9983E-01 .926SE-01 .13S2 .1535

SRO .6563E-02 S�82E-02 .5442E-02 .442SE-02 .4949E-02 .5447E-02 .7221E-02 .783GE-02

LA203 .3607F-02 �2328E-02 f612E-02 .6170E-03 .44SOE-03 .312SE-03 .131RE-03 .249SE-05

CE02 .1587E-03 .9156t'-04 .169DE-09 .2314E-05 .2727E-05 .3313E-05 .3959E-05 .429RE-05

N9209 0. 0. 0. 0. 0. 0. 0. 0,

CS1 .2393E-14 .279RE-14 .3203E-14 .4470E-t4 .5209E-14 .6402E-14 .74rgE-f4 .830SE-14

co 23.63 23.32 22.17 27.38 23.10 19.11 14.55 1�679

OXIOE MELT TEMP(kl 2135. 2095. 2003. 19R2. 19S4. 1924. t9O4. IMAR.

SOURCE RATEICM/S) 21.41 te.19 14.20 iO.43 8.044 5 rll 3 q 1; 7 3.30R

AEROSOL VENSITYfGM/CM3) 5.179 4.882 4.676 4.130 4.041 3.946 4.076 4 032

At"nSU SZE(MIr"ON1 .4500 .425S 4041 .3060 .3415 .3t4, 2971 .27AG

TABLE 48. AEROSOL RELEASE DURING CORE-CONCRETE ATACK FOR S3HF1/S3HF2(continued)

SPECIES TIME 19200 20400.0 2 1600 22800.0 24000.0 25200 26400.0 27600

rEO 40 f2 5444 5978 0309 6706 .895 7 14 7295

CR203 1072E - I 1472E-01 12119E-01 114GE-01 103SE-01 .94SIE-02 .07 ISE 02 8097E-02

Nt 1951 1978 1770 IGA3 Is 13 .1555 1505 1463

MO 24f3E-02 .255JE-02 .257RE-02 .26OOE-02 �2622E-02 2643E-02 2664E-02 268SE-02

RU 320it-07 2789E-07 .2444E-07 .21DOE-07 i980E-07 .1821E-07 1692E-07 .159SE-07

SN 4 50 5 072 5.003 4.944 4.899 4,960 4 829 4,1104

so 0 0. 0. 0. 0. 0. 0. 0.

YE 1.123 1.192 1 206 1.218 1.227 1 23S 1 242 I 748

AG 51.47 52 2 51.18 50.32 49.81 49 02 411 52 4R.09

mm 0. 0 0. 0. 0. 0. 0. 0.

CAD .8674 .11902 .8789 .8611t RS82 .8491 .8409 .9320

AL203 966SE-04 .9194E-04 .857RE-04 .8071E-04 .7062E-04 .7320E-04 .553JE-05 .5697E-09

NA20 1.156 1 220 I 241 1.252 1.259 1.264 1,267 1.270 -D.

K20 33.72 36.79 37.99 39.Ot 39.84 40.54 41.13 41.64 1�00

S102 .217SE-01 1992E-01 .1809E-01 .16BIE-01 t544E-01 .1447E-01 .136SE-01 .129SE-01

U02 1.787 i.668 1.517 1, 393 1.292 1.207 1.139 1,074

ZR02 .1619E-03 .1727E-03 .1737E-03 .1740E-03 .173DE-03 .1730E-03 .1723E-03 .17t4E-03

CS20 .1529 0. 0. 0. 0. 0. 0. 0.

BAD .1608 .1834 .1582 .1533 . f4as .1447 .1409

SRO .78ROE-02 .7812E-02 .742BE-02 .70SIE-02 .6781E-02 .6917E-02 .6292E-02 .6073E-02

LA203 .274SE-05 .2931E-05 .2947E-05 .2952E-05 .2947E-09 .2937E-05 .2924E-05 .290SE-05

CE02 .4734E-05 .5049E-05 .507SE-05 .508SE-09 .5077E-05 .5059E-05 .5037E-05 soloc-05

N8209 0. 0. 0. 0. 0. 0. 0. 0

Cst .9147E-114 .975GE-t4 .9811E-14 .9827E-14 .9809E-14 .977SE-14 .9732E-t4 .90ROE-14

CD 3.904 0 0. 0. 0. 0. 0. 0.

oxinE MELT TEMNK) 1875, IRS3. 1854. 1847. 1841. 1835. t83O. t828

SOURCE RATEIGM/Sl 2 8 1 2 524 2,309 2 085 1.896 1.741 1.012 1.907

AEROSOL DENSITYiGM/CM3) 3 945 1 ass 3.791 3.730 3.092 3 650 3.626 3.601

AEROSOL SIZEWICIIIIN) 2680 2609 .25811 .2571 .?q57 714M 7�16 7�,77

TABLE 48. AEROSOL RELEASE DURING CORE-CONCRETE ATTACK FOR SHF1/S3HF2(continued)

SPECIES TIME 28ROO.O 30000.0 31200.0 32400.0 33600.0 34800.0 36000.0

rEO .74iS .75i3 .7592 .7658 .7713 .7757 .7794

CR203 .7564E-02 7102E-02 .6697E-02 .634SE-02 .6033E-02 .5754E-02 SS03E-02

MI .1426 .1393 .1363 t338 .1315 .1295 .1277

Ma .270SE-02 .272BE-02 275OF-02 .2772E-02 .2794E-02 .29tGE-02 .2839E-02

RU .1493E-07 .14t3E-07 f345E-07 .1286E-07 .123SE-07 lt89E-07 .1149E-07

SN 4.783 4.706 4.753 4.742 4.735 4.730 4.727

se 0. 0. 0 0. 0. 0. 0

TE 1.254 1.259 24 1.268 1.272 t.276 1.279

AG 47.71 47.37 47.08 46.93 46.62 46 44 46.28

mm 0. 0. 0 0. 0. 0. 0

CAD .8250 8179 .8111 8047 .7986 .7929 .7874

AL203 .504RE-05 .598RE-05 .6ttSE-05 .6233E-05 .6340E-OS .6439E-OS GSJOE-05

NA20 1.271 1.27t t.271 1.270 1.289 1.267 1.269

K20 42.09 42.49 42.84 43.14 43.39 43.62 43.82

S102 t233E-01 .117SE-01 .1130E-01 .1087E-01 .104SE-01 .1014E-01 .981RE-02

002 1.021 .9731 .9308 .8931 .9592 .8284 .8004

7RO2 .170111-111 .1694E-03 i653E-03 i672E-03 .1660E-03 .164RE-03 .163SE-03

CS20 0. 0. 0. 0. 0. 0. 0.

SAO .1344 .1315 .1287 .1262 t23R .12ig ltg3

SRO .5884E-02 .57IIE-02 .5553E-02 .5407E-02 .5271E-02 .5145E-02 .502RE-02

LA203 .2892E-05 .2117SE-05 .2857E-09 .2837E-05 .28t7E-05 .279GE-09 .277SE-05

CE02 .4982E-05 .4953E-05 .4922E-05 .4088E-09 4853F-05 .4817E-09 .47RIE-05

N820S 0. 0. 0. 0. 0. 0. 0.

CS1 .9627E-t4 .9570E-t4 .95tOE-14 .9444E-14 .937GE-14 .9307E-14 .923SE- 14

co 0. 0. 0. 0. 0. 0. 0.

OXIDE MELT TEMP(K) t822. 1819. isle. 1813. islo. t8oll, 1006.

SOURCE ATE(GM/S) 1.420 1.344 1.278 1.219 1.107 I.tig t.092

AEROSOL E?4SITY(GM/CM3) 3.579 3. S130 3.544 3.530 3.S18 3.507 3. 498

AEROSOL SIZE(MICRON) 2519 .2512 2506 .2500 249S 2491 74R7

TABLE 4.9. AEROSOL RELEASE DURING CORE-CONCRETE ATTACK FOR SHF3

SPECIES TIME .0 1200.0 2400.0 3600.0 4800.0 5000.0 7200.0 8400.0

FEO .287JE-17 .3720 3.308 9.043 8 07t 10 1 12 35 .11180

CR203 3433E 27 .1969E-05 .2002E-03 79RIE-02 S73GE-01 99ABE-01 7123E-01 .2205

NI 1603E-OS .45IIE-02 iS89 1.064 i.150 W4 .2408 4417

MO .1138E-15 5898E-10 .336SE-07 tO37E-OS .119GE-05 .2266E-06 7432E-07 554SE-04

:U 554E-15 .4:7:E-0: .2542JE-06 7237E-05 .:333E-05 .:G02E-OS .5306E-06 .6656E-06

N :69SE 2 .2 4 -0 i 29 5008 . 255 . 736 .1871 i.2t4

So 0 0. 0. 0. 0. 0. 0. 0.

TE .8050E-02 .711SE-01 .1189 .1462 1436 1137 .999SE-01 .2638

AG .9622E-02 1 01 i 3 66 17.63 17 30 20 16 37 3 20

MN 0 0. 0 0 0 0. 0. 0.

CAD 0. 654SE-01 Is 91 15.38 17 9 20 79 24 90 41 52

AL203 0. 1121E-04 1232 Si49 1.516 8708 .3265 .142BE-03

MA20 0. 1.258 1.766 2.038 2.688 1 939 .9937 .1965

X20 0. 21 77 12.51 7.764 7.789 9.315 10.94 3.999

S102 0. 3033 i2.Si iO.21 10.78 12 02 13.81 .2520 C>

L002 .542SE-04 .1724E-02 .984SE-01 .8684 .9252 .2925 .1307 1.438

ZRO2 .1241E-04 .1302E-04 .7119E-03 tl96E-Ol .13OOE-01 .291GE-02 .7572E-03 .3549E-04

CS20 3.761 2.701 IA49 .6796 .6406 6972 .6633 .5299

BAD t339E-01 .7161 2.350 2.050 1.913 1.529 .8784 .4235E-01

Vto .1065 1.328 2.571 3.453 3.280 2.075 .9907 .4469E-02

LA203 .2860E-0(3 .869SE-03 SS44E-Oi .9954 i.078 .2930 .9849E-Oi .356GE-02

CE02 .4932E-06 71GOE-02 .3050 2 228 2.298 .7077 .2078 .1753E-03

N8205 .2690E-08 .2820E-08 1.505 4.887 4.054 .3118 0. 0.

CST .2502E-01 .680i i SRO 2.793 .7112 iO3SE-O2 304SE-15 il52E-14

co 96.07 68.99 29.34 17.36 t 3 t7.01 t6.94 13 53

OXIDE MELT TEMP(Ki 1382. lots 2t86. 2459 2472 2336, 2247. 2184.

SOURCE RATE(GMIS) .2353 i.280 7.101 29.2S 119.8 278.0 249.9 50.99

AEROSOL DENSITY(GM/CM3) 3. 74 t 3 i4 3.303 3 726 3.652 3.581 3.428 4.057

AEROSOL SIZEiMICRON) .6536 .7445 .959i i 083 1.097 1.051 1.004 .8359

TABLE 49. AEROSOL RELEASE WRING CORE-CONCRETE ATTACK FOR SHF3(continued)

SPECIES TIME 9600.0 10900.0 12000.0 13200.0 14400 tssoo 0 16600.0 16000.0

FED 4286 .5202 .4766 .4348 .3998 .3280 2053 .3629

CR203 .4151 3655 .2327 .1570 .1048 .545SE-01 .2574E-01 .191SE-01

MI 8011 .5049 .3380 .2653 �242S .2164 .2094 .2011

MO 712SE-04 .6217E-04 .3831E-04 .37SIE-04 .8840E-04 2179E-03 .1284E-02 .217SE-02

U 6:2:E-06 .4313E-06 .17:4E-06 .10:OE-06 .7:02E-07 534:E-07 .437SE-07 .3600C-07

:N t 4 1.9 2 1. I 1. 4 t. 23 2 Ss 4.086 4.740

so 0. 0. 0. 0. 0. 0. 0. 0.

TE .4883 .5280 .5882 .6497 .7321 .8396 .9476 1.038

AG 57.93 5 4 47,96 44.81 45 72 46.6S 48 66 50.04

mm 0. 0. 0. 0. 0. 0 0. 0.

CAD .5657 .5905 .5220 .5102 .5605 .6322 .7630 .828S

AL203 .190OE-03 i849E-03 .1067E-03 .6S86E-04 .9621E-04 .9914E-04 .101SE-03 .100SE-03

MA20 .5731 .7364 .8901 i.010 t.097 1.106 1.025 1.068 4�b1

K20 10.32 13.38 17.70 21.50 24.62 26.9S 27.47 30.18 I.-

3102 .3661 .30sl �1841 .1254 .9064E-01 S36SE-01 .2SS3E-01 .238SE-01

1102 1.868 I 54S 1.007 .8281 .8920 1.119 1.752 1 879

ZRO2 .424SE-04 .494SE-04 .6527E-04 .7967E-04 .9393E-04 .1142E-03 1329E-03 .1479E-03

CS20 .9238 .9078 1.070 1.0se .6670 SS04 .3560

BAD .6659E-01 .6392E-01 .5899E-01 .602BE-01 .705SE-Ol .9391E-01 .1368 .1544

SRO .6561E-02 .5892E-02 .48SOE-02 .4452E-02 .468RE-02 SSIOE-02 .7292E-02 .7878E-02

LA203 .359GE-02 .2340E-02 .10SBE-02 .617SE-03 .4457E-03 .3i3lE-03 i324E-03 .251if-os

CE02 .1581E-03 .9203E-04 .190BE-05 .2329E-OS .274SE-OS .333SE-05 .3805E-05 .432SE-05

N0205 0. 0. 0. 0. 0. 0. 0. 0.

CST .2353E-14 .2743E-14 .3617E-14 .44i6f-14 .620SE-14 .632BE-14 .7364E-14 .319SE-14

co 23 0 23.19 27.34 26.96 22.65 18.63 14.06 9.093

OXIDE MELT TEMP(K) 2134. 2095. 2026. 1281. 1954. 1924. 1904. 1888.

SOURCE RATE(GM/S) 21.33 W19 12.14 10.36 7.975 S.3iB 3.928 3.286

AEROSOL DENSITY(GM/CM3) 5.t77 4 87 4.388 4.i32 4.044 4.001 4.081 4.034

AEROSOL SIZEIMICRON) .4497 4255 3915 .3650 .3406 .3136 .2913 .2780

TABLE 49. AEROSOL RELEASE DURING CORE-CONCRETE ATTACK FOR S3HF3(continued)

SPECIES TIME 19200.0 20400.0 21600 22800.0 24000 25200.0 26400,0 27600,0

FED .4644 .5445 .5979 .6388 .6704 8948 7229 8322

CR203 i67SE-01 i4sSE-0i il86E-Oi iWE-Ol iO33E-Oi .944DE-02 9996E-02 lit2E-01

NI .1960 .1875 .1768 .1581 .1611 .1553 .1536 1802

MO .242SE-02 .25SIE-02 .2577E-02 .2599E-02 .2620E-02 .2541E-02 2683E-02 .2692E-02

RU .3213E-07 .2762E-07 2439E-07 .2176E-07 .1977E-07 ISISE-07 176SE-07 .250SE-07

SN 4.988 5.071 5.002 4.943 4.897 4 8 4 800 5 115

so 0. 0. 0. 0. 0. 0. 0 0

TE 1.133 1.196 t.209 1.221 1.230 1.238 1.237 1.168

AG 51.79 52�21 S1.18 50 32 49�62 49 02 48 89 5 as

mm 0. 0. 0. 0. 0. 0. 0. 0

CAD 8723 .8996 .8784 .8676 .8577 .8486 .8443 .8737

AL203 .97IIE-04 .9180E-04 .856SE-04 .8062E-04 .76559-04 .7313E-04 .7171E-04 .937SE-04

MA20 1 163 1.226 t.241 1.2SI 1.2S9 1.264 1.263 1 227 C)

K20 33.94 36.80 38.00 39.01 39.83 40.53 40.70 37.36

S102 .216SE-Ot .198SE-01 .180SE-01 .1659E-01 .1542E-01 .1446E-01 .14IOE-Ot .1804E-01

U02 1.795 1.665 I.Sts 1.391 1.290 1.206 1.1so 1.222

ZR02 .1629E-03 .1727E-03 t73SE-03 .1739E-03 .1735E-03 .1730E-03 .1693E-03 .143SE-03

C320 �1299 O� 0, 0� 0. 0. O� 0.

SAO .16t7 i633 .158i .1532 .1487 i446 .1398 .1262

SRO .7922E-02 .7803E-02 .741DE-02 .707SE-02 .8774E-02 .05liE-02 .629SE-02 .5931E-02

LA203 .276SE-05 .2930E-OS .294SE-05 .29SIE-05 .294SE-OS .293SE-05 .2874E-05 .243SE-05

CE02 .4764E-05 .504SE-05 .507SE-OS .5084E-05 .5074E-05 .5057E-05 .49SIE-05 419SE-05

N8205 0. 0. 0. 0. 0. 0. 0. 0.

CSI .9030E-14 .9569E-14 .962iE-14 .9637E-14 .960E-14 .958BE-14 .936SE-14 .7953E-14

CD 3.3i7 0. 0. 0. 0. 0. 0. 0.

OXIDE MELT TEMP(K) 1874. 1863. 1854. t847 184t. 183S. 1833. 1855.

SOURCE RATE(GM/S) 2.832 2.522 2.307 2.083 1.695 1.739 1. 670 2.607

AEROSOL DENSITY(GM/CM3) 3.946 3.857 3.790 3.738 3.892 3.658 3.648 3.825

AEROSOL SIZE(MICRON) .2674 .2609 2588 .2571 .2558 .2546 .2547 .2639

TABLE 49. AEROSOL RELEASE DURING CORE-CONCRETE ATTACK FOR S3HF3(continued)

SPECIES TIME 28800.0 30000 31200 32400 33600 34800.0 36000

rEo 9180 .989S 1.047 1.091 1.12S I ISO I t66

CR203 .122SE-01 1277E-01 1286E-01 126SE-01 .122GE-Ot .1178E-01 112rE-01

NI 1967 2081 .2156 .2205 .2236 .2258 2273

MD .2724E-07 2762E-02 .280SE-02 .2854E-02 .2908E-02 .296BE-02 .3034E-02

RU .3033E-07 3417E-07 .3674E-07 .3833E-07 .3923E-07 .3967E-07 .3993E-07

sm 5.269 5.391 5 469 5.54t 5.607 5.670 5 733

so 0 0. 0. 0. 0. 0. 0.

TE i.i24 1.092 1.069 1.051 1.038 1.024 1 013

AG 53 SS 54 70 55.51 W12 56.62 57 05 57 45

mm 0 0 0. 0 0. 0 0.

CAD .8872 8927 .0928 .8892 .8829 .8747 .8651

AL203 .907SE-04 949SE-04 .9697E-04 .9748E-04 .9699C-04 .958SE-04 943iE-04

NA20 t.202 1.181 1 184 t.149 1.13S 1.121 I.tOG

K20 35.46 34 19 33.32 32.68 32.18 31.75 31.36

S102 .205SE-01 221SE-01 .230SE-01 .2341E-01 .2340E-01 .231SE-01 2274E-01 w

W2 I 232 t.210 1.169 1.119 1.064 1.010 .9564

ZRO2 .1279E-03 116SE-03 �1084E-03 .1017E-03 967'AF-04 914SE-04 9727E-04

CS20 0. 0. 0. 0. 0. 0 0

BAD .1161 .1078 .1007 .2459E-01 .891BE-01 .0433E-01 .7993E-01

SRO SS99E-02 527RE-02 .497SE-02 4696E-02 .443SE-02 .4ig6E-02 3974E-02

LA203 .2172E-05 11982E-05 .1840E-OS .1727E-OS .1633E-05 iSS3E-OS 1,49IE-09

CE02 .3741E-05 .341SE-05 .3169E-05 .297SE-05 .2814E-05 267SE-05 .2552E-05

NO205 0. 0. 0. 0. 0. 0 0.

CS1 .7092E-14 .6473E-14 .6008E-14 .5639E-14 .5334E-14 .9071E-i4 .4837E-14

co 0 0 0 0 0. 0. 0

OXIDE MELT TEMP(K) 1867. 1874 1879. 1881. 1882. 1882. 1882.

SOURCE ATEIC.M/S) 3 232 3 9R 4.049 4.307 4.493 4 629 4 724

AEROSOL OENStTY(GM/CM31 3 934 4.011 4.066 4 107 4.140 4,189 4.196

AEROSOL SIZE(MICRON) .2697 2739 2769 .2791 2807 .2819 .2829

TABLE 410. AEROSOL RELEASE DURING CORE-CONCRETE ATTACK FOR TB

SPECIES TIME .0 1700 2400 3000.0 4ROO.0 6000.0 7200.0 R400

rEo .7666E-17 SG14 3.RS7 8.5139 2.264 10.9s 7 441 .40OR

CR203 .17SOE-26 .4972E-05 .433SE-03 .1154E-01 .7469E-Ot 1036 .5289E-01 .3213

ml .4474E-05 .850dE-02 .2259 1,363 .8834 .3368 .2192 .4531

MO .6050E-IS .171SE-09 .6423E-07 .1012E-05 .7593E-06 .139SE-06 Ssotr-o? .429q[-04

RIP SOISE-14 .1291E-os .459SE-06 .1120E-04 .526SE-05 9902E-05 .393GE-06 .3899E-06

SN .243RE-02 3699r-ot .1891 .5920 .4401 .2300 IR84 1 44

so 0 0. 0. 0. 0. 0. 0. 0.

If .1117E-01 .844SE-01 .1221 IS36 t37S .1079 .1122 .5160

AG tgOSE-01 2.546 16.62 16.69 18.06 19.90 16.47 51.12

mm 0. 0. 0. 0. 0. 0. 0. 0.

CAD 0. .1099 14 68 16.0s 10.72 22.08 23 4 3 .4800

AL203 0 .2717E-04 .1699 1.148 1.313 6304 .862SE-01 .134SE-03

NA20 o 1.441 I 31 2.031 2.843 1 324 9918 .677R

K20 0. 22.98 11.23 7.392 8.229 9.832 13.40 11.92 C)

S102 0. .5836 14.23 9.823 11.12 12.45 16.72 .2644

1102 56SIE-04 .3452E-02 .1473 1 t49 .0770 .2065 .1095 1 345

ZR02 i3olE-04 .2i7flE-04 .1203E-02 .1714E-01 .8703E-02 .175SE-02 .264GE-03 .4993E-04

CS20 3.858 2.839 1.055 .6499 .6074 .7076 7666 I.t92

�AO .226RE-01 .9608 2.233 2.011 t.837 1. 290 .4145 .6054E-01

SRO .1507 I.S77 2,859 3.564 2.957 1.824 .4298 .5093 02

LA203 .2993E-08 .1912E-02 .134S 1.359 .7579 .1900 .916GE-Ol .21ROE-02

C102 .51021-06 .1379E-01 .4370 2.8SI 1.687 A921 .72S3E-01 .2187E-04

P18205 .281SE-08 .2702E-08 1.843 5.598 2.897 0. 0. 0.

CSI .404RE-01 .8959 2.212 2.883 .3263 .686GE-05 .4022E-IS .279GE-14

co 9S.90 65 58 26 22 18.15 17.08 17.59 Iq 05 29.63

OXIDE MELT TEMP(K) 1424. f867 , 2234. 2498 . 2433. 2298. 220. 2089,

SOURCE RATE(GM/S) .3033 1.500 8.695 39.59 2C4.4 303.2 287.3 31.07

AEROSOL DEASITY((,M/CM3) 3.741 3.157 3.387 3.739 3.oi3 3.55G 3 749 4.789

AEROSOL SIZEIMICRON) .6498 .7572 .90104 1.102 1.088 1 032 .9476 .4325

TABLE 410. AEROSOL RELEASE DURING CORE-CONCRETE ATTACK FOR T8(continued)

SPECIES TIME 9600.0 10800.0 12000 13200.0 14400 15600.0 frA0 0 111000.0

rEO .4441 .4212 .3820 .3102 .3025 .4008 5055 .57GR

CR203 .2081 .148S .919SE-01 .4487E-01 .2291E-01 .1859E-ol 1644E-01 .143GE-01

Hi 3040 .2628 .2376 .2189 .2094 .2013 .1971 .1870

MO .3273E-04 .4192E-04 .861RE-04 �3618E-03 t72RE-02 .2267E-02 .249SE-02 .2564E-02

Rif .1449E-06 974BE-07 .7019E-07 .5214E-07 .4284E-07 .3620E-07 .3169E-07 .2763E-07

SH 1.537 1 sag 2.048 2.962 4.436 4.838 5.080 5.0115

se 0. 0. 0. 0. 0. 0. 0 0.

YE S9qA 6692 .7621 .8758 .9793 t.072 1 173 1.208

AG 45.54 44 97 45.89 47.39 49.15 50.59 52 55 52 14

Awl 0 0. 0. 0. 0. 0. 0 0

CAD .4904 .5180 .5734 .6671 .7897 .8405 .8R63 .8077

AL203 .9403E-04 8649E-04 .97OIE-04 .942SE-04 .1039E-03 .1004E-03 .971SE-04 .911SE-04

MA20 .8989 1.024 1.101 1.078 1.016 1.090 1.192 1 228CO

K20 18.27 21.91 25.08 26.80 27.65 30.97 34 94 36 as (J"

S102 .1613 .1198 .818SE-Of .4654E-01 .2740E-01 23SOE-Ol .2171E.01 .1974E-01

U02 .8974 8420 .9328 1.280 1.882 1.864 1.775 1 674

ZRO2 .6834E-04 .8119E-04 .9737E-04 .1177E-03 .13SIE-03 'ISOOE-03 .165SE-03 .1699E-03

CS70 1.183 1.059 .6728 .7042 .5177 .3067 S72RE-01 0.

BAD .572SE-01 .617tE-01 .745SE-01 .1044 .1443 .1555 .1627 .1800

SRO 460SE-02 .4479E-02 .4814E-02 .5959E-02 .7592E-02 7RBOE-02 .7919F-02 .767SE-02

LA203 asset-03 SgOOE-03 .412SE-03 .298SE-03 .1331E-03 .25489-05 .2807E-05 .2883E-05

CE02 .199BE-os .2373E-05 .294SE-OS .3439E-05 .39SOE-05 .4386E-05 .4R37E-05 .496GE-05

H82OS 0. 0. 0. 0. 0. 0. 0. 0.

CS1 .3827E-14 .4547E-t4 S453E-14 .6589E-14 .75GRE-14 .8403E-14 .920GE-14 AS14F-14

co 29.41 26.3f 21.87 17.50 12.67 7.623 1.424 0.

OXIDE MELT TEMP(k) 2012. 1978. 1948. 1921. t9ol. less. f872. t802.

SOURCE RATE(C.M/5) 0.02 10.67 7.589 5.047 3.916 3 308 2.824 2 524

AEROSOL DENSITYjGM/CM3) 4 26 4.122 4.035 4.037 4.098 4 021 3.932 3.854

AEP(ISOL STZE(MICRON) 3873 .1618 .3352 3082 .2081 27SR 761,2 2r1:1

TABLE 410. AEROSOL RELEASE DURING CORE-CONCRETE ATTACK FOR T8(continued)

srEctes TIME 19200 20400 21600 22800.0 24000.0 75200.0 26400 27000.0

rEo 625 1 .6616 .689n . 7113 .7281 .74 15 .7524 .8353

CR203 1260C -01 . 112SE-01 1019E-ol .9334E-02 .8617E-02 ROt2E-02 .749SE-02 R9O4E 02

pit .1765 .1081 .1613 .1557 .1508 .14136 1429 t629

Ho .258BE-02 .20IOE-02 .263IF-02 .2653E-02 .2074E-02 .209SE-02 2717E-02 .274SE-02

RU .242SE-07 �2174E-07 .19ROE-07 A824E-07 .169GE-07 .156SE-07 .1499E-07 ISR7E-07

SN 5.017 4.962 4.918 4 882 4.852 4. 28 4.808 5.023

SR 0. 0. 0. 0. 0. 0. 0. 0.

IE 1.222 1.233 1.243 1.251 i.2SO 1.265 1 270 i.210

An 51.14 50 2 49 64 49.07 48.58 49.15 47.79 so 24

MN 0 0 0 0. 0. 0. 0. 0.

CAD .8702 Bass asse .64GR .6363 .8303 .8229 .8460

AL203 RSIOE-04 .802qE-04 �76211E-04 .7294E-04 .5530E-05 5002E-09 .5839E-05 7390E-04

NA20 1 242 1 251 1.2SR t.263 1.206 28 1,268 1 241 C)

K20 3R.03 38 99 39,79 40.46 4i.04 4i.SS 41 so 3 23

3102 1794E-01 165JE-01 .1539E-ol .1444E-01 .1363E-01 .120311-01 .1232E-01 .1503c-ot

U02 1 479 1.362 1,260 1,186 1.118 1.05R t.007 1.058

?no? .170SE-03 .111it-03 t7f9E_0l i7O4E-03 isgaf_03 16901-03 .1681E-03 .1474E-03

CS20 0 0. 0. 0. 0. 0. 0. 0.

"AO .1550 1503 .1461 .1423 1387 135S .1325 .1222

SRO .725SE-02 .6912E-02 .6649E-02 .64OOE-02 .6178S-02 .59791-02 .579SE-02 SS52E-02

LA203 .2899E-05 .2903E-05 .29OOE-05 .28gtE-05 .2880E-OS .2847E-05 .2852E-05 .25OIE-05

CFO2 .4994E-05 q002E-05 .4995E-05 .49SIE-05 .4902E-05 4940E-05 .4911E O .43091 or.

tin2os 0 0 0 0. 0. 0. 0. 0.

CSI .95GRE-14 .9583E-14 .9570E-14 .9543E-14 .9507E-14 .9403E-14 .9412E-14 .8254 14

CD 0 0 0. 0. 0 0. 0. 0.

OXIDE MELT EMPIX) 1854 11147. tB40 1835. ill3l. 1826. 1823. 1840.

SOUnCE RATE(GM/Sl 2 20 2.040 1 859 1.713 1.593 1.495 1.515 2.081

AEROSOL DENSITYIGM/CM3) 3 788 3.736 3.095 3.660 3.631 3.605 3.584 2.724

AEROSOL SIZEIMICRON) 2592 .2575 .2562 2550 .2540 .2531 .2S21 2Sq4

TABLE 410. AEROSOL RELEASE DURING CORE-CONCRETE ATTACK FOR TB,(continued)

SPECIES TIME 211800.0 30000.0 31200.0 32400.0 33600.0 34800.0 36000

rEo .9282 1.005 1.065 1.112 t.146 1.170 t.lRR

CR203 .104SE-01 .1143E-01 118SE-01 .119SE-01 .117BE-Of .1142E-01 .1099E-01

tit .1850 .2010 .2121 2198 .2251 .2288 2314

MD .277SE-02 .2813E-02 .2057E-02 .2907E-02 .2964E-02 .3027E-02 .3097E-02

RU .2620E-07 .3133E-07 .35f3E-07 .377SE-07 .394SE-07 .4049E-07 .4109E-07

SN 5.238 5.398 5.504 S.S97 S.879 5.754 5.1126

so 0. 0. 0. 0. 0. 0. 0.

TE 1.159 1.118 1.087 1.063 1.04S 1.030 1 Ole

AG 52.64 54.28 55.44 56.30 50.97 57,S3 S8.02

mm 0. 0. 0. 0. 0. 0. 0.

CAD .8667 .8770 .9802 8784 .8733 .8658 8562

AL203 .8334E-04 .6964E-04 .9337E-04 .951?E-04 gSSSE-04 SSOSE-04 939SE-04

MA20 I 208 1.181 1.159 l.t39 1.122 i.105 1.088CO

K20 36.52 34.69 33.42 32.51 31.80 31.23 30.74 -4

S102 .1821E-01 .2049E-01 .2tgSE-01 .227SE-01 .23OGE-Ol .2302E-01 .227SE-01

U02 1.102 1.108 1.089 1.0154 1.012 .9653 .9183

ZRG2 i;Is3E-uj .1150E-03 .1053E-03 .978SE-04 .9193E-04 .869GE-04 .8264E-04

CS20 0. 0. 0. 0. 0. 0. 0.

�AO .1ill? .1033 .9620E-Ot .9012E-01 .8482E-01 .6010E-01 .75051-01

SRO .5273E-02 .4987E-02 .4712E-02 .44S3E-02 .4212E-02 .396GE-02 .3777E-02

LA203 .2177E-OS .19SIE-05 t787E-03 t6slEOs ISSOE-OS .1475E-05 .14o2E-OS

CE02 .37SOE-05 33R2E-09 307RE-05 .28SIE-05 2687E-OS .2542E-05 .211GE-OS

PIR20S 0. 0. 0. 0. 0. 0. 0.

CSI .718SE-14 6440E-14 S89SE-14 .5492E-14 .514SE-14 .4870E-14 .4n28E-14

co 0. 0. 0. 0. 0. 0. 0.

OXIDE MELT TEMP(K) 1857. lase. 1675. JARO 1982. tRR3 IR94.

S01111CE RATE(GM/Sf 2 819 3 417 3.888 4 248 4.S17 4.718 4 83

AEROSOL DENSITY(C.M/CM3) 3.873 3.980 4 059 4.118 4.105 4.204 4 29

AEROSOL SZEIMICRON) 2970 .2720 2767 .2798 .21118 2434 21147

TABLE 411. AEROSOL RELEASE DURING CORE-CONCRETE ATTACK FOR TBA

WECIES TIINE .0 1200.0 2400.0 3000.0 4800.0 mmmm.o 7200.0 4400.0

FEO ". so 10.70 5.764 9.404 10.32 12.50 .8831 .7797

MM .5021E-20 .7880E-19 .9570E-14 .1250E-17 . ISISE-17 .2807E-17 . i269 .2119

"I .831st-ol .4011 1.264 .7063 .3942 .2233 .41147 .3366

.6019-00 .20069-00 .1541E-06 .670GE-08 . NME-00 .7131E-07 .8417E-04 .480BE-04

.6340E-07 .1454E-09 .1070E-04 .474GE-05 .1279E-05 .5088E-Os .741411-04 .4164E-08

sm .9882E-01 .2072 .4461 .2214 .1096 .1417 1.047 .8906

0. 0. 0. 0. 0. 0. 0. 0.

low I"I .1093 .1510 .1229 .1139 .2291 .3150

AG 10.41 20.83 17.02 12.30 19.73 14.93 30.30 90.88

IN 0. 0. 0. 0. 0. 0. 0. 0.

CAD 0. 13.76 17.06 Igloo 20.80 26.19 44. NO 90.41

AL202 0. so" I.S80 1.95i SO" .49111 .17199-03 .12*U-CM

NA20 0. 1.806 2.243 2.854 2.902 1.021 .1403 .2795 00

X20 0. 9.137 7.518 8.216 9473 Ii.24 3.825 5.221

S102 0. II.66 10.23 11.26 12.14 13.96 .2019 .2276

U02 .61181-01 .22" 1.208 .61032 .2093 .1274 1.777 I.M

ZM02 .2672E-03 .3264E-02 .172GE-01 .8402E-02 .2573E-02 .9012E-03 AME-04 .2343E-04

C520 .591SE-01 .2172E-01 .164SE-01 .1744E-01 .1413E-01 .170SE-01 .1027E-01 .92241-02

an 2.976 1.982 1.900 IAO3 1.403 1.003 .447GE-01 .281101-el

Sao 3.173 2.874 3.594 2.987 2.006 1. In .4809E-02 .3me-02

LA202 .392111-01 .3"s 1.450 .7794 .2750 .11" .43M-02 .2471E-02

CE02 .172t .9105 2. 983 1.719 .6722 .2584 .20799-03 .1070E-03

W205 1.237 2.833 5.672 3.006 .6790 0 . 0. 0.

csI .7087E-01 .7304E-01 ."22E-01 .1074E-01 .16222-03 .440SE-18 .1402E-14 .147SE-14

co 54.78 21.67 16.24 17.33 19.01 17.46 10.20 9. In

OXIDE ELT TEMPW 2070. 2328. 2504. 2434. 2330. 2253. 2199. 2191.

SOLMICE RATE(OW/5) 3.982 12.16 32.94 87.50 271.8 284.1 44.07 32.37

AE1OsOL DENSITY(401/00) 4.9810 3.704 2.706 3.610 2.549 3.393 4.044 3.778

AEROSOL SIZE(MICID ) .7008 1.027 1.099 1.001 1.054 1.004 .8216 .6103

TABLE 411. AEROSOL RELEASE DURING CORE-CONCRETE ATTACK FOR TBA(continued)

SPECIES TIME 91100.0 Imerim.0 12000.0 13200.0 14400.0 104100.0 16800.0 18000.0

FED 1.489 1.401 1.310 .9562 .7938 .5794 .W92 .6815

=203 .4131 .3090 .3165 .1727 .1007 .909711-01 .27118E-01 .2232E-01

m .9790 .50110 .4574 .2846 .2613 .2409 .2315 .2122

.411792-04 .7574 04 .7694E-04 .4997E-04 .104GE-03 BME-03 .20M-02 .225BE-02

.91ME-08 .4067:--06 .311se-als .106it-08 .7943E-07 .1179GE-07 .47922-07 .3944E-07

gm f.730 1.717 1.748 1.501 1.879 2.838 4.071 4.082

SO 0. 0. 0. 0. 0. 0. 0. 0.

TE .6779 .7238 .7696 .8000 1.0" 1.217 1.353 1.377

50.40 $7.72 98.99 48.27 49.84 92.12 64.13 62.52

0. 0. 0. 0. 0. 0. 0. 0.

.7228 .7423 .7624 .6537 .7304 .0839 i.038 1.029

AL202 "48E-M .2017E-03 IME-03 .11IOE-03 .114OL-03 .1261E-03 .136SE-03 .12401-03

NA20 .7249 .8789 .9961 1.20i 1.321 1.317 1.274 1.295

K20 Is." 16.41 Well ". 02 Will 33.19 34.90 38.52

SIDE .2929 .3363 .2878 .1487 tO33 .91HPOE-01 .3540E-Ol .30ilE-01

U02 2.102 1.837 1.079 1.0111 i.190 1.622 2.290 2_0105

ZROM MM-04 .907W-04 .974SE-04 gay.1-04 . I low-as .144OL-03 .1922E-03 .1643E-03

C220 .1819E-Ol .1737E-01 .1901E-01 .1867E-01 .1263E-01 .98M-02 0. 0.

an .749211-oi .7343E-01 .7382E-01 .72019-01 .07539-01 .1272 I"� .1819

No .7227E-02 .972GE-02 .937SE-02 .938GE-02 U37E-02 .74249-02 .90ME-02 .854SE-02

LA208 .3297E-02 .2447E-02 .1492E-02 .7293E-02 .62ON-02 .367SE-03 .law-09 .27OIE-05

cm .1341E-03 .88SOE-04 .1974E-05 .2831E-08 .32949-00 .4213E-05 .474SE-05 .4808E-06

W209 0. 0. 0. 0. 0. 0. 0. 0.

CSI .34OOE-14 .38VA-14 .425a -14 .810BE-14 .73IOE-14 .9006E - 14 .10231!-13 .10971-i3

co i 9. 07 17.26 15.90 18.95 12.48 9.693 0. 0.

OXIDE MELT EW(K) 2119. 20". 2m. fell. 1962. Ml. 1003. 1890.

SOMM RATE(M/S) 12.77 10.39 12.11 10.17 7.672 5.196 4.069 3.861

AEROSOL ENSITY(M/Cna) 5.020 4.837 4.957 4.0119 3. 999 3.940 3.971 3.888

AEROSOL IZE(MI .4105 .3937 .3788 .3410 .3199 .2874 .2701 .20"

TABLE 411. AEROSOL RELEASE DURING CORE-CONCRETE ATTACK FOR TBA(continued)

SPECIES TIM 19200.0 20400.0 219M.0 22600.0 24000.0 25200.0 26400.0 276M. 0

PEO .7012 .8636 .9332 .9920 1.045 1.099 IA28 1.102

C0203 AME-01 .1753E-01 AME-01 .15189-01 .1434E-01 .1363E-01 .130if-ol .124SE-01

ml .2007 .1940 .1099 .1873 A887 .1847 .1843 .1841

'O .:2:7:-02 .231::-:2 .2337:-02 .2359:-02 .2380E-02 .2404:-02 .243::-02 .24:::-02

Itu 4 4 07 .322 - 7 .3071 07 .297i 07 .29022-07 .2854 07 .202 -07 .27 -07

sm 4.035 4.007 3.995 3.993 3.999 4.011 4.027 4.047

so 0. 0. 0. 0. 0. 0. 0. 0.

TE 1.286 1.389 i.322 1.376 1. 3" i.362 1.355 1.348

AG 51.49 90.89 50.54 50.37 90.30 50.31 50.39 50.52

m 0. 0. 0. 0. 0. 0. 0. 0.

CAO i.016 1.010 1.004 .9982 .9929 .9875 .9822 .9706

AL202 I IWE-03 A109E-03 .10749-03 .10471-03 .10202-03 .1008E-03 .99279-04 .9782E-04

NM 1.314 1.325 1.231 i.334 1.324 i.232 1.328 1.323

K20 27.71 38.44 38.67 I 39.23 39.27 39.22 39.13 CD

S102 .27532-Oi .29032-01 .25"E-01 .2437E-01 .2383E-01 .2339F-01 .2302E-Oi .2287E-01

W2 1.852 1.692 1.865 i.400 1.370 1.292 i.222 1.180

ZWO2 A6232-03 .1582E-03 .1532E-03 .14799-03 .1426E-03 A374E-03 .1323E-03 .12759-03

C520 0. 0. 0. 0. 0. 0. 0. 0.

VIAO .1538 .t4w .1300 .1908 .1243 .1182 A1218 .1075

sm .78799-02 .7300E-02 .6942E-02 .06529-02 .620OL-02 .9082E-02 .659it-02 .6324E-02

LA203 .2757E-05 .2687E-05 .26ME-06 .29121-05 .24221-06 .2322f-05 .22479-05 .216SE-05

CE02 .4749E-05 .4629E-05 .4483E-09 .43209-05 .4172E-05 .4020E-09 .38721-09 .3731E-05

W205 0. 0. 0. 0. 0. 0. 0. 0.

C31 .1024E-13 .99911-14 .911901-14 .93319-14 .999BE-14 SMOE-14 .8349E-14 .9045E-14

CD 0. 0. 0. 0. 0. 0. 0. 0.

OXIVE WLT TOW(K) loll. 1876. 1873. 1870. lose. 1867, low. lees.

6011 CE RAT11(001/2) 3.372 3.210 3.124 3.090 3.099 3.047 3.049 3.005

AER050 0EMITY(ON/CM) 3.801 3.761 3.737 3.724 3.718 3.716 3.719 3.724

AEMML SIZE(MIMM) .2661 .2647 .2947 .2690 .2694 .2099 .201155 .2671

TABLE 411. AEROSOL RELEASE DURING CORE-CONCRETE ATTACK FOR TM(continued)

SPECIES TIM 28800.0 30000.0 31200.0 32400.0 33800.0 24800.0 38000.0

rEO i.190 1.213 1.232 1.248 1. m 1.269 1.275

CH203 iI931-01 Ii4ff-oi .1097E-01 W92E-01 . 10101!-Oi .989SE-02 .9317E-02

"I .1840 .1840 .1840 .1041 .1843 .1645 .1849

no .24SOE-02 .2521E-02 .25639-02 .2591E-02 .2630E-02 .2971E-02 .2714E-02

NJ .2777E-07 .2797E-07 .2737E-07 .2717E-07 .2999E-07 .2683E-07 .287OF-07

gm 4.069 4.093 4.120 4.i4d 4.178 4.210 4.244

se 0. 0. 0. 0. 0. 0. 0.

TE 1.341 i.334 1.228 1.222 1.319 1.31i 1.309

AG go. SO 50.83 51.01 51.21 51.43 91.67 51.93

RN 0 0. 0. 0. 0. 0. 0.

CAO .9705 ."39 . 9" .9495 W7 .9337 .9254

AL202 .964OE-04 .9498E-04 .9354E-04 .92091-04 .905SE-04 .8922E-04 .8783E-04

NM 1.317 i.310 1.303 1.295 1.286 1.276 1. "

920 39.02 39.88 30.73 38.96 38.27 38.19 37.92

3102 .22322-01 .21961-01 .2158E-01 .21192-01 .20799-01 .2040E-01 .200iE-01

tw2 1.102 1.049 .9994 .9524 .9105 .8705 .8333

ZRO2 .1230E-03 .1197E-03 .1147E-03 .1iOPE-03 .1073E-03 .1039E-03 .100GE-08

C520 . 0. 0. 0. 0. 0. 0.

VW '.1027 .98279-01 .94169-01 .9033E-01 .067SE-01 .233BE-01 .9021E-01

Sao .90792-02 .48912-02 .463VE-02 .444if-02 .4257E-02 .4064E-02 .3921E-02

LA202 .20492-06 .2Oi7E-05 WSE-05 .1804E-08 .1823E-05 .176SE-05 .1700E-05

cm .3999E-05 .3474E-00 .3357E-08 .324GE-08 .31411-05 .30409-05 .2944E-05

M205 0 . 0. 0. 0. 0. 0. 0.

Csl .7759E-14 .749it-i4 .723BE-14 .699ge-14 .8772E-14 MBE-14 .6347E-14

co 0. 0. 0. 0. 0. 0. 0.

OXIDE T IMM 1064. 1004. 1863. 1062. 1861. 1890. 1899.

SOURCE RATE(SP/2) 3.006 3.106 3.124 3.141 3.1se 3.171 3.191

AEROSOL OEWlTV(W/CW3) 3.720 3.738 3.746 3.756 3.787 3.779 3.793

AEROSOL SIZE(NICRON) .2670 .2601 .2880 .2690 .2694 .2998 .2702

5-1

5. RADIONUCLIDE RELEASE AND TRANSPORT

5.1 _S3HF Sequence

Three scenarios are examined which each have the same primary system

behavior.

5.1.1 Release and Transport in RCS

An overview of fission product behavior in the RCS during the period

from beginning of radionuclide release from fuel pins to the time of head

failure is provided in Tables 5.1 and 52 and Figures 5.1 through 54.

Table 5.1 gives the progression, in 23 minute intervals, of the

total deposit on RCS internal structural surfaces of three volatile fission

product species, CsI, CsOH, and Te, as well as that of aerosol material. For

comparison the total masses of these species released from the fuel as

functions of time are also presented. The data show a generally monotonic

increase in captured mass for all four species with slight reevaporation of

the volatile fission products late in the sequence. At the time of head

failure, 68.9 percent of CsI, 76.1 percent of CsOH, 91.0 percent of Te, and

74.1 percent of aerosol mass released from the fuel to that time are retained

on internal RCS structural surfaces. The differences between these retention

fractions tell the physico-chemical story. CsI can only condense on structural

surfaces. CsOH has a moderate reaction rate with stainless steel while Te has

a high reaction rate with stainless steel. Aerosol particles are removed

predominantly by settling, except in the steam generator, where high

turbulence, surface area, and plug flow combine -to yield inertial deposition as

the dominant process.

Table 52 gives a summary of conditions at the time of head failure

for each of the elemental groups of fission products.

Figure 5.1 through 54 give a more detailed view of fission product

and aerosol behavior in the RCS as a function of time. Released masses are

shown for CsI, CsOH, Te, and total aerosols for each control volume with time.

These are the masses of the given species that have transported beyond the

indicated control volume and are therefore no longer subject to capture by

TABLE 5.1. MASSES OF DOMINANT SPECIES RELEASED FRO14 FUEL (TOTAL)AND RETAINED ON RCS STRUCTURES (RET) AS A-FUNCTION OFTIME -- S3HF SEQUENCE

CS1 CSOH TE AEROSOLTIME PET TOTAL PET TOTAL PET TOTAL PET TOTAL(S) (KG) (KG) (KG) (KG) (KG) (KG) (KG) (KG)

21977. .2 1.7 3.2 i3.2 .2 .7 3.3 32.0

22115. 1.2 3.7 11.1 25.3 .8 1.3 17.5 47.4

22255. 2.8 6.3 22.0 40.6 1.3 2.2 32.3 64.6

22392. 4.8 9.1 35.2 57.1 2.1 3.3 47.5 85.6

22531. 6.9 12.i 5O.i 74.6 3.1 4.6 66.4 iO9.5

22669. 9.0 15.1 64.7 92.3 4.2 8.3 87.2 132.1

22810. 10.7 17.9 79.8 108.9 5.9 9.9 108.5 155.8

22950. 12.6 20.4 91.1 i23.5 8.1 il.8 130.0 179.2 Ln

23087. 15.6 22.4 101.4 135.7 10.8 14.2 153.3 203.3

23225. 18.1 24.2 12O.i 146.2 i3.7 17.i 178.8 228.1

23384. 19.9 25.8 132.0 155.8 16.9 20.9 204.0 254.7

23502. 2i.3 27.6 141.2 167.3 20.8 24.1 230.3 296.8

23641. 2i.6 29.6 143.6 i8O.2 23.3 25.5 258.4 329.9

23780. 2i.4 30.0 143.6 183.0 24.0 26.0 263.1 343.6

239i9. 21.i 30.1 i42.2 183.5 24.4 26.4 264.4 352.7

24059. 21.0 30.1 141.3 183.9 24.6 26.6 264.9 357.1

24i9g. 20.9 30.i 140.9 i84.i 24.5 28.7 265.1 357.5

24339. 20.9 30.2 140.4 i84.i 24.4 26.7 265.1 357.6

24475. 20.8 30.2 140.2 184.2 24.4 26.7 265�1 357.6

24618. 20:8 30.2 140.1 184.2 24.3 26.7 265.i 357.6

TABLE 52. MASSES OF RADIONUCLIDE RELEASED FROM FUELAND RETAINED ON RCS (BY GROUP) -- S3HF SEQUENCE

RELEASED RETAINEDGROUP (KG) (KG)

1 14.7 10.2

Cs 178.8 i35.0

TE 28.7 24.3

SR .0 .0

RU .0 .0

LA .0 .0

NG 336.4 .0

CE .0 .0

OA .9 .7

35

30

25-

20-

15- LegendFUEL

(nIAJ CORE

10

of UPPER PLENUMHOT LEG

STEAM GENERATOR

021500 22000 22500 23600 23500 24800 24900 25600

TIME (sec)

FIGURE 5.1. MASS OF CsI RELEASED FROM INDICATED RCS COMPONENT AS A FUNCTION OFTIME - 3HF SEQUENCE

200

150

loo

LegendFUEL

CORE50 UPPER PLENUM

HOT LEG

STEAM GENERATOR

04-21500 22000 22500 23000 23500 24000 24500 25600

TIME (sec)

FIGURE 52. MASS OF CsOH RELEASED FROM INDICATED RCS COMPONENT AS A FUNCTION OFTIME - S3HF SEQUENCE

30-

25

20

15

LegendFUEL

loCORE

UPPER PLENUM

5 HOT LEGSTEAM GENERATOR

0i1500 22600 22iOO 23000 23500 24000 NgOO 25600

TIME (sec)

FIGURE 53. MASS OF TE RELEASED FROM INDICATED RCS COMPONENT AS A FUNCTION OFTIME - S3HF SEQUENCE

400-

350-

300-

op-%co

250-

200

Legend150 FUEL

CORE

100 UPPER PLENUM

HOT LEG

50 STEAM GENERATOF

J

021500 22000 22500 2.3000 2.3;00 24000 24�00 25000

TIME (sec)

FIGURE 54. MASS OF AEROSOL RELEASED FROM INDICATED RCS COMPONENT AS A FUNCTION OFTIME - 3HF SEQUENCE

5-8

that volume. The escaped mass from the steam generator is therefore the

(cumulative) mass released to the containment. The escaped mass from the fuel

gives the (cumulative) mass released for transport through the RCS. Only the

upper plenum and steam generator tubes are seen to play a significant role.

5.1.2 Release and Transport in Containmentfor SiHF1 Scenario

In this scenario fission products are released into the lower

compartment both before and after vessel failure. However, the fission

products released from the primary system after vessel failure are released

through the reactor cavity water and thus are subject to scrubbing. The

fission products are released into the lower compartment throughout the entire

release period. From the lower compartment the fission products flow through

the ice condenser and into the upper compartment, where they are released to

the environment. The calculational procedure for this scenario is given in

Figure 5.5.

Table 53 summarizes the release of radionuclides to the containment

from the reactor coolant system during the in-vessel melting period, the puff

at the time of vessel failure, and during core-concrete interaction.

The size distribution of airborne particles in the upper compartment

is shown in Table 54, and the fraction of the core inventory released to the

environment from the upper compartment is listed in Table 5.5. Table 56

presents the locational distribution of each fission product after the scenario

is completed.

As can be seen in Table 56, the majority of the fission products

which are released from the core are retained in the primary system, the

cavity water, or the ice condenser. The integrated decontamination factor for

the ice condenser varies somewhat with the species but is about 4 As a

result of these factors, the amount of fission product released to the

environment is fairly low, about percent of the I, 4 percent of the Cs, and

1 percent of the Te inventories. The decontamination factor for the ice

condenser will be discussed further in Section 56.

TABLE 53. SUMMARY OF RELEASE TO CONTAINMENT FOR THE S3HF SEQUENCE

During During DuringGroup In-Vessel Puff Core-Concrete

Release Release AttackS3HF2/S3HF3 S3HFj S3HFI/S3HF2 S3HF3

.2999 1.4758E-04 3.46E-06 9.3423E-04 5.32E-03

Cs .2369 2.6318E-04 6.17E-06 1.4375E-03 7.94E-03

Pi 6.8057E-04 4.8690E-08 1.14E-09

TE 7.5049E-02 1.7069E-04 4.OOE-06 5.1263E-03 3.62E-02

SR 1.5180E-04 1.5863E-09 3.72E-11 5.8901E-03 3.30E-02

RU 2.5856E-07 4.2563E-13 9.97E-15 4.2590E-07 2.19E-06

LA 2.4039E-08 4.8395E-15 1.13E-16 2.7983E-04 1.60E-01

CE 0 0 0 2.1797E-04 1.23E-03

BA 2.7774E-03 8.5635E-08 2.01E-09 3.65OOE-03 2.06E-02

5-10

TRAP-MERGE VANESA

NA(Lower

Compartment)

ICEDF(Ice

Condenser)

NAUA(Upper

Compartment)

Environment

FIGURE 5.5. SCHEMATIC DIAGRAM SHOWING CONTAINMENTCALCULATION PROCEDURES FOR THE S3HFloS3HF2,o S3HF3.o AND TB SCENARIOS

TABLE 54. SIZE DISTRIBUTION OF AEROSOLS IN CONTAINMENT - S3HF1 SCENARIO

T I ME (fr) 7.000 7.501 B. 01 9.001 9.501 10.000 10.500 11-000 Is. 01 20.024

DENSITY(G/CN3) 3OE*OO 3 OOE+00 3.04E+oo 3. 15E+00 3.26E+00 3. 3SE400 3.43E+00 3.48E+00 3.62E+00 3.53E+00

PARTICLEDIAMETER(MICRONS)

5.001!-03 0. 0. 0. 0. 0 . 0 . 0 . 0 . 8.99E-23 0 .8.20E-03 0. 0. 0. 0. 0. 7.70E-26 2.47E-24 9.37E-23 9.41E-18 0.1.352-02 I.M-22 3.412-23 9.411-23 i.69E-29 1.091-2i 7.89E-20 I.OBE-18 1.72E-17 2.14E-13 0.2.21E-02 2.359-16 9.41E-19 3.6OE-17 1.44E-19 2.659-16 1.079-14 6.94E-14 5.34E-13 1.03E-09 2.OGE-iS3.62E-02 3.UE-12 $.oil-13 4.901!-12 1.599-14 7.GM-12 1.919-10 5.86E-lo 2.64E-09 9.91E-07 1.14E-Od5.95E-02 8.75E-00 2.479-00 2.449-06 2.84E-10 2.33E-06 3.881-07 6.78E-07 1.89E-06 1.80E-04 1.28E-05

7GE-02 1.6911-00 6.451-07 1.142-09 i.481-00 l.i9E-OG 8.64E-05 i.23E-04 2.20EE-04 8.69E=03 ft.48E-039. 6.59E-02 ;.57E-021.90E-Ol 4.991-05 3.94E-05 7.itE-04 4.091-04 5.08E-04 2.GSE-O3 3.72E-03 5.20E-032.63E-01 1.629-03 8.21E-04 6.71E-03 1.22E-02 7.54E-03 1.62E-02 2.36E-02 3.22E-02 2.16E-Ol 3.02E-Ol4.31E-01 1.23E-02 8. IOE-03 3.33E-02 8.82E-02 4.87E-02 4.73E-02 5.61E-02 7.37E-02 2.64E-01 4.32E-017.079-01 5.811-02 4.94E-02 9.63E-02 1.54E-01 1.02E-01 1.91E-Ol 1.45E-01 1.47E-01 1.6SE-01 1.64E-011. l6E+O0 I.Me-01 1.98E-Ol 1 3SE-01 1.63E-01 2.57E-01 2.80E-Ol 2.79E-Oi 2.73E-01 1.36E-01 3.OOE-021. 01+00 2.92E-01 3.00E-Ol 2.87E-01 2.31E-01 2.472-01 2.619-01 2.66E-Ol 2.61E-Ol 1.01E-Ol 1.11E-023. 09+00 8.32E-01 2.46E-01 3.392-01 2.53E-01 2.04E-Ol 1.84E-01 1.76E-01 1.64E-01 4.OGE-O2 2.93E-035. 12E+W 1.221-01 1.27E-01 1.239-01 8.811-02 0.989-02 9.34E-02 4.73E-02 4.07E-02 4.64E-03 1.56E-048.419400 I.Mg-02 1.672-02 1.906-02 I.OGE-02 7.03E-03 4.599-03 3.47E-03 2.53E-03 5.23E-05 3.23E-071. =E+01 1.4611-03 1. 132-03 8.14E-04 4.279-04 2.30E-04 9.92E-05 5.32E-05 2.81E-03 2.89E-08 1.75E-112.2W+Ol 9.869-09 2.239-05 1.099-05 3.9OE-OG I.WE-06 3.96E-07 1.24E-07 4.53E-08 1.48E-12 i.34E-093.711+01 4.932-07 8.78E-08 3.289-08 7.12E-09 2.48E-09 2.28E-10 5.02E-11 1.32E-ii 3.01E-10 4.15E-096.09t+01 5.409-10 7."E-11 2.12E-11 2.72E-i2 7.83E-13 2.719-14 4.07E-15 6.48E-ii 1.01E-09 1.39E-081.001!+02 1.23E-13 1.32E-14 2.98E-13 4.OSE-11 3.99E-ii 9.48E-11 1.26E-10 1.59E-10 2.47E-09 3.40E-08

TABLE 5. FRACTION OF CORE IVENTORY RELEASED FROM COMMENT - S3HF1 SCENARIO

Tim FISSION PRODUCT(Hit) I Cs Pt TE SR PU LA CE BA PE Tot

7.000 3.0511-02 2.52t-oa 7. 3E 06 8.032-03 1.982-05 2. 87E 08 2.64E-09 T. 2GE- 16 3.08f-04 9. OSE-04 2.2419-os7.501 3.639-02 2.939-02 8.852-06 0. "E-03 1.95E_05 3.33E-08 2.06E_09 S. 13C - 12 3.59E-04 3.28E-03 2.752-032.501 3.679-02 3.04E-02 9.209-06 9.69E-03 3.40E-05 3.47E-00 i.IIE-06 6.56E-07 3.76E-04 5.05E-ol 2.98E-039.001 4.122-02 3.4it-02 1.03E-04 i.09E_02 1.632-04 3.812-09 1.04E-05 9.9st-06 5.021-04 6.499+00 3.261-03

1:.901 4.WE-02 3.8JE-02 1.159-04 1.22E-02 5.33E-04 4. ne -Os 2.61E-05 1.98E_05 7.55E-04 2. M+ 1 5.97f-os.000 5.09E-02 4.222-02 1.27E-04 i.36E-02 9.991-04 4. "E -08 4.581-05 3.62E-05 i.O$E-03 5.26f*ol 9.921E-03

10.5w 6.129-02 4.2ff-02 i.289-04 1.37E-02 1.03E-03 S.012-06 4.70E-05 3.75t-05 i.IIE-03 S.IM+01 1.00f-0211.000 5.14E-02 4.279-02 1.292-04 1.3$E-02 1.079-03 5.05E-08 4.03E-08 3.86E_05 1.13E-03 5.73E+01 1.04E-0219.501 6.191-02 4.31E-02 i.30E-04 1.40E-02 1.12E-03 5.wt-Oe 5.07E-05 4.08E-05 1.17E-03 1.139-02

5.19E-02 1.12E-03 6.67E-06 5.07E-05 4.0GE-05 1.19E-03 6.55f+of 1.13E-0220.024 4.31E-02 1.30E-04 1.41E-02 E*01

TABLE 56. DISTRIBUTION OF FISSION PRODUCTS BY GROUP - S3W1 SCENARIO

Cavity Lower Ice UpperSpecies RCS Water Melt Compartment Bed Compartment Environment

I 0.67 2.9 x 1-2 0 3.8 x 1-2 0.18 9.2 x 10-3 5.2 x 1-2

Cs 0.73 3.1 x 1-2 0 3.0 x 10-2 0.14 7.4 x 10-3 4.3 x 10-2

Te 0.77 5.9 12 9.3x 10-2 1.2 x 10-2 4.7 x 1-2 2.4 x 10-3 1.4 x 10-2 Ln

Sr 4.9 x10-4 0.16 0.83 1.7 x 10-3 3.6 x 10-3 1.0 x 10-4 1.1 x 10-3

Ru 8.0 x10-7 2.0 x 10-6 1.0 1.2 x 10-7 3.5 x 10-7 1.9 x 10-8 6.9 x 10-8a 0 -IA-A A - I -1 1% nn , -- -, ._r . ^ q--A . - .-- c I .-

U..L AIV Q.'t A IV - V. 10 D.D x IV i.0 x IV -T 4. 1jxIu-V D.1 x lu-j

Ce 0 6.2 x 10-3 0.99 5.6 x 10-5 1.4 x 10-4 3.8 x 10-6 4.1 x 10-5

Ba 9.0 x10-3 9.6 x 1-2 0.89 1.4 x 10-3 3.7 x 10-3 1.5 x 10-4 1.2 x 10-3

Tr 0 0.98 0 2.2 x 10-2 3.5 x 10-2 1.5 x 10-3 1.1 x 10-2

5-14

5.1.3 Release and Transport in Containmentfor SiHF2 Scenario


compartment both before and after vessel failure. Unlike S3HF1, the fission

products released from the primary system after vessel failure are released

directly into the lower containment. The fission products are released into

the lower compartment throughout the entire release period. From the lower

compartment the fission products flow through the ice condenser and into the

upper compartment, where they are released to the environment. The

calculational procedure for this scenario is again given in Figure 5.5.


is shown in Table 57, and the fraction of the core inventory released to the

environment from the upper compartment is listed in Table 5.8. Table 59

presents the locational distribution of each fission product after the scenario

is completed.

As can be seen in Table 59, the majority of the fission products

which are released from the core are retained in the primary system, the

cavity water, or the ice condenser. The integrated decontamination factor for

the ice condenser varies for the various fission product groups, but is

generally about 3 to 5. As a result of these factors, the amount of fission

product released to the environment is fairly low, about 4 percent of the I,

3 percent of the Cs, and percent of the Te inventories.

5.1.4 Release and Transport in Containmentfor SqHFq Scenario


compartment both before and after vessel failure. From the lower compartment

the fission products flow through the ice condenser and into the upper

compartment. The fission products are released to the environment from the

upper compartment. The calculational procedure for this scenario is also

given in Figure 5.5.


is shown in Table 5.10, and the fraction of the core inventory released from

the upper compartment is listed in Table 5.11. Table 512 provides the

TABLE 57. SIZE DISTRIBUTION OF AEROSOLS IN CONTAINMENT - S3HF2 SCENARIO

TIME (r) 7.000 ?.GM 8.900 9.000 S. SW io.001 10.500 11.000 15.900 20.026

DENSITY (G/CN3) 3.ooE+oo 3.009+00 2. 06E+00 3.21E+00 3.30E+00 3.381!+00 3.43E+00 3.47E+00 3.62E+00 3.54E+oo

PMTICLEDIAWTER(Mir, la NO

S.00E-03 0. 0. 0. 0. 0. 0. 0. 0. 6. 42E-23 0 .8 2K-03 2.192-29 0. 0. 0. 0. 3.789-26 5.95E-24 6. 79E-23 6.96E-le 0 .1. 5E-02 1.661-17 9.9311-23 1.520-22 3.572-29 1. 3E-21 3.$OE-" 2.17E-14 1. 23E- 17 1.659-13 a 2.211-02 3.891-12 9.161-19 4.141-17 2.811-19 2.929-19 4.919-15 1.04E-13 3.761-13 8.07E-10 I.SSE-163.028-02 2."2-10 0.792-13 6.009-12 2.34E- 14 7.38E-12 8.339-11 6.57E-10 1.851-09 7.70L-07 1. 2SE 09B. ff-02 3AW-08 2.239-09 3.3u-" 3.87E-10 1.97E-04 1 75E-V? 6.08E-07 1.33E-06 1. 37E 04 1. ISE-059.71SE-02 3.079-08 7.14E-07 1.049-06 1.632-09 5.94E-09 4.221E-05 S. 3SE -05 1.54E-04 S.OVE-03 2.21E-031.190E-ol 1. ISE-04 5.01E-06 1.05[-03 4.OOE-04 3.29E-04 1.372-03 2.60E-03 3. SSE 03 4.98E-02 5.07E-022.63E-01 1.81E-03 i.041-02 1.279-02 1.15E-02 5.919-03 8.87E-03 1.57E-02 2.14E-02 1.72E-01 2.84E-014.3if-01 1.38E-02 9.729-03 4. am-02 2.2SE-0 4.141-02 3-241-02 3z�OE-0-2 4afff-02 2.331-01 4.29E-017.072-01 6.249-02 5.239-02 8.199-02 1.87E-01 i.679-01 1.41E-01 1.2SE-01 1.22E-01 1.62E-01 1.74E-01I.IGE+00 1.752-01 1.871-01 1.539-01 2.009-01 2.92E-01 3.16E-01 3.0�E-Oi 3.OOE-01 1.71E-01 3.92E-021.9"+00 2.919-01 3.01E-01 2.79E-01 2.24E-01 2.57E-01 2.91r-ol 3.04E-01 3.07E-01 1.49E-01 1.849-023.i2E+00 3.22E-01 3.341-01 3.079-01 2.14E-01 1.79E-01 1.19SE-01 1.64E-01 1.60E-01 5.19E-02 4.17E-039.12E+00 1.189-01 1.20E-01 1.059-01 6.8se-02 6.24E-02 4.21E-02 3.89E-02 3.51E-02 5.11E-03 1.84E-044.40+00 1.47E-02 1.46E-02 1.2011-02 7.09E-03 4.912-03 3.371-03 2.74E-03 2.151-03 5.73E-05 3.71E-071.309+01 1.259-03 8.87E-04 4.9711-04 2.321-04 1.3�E-04 7.19E-05 4.50E-05 2.70E-05 3.569-08 1.97E-112.269+01 4.471-05 1.449-05 5.029-08 I.SSE-08 S.60E-07 3.07F-07 1.3SE-07 5.8at-os 1.99E-12 I.SSE-093.71E+01 2.719-07 4. 209-08 1.10-08 2.49E-09 1.19E-09 2.639-10 7.72E-11 2.3SE-11 3.39E-10 5.68E-096.091+01 2.279-10 2.51SE-11 SASE-12 7.82E-13 3.64E-13 4.3it-14 8.90E-is 4.57E-11 1.10E-09 1.84E-081.001!+02 3.92E-14 3.34E-15 4.171-11 4.09E-ii 3.502-11 9.00E-11 8.92f-11 i.12E-10 2.97E-09 4.48E-08

TABLE 5.8. FRACTION OF CORE INVENTORY RELEASED FROM CONTAINMENT - S3HF2 SCENARIO

lim fission PRooUclr GM"(MR) I Cs pi TE SR Ru LA CE DA PE TR

1.000 2.779-02 2.2911-02 S."t-06 7.211-03 1.53E-00 2.629-06 2.409-09 7.64E-16 2.799-04 9.29E-04 9.539-027. DM 2.849-02 2.39E-02 7.142-06 7.511-08 1.57E-05 2.699-08 2.47E-09 1.90E-14 2.86E-04 1.009-03 9.921-028.900 2.92K-02 2.431-02 7.24K-05 7.739-00 3.15E-05 2.779-08 1.231-06 7.3GE-07 2.03E-04 9.59E-01 1.051-01

000 3.28E-02 2.722-02 8.221-06 8.989-03 2.49E-04 3.101-08 1.51E-05 1.01E-05 4.961-04 9.26E+oo 1.33E-01:-Boo 3.67E-02 2."1-02 8.959-09 9.51E-os 5.54E-04 3.419-08 3.03E-05 2.18E-06 6.62f-04 2.449+01 i.57E-01

10:001 7.97E-02 3.30L-02 9.952-05 1.072-02 1.09E-03 3.919-09 5.33E-05 4.09E-05 1.02E-03 5.52E+01 1.91t-ol10. gm 4.046-02 3.31BE-02 1.011-04 1.0se-02 i.IBE-03 4.009-09 6.73E-013 4.43E-05 1.09E-03 O.iSE+Oi 1.992-0111.000 4.08E-02 3.40E-02 i.02E-04 i.liE-02 1.25E-03 4.08E-08 6.029-05 4.67E-05 1.13E-03 6.63E+01 2.0*E-0115.800 4.18E-02 3.49E-02 1.05E-04 1.15E-02 1.40E-03 4.939-08 6.63E-05 5.20E-05 1.24E-03 7.961+01 2.089-0120.026 4.18E-02 3."E-02 I-OSE-04 1.16E-02 1.40E-03 7.02E-08 6.69E-05 5.21E-05 1.24E-03 8.33f+01 2.0BE-01

TABLE 5 9 DISTRIBUTION OF FISSION PRODUCTS BY GROUP S3HF2 SCENARIO


I 0.67 2.9 x 1-2 0 4.2 x 1-2 0.19 7.5 x 10-3 4.2 x 1-2

Cs 0.73 3.1 x 1-2 0 3.2 x 10-2 0.15 6.0 x 10-3 3.4 x 10-2

Te 0.77 5.9 x 1-2 9.3 x 10-2 1.2 x 10-2 4.9 x 1-2 2.0 x 10-3 1.2 x 10-2

Sr 4.9 x 10-4 0.16 0.83 1.6 x 10-3 2.7 x 10-3 1.4 x 10-4 1.4 x 10-3

Ru 8.0 x 10-7 2.0 x 10-6 1.0 1.3 x 10-7 3.5 x 10-7 1.1 x 10-8 7.0 x 10-8

La 8.1 x 10-8 8.4 x 10-3 0.98 5.8 x jo-5 1.4 x in-4 7 n 10-6 r, r in-5

Ce 0 6.2 x 10-3 0.99 5.1 x 10-5 1.o x 10-4 5.4 x 10-6 5.2 x 10-5

Ba. 9.0 x 10-3 9.6 x 1-2 0.89 1.4 x 10-3 3.3 x 10-3 1.5 x 10-4 1.2 x 10-3

Tr 0 0 0 0.31 0.38 2.9 x 10-2 0.21

TABLE 5.10. SIZE DISTRIBUTION OF AEROSOLS IN CONTAINMENT - S3HF3 SCENARIO

TIME (r) 7.001 7.500 8.502 9.001 9 SW 10.000 10.500 11.000 15.500 20.002

DENSITY G/CN3) 3.OOE+00 3.OOE+00 3. iE+00 3.32E+00 3.51E+00 3.61E+00 3.67E+00 3.72E+00 3.79E+00 4.02E+00


S. OOE-03 0 0. 0. 0. 7.529-26 2.77E-24 9. i7E-21 I.OSE-19 1. 76E- 17 0 8.20E-03 1.24E-23 0. 0. 0. 7.59E-21 1.19E-19 7.77E-17 5.72E-16 2.14E-14 0.1.35E-02 1.84E-17 0. 0. 0. 2.22E-16 1.40E-15 2.20E-13 1.04E-12 1.1GE-11 O.2.21E-02 4.OOE-13 3.99E-19 4.02E-24 $.ODE-23 2.28E-12 4.50E-12 2.21E-10 6.64E-10 2.83E-09 0.3.921-02 2.75E-10 7.51E-13 9.03E-18 3.28E-17 3.68E-09 3.83E-00 6.72E-08 i.42E-07 2.98E-07 0.5.95E-02 3.81E-Ol 2.44E-09 8.46E-12 2.54E-12 1.02E-06 6.39E-07 5.7#E-08 9.80E-06 1.31E-05 i.09E-149.76E-02 3.20E-06 7.62E-07 5.71E-08 2.23E-08 6.07E-05 4.97E-05 1.62E-04 2.37E-04 2.52E-04 7.44E-09I.GOE-01 1.22E-04 5.21E-05 2.74E-05 2.029-05 1.09E-03 9.38E-04 1.45E-03 2.48E-03 2.50E-03 5.54E-082.63E-01 1.84E-03 1.079-03 2.03E-03 2.07E-03 8.93E-03 7.45E-03 I.OSE-02 1.42E-02 1.57E-02 3.67E-044.31E-01 1.39E-02 9.89E-03 2,72E-02 3.30E-02 3.99E-02 3.2GE-02 3.43E-02 4.92E-02 6.93E-02 6.1GE-037.07E-01 6.27E-02 5.27E-02 9.59E-02 1.39E-01 1.20E-01 1.02E-01 9.87E-02 1.10E-Ol 1.88E-Ol 5.47E-021.19E+00 1.75E-Ol 1.68E-Ol 1.85E-Ol 2.22E-01 2.84E-01 2.66E-Oi 2.45E-Ol 2.259-01 2.81E-Ol 2.60E-011.90E+00 2.911-01 3.OOE-O1 2.81E-Ol 2.94E-01 3.46E-01 3.WE-01 3.66E-01 3.54E-01 2.88E-Ol 4.54E-013.12E+00 3.22E-01 3.33E-01 2.96E-01 1.88E-01 1.04E-01 1.86E-Ol 1.97E-01 1.95E-Ol 1.37E-01 2.OOE-015.129+00 1.10E-Ol 1.20LI-Of I.Offf-of 5.48E-02 3.29E-02 3.63E-02 3.83E-02 3.702-02 1.79E-02 2.36E-028.4iE+00 i.499-02 1.451-02 1.14E-02 5.88E-03 2.79E-03 2.87E-03 2.47E-03 2.57E-03 G.WE-04 9.31E-041.3$E+01 1.22E-03 8.03E-04 4.GOE-04 2.16E-04 9.43E-05 7.4$E-05 6.74E-05 5.15E-05 3.49E-05 1.07E-052.26E+01 4.28E-05 1.38E-05 4.50E-06 2.07E-06 6.65E-07 4.82E-07 3.43E-07 2.33E-07 4.50E-09 3.0GE-083.71E+01 2.5&E-07 4.OOE-08 9.809-09 4.70E-09 1.13E-09 7.OOE-10 4.79E-10 2.24E-10 1.25E-12 2.02E-116.09E+01 2.22E-10 2.3GE-11 4.74E-12 2.48E-12 3.98E-13 2.27E-13 1.30E-13 4.efE-14 4.73E-11 2.49E-11I.OOE+02 3.51E-14 3.02E-15 3.SSE-11 2.01E-11 2.72E-11 2.8GE-11 3.5GE-11 4.57E-11 1.039-10 5.45E-li

TABLE 5.11. FRACTION OF CORE INVENTORY RELEASED FROM CONTAINMENT - S3HF3 SCENARIO

Tin FISSION ROOMI on"(HP0 Cs pi TE itu LA CE BA Pe Tlt

7.001 2.72E-02 2.259-02 6.429-05 7.121-03 I-WE-06 2.57E-08 2.3GE-09 3.83E-15 2.74E-04 4.73E-03 7.679-02T. Om 2. IOE -02 2.329-02 7.01E-05 7.329-02 1.54t-os 2.64E-08 2.42E-09 8.021-14 2.812-04 S. 2E-03 8.022-029.902 3.09E-02 2.561-02 7.75E-05 8.159-03 4.38E-05 2.92E-00 2.64E-06 1. 23E-06 3.25E-04 9. 37E -0 I 9. 84E-02

001 3.459-02 2.8ff-02 2.02E-05 9.139-03 4.189-04 3.26E-08 2.52E-06 1. 70E-05 5.989-04 1.98E+01 1. ISE-01:-moo 4.29E-02 3.572-02 1.01YE-04 1. IOE-02 2.02E-03 4.429-08 1. 30L-04 1.09E-04 2.21E-03 1.61E+02 1. 63E -0 I

10.000 4.34E-02 3.03E-02 1.08E-04 1.219-02 2. 2SE-03 4. BIE-08 1. 39E-04 1. 7E-04 2.37E-03 1.78f+02 1.98E-ol10.000 4.40E-02 3.UE-02 1.1011-04 1. 24E-02 3.51E-03 4.79E-08 1.49E-04 1. ZOE-04 2.53E-03 1. 97E+02 1. 70E-Ol11.000 4.459-02 3.721!-02 1.119-04 1.27E-02 3. 70E -09 4.9SE-08 I.WE-04 1.32E-04 2.68E-03 2.14E+02 1.72E-0116.900 4.519-02 3.792-02 1.122-04 1. 33K-02 3.98E-03 7.27E-08 1.661-04 1.429-04 2.85E-03 2. "E+02 1.76E-0120. oo2 4.549-02 3.81E-02 1.139-04 1.�W-02 4. 1 IE-03 5.09E-07 1.71E-04 1.46E-04 3.02E-03 3.36E+02 1.7GE-0i

TABLE 512. DISTRIBUTION OF FISSION PRODUCTS BY GROUP - S3HF3 SCENARIO


I 0.67 2.4 x 1-2 0 4.6 x 1-2 0.19 7.5 x 10-3 4.5 x 1-2

Cs 0.73 2.3 x 1-2 0 3.6 x 10-2 0.15 6.1 x 10-3 3.8 x 10-2

Te 0.77 2.4 x 1-2 8.6 x 10-2 2.1 x 10-2 6.4 x 1-2 2.8 x 10-3 1.6 x 10-2

Sr 4.9 x 10-4 0.14 0.83 1.8 x 10-2 1.1 x 10-2 4.4 x 10-4 4.1 x 10-3

Ru 8.0 x 10-7 6.5 x 10-7 1.0 7.2 x 10-7 1.6 x 10-6 1.2 x 10-7 5.1 x 10-7

La 8.1 10-8 4.8 x 10-3 0.98 9.7 x 10-4 4.4 x 10-4 1.8 x 10-5 1.7 x 10-4 CD

Ce 0 5.2 x 10-3 0.99 7.6 x 10-4 4.2 x 10-4 1.7 x 10-5 1.5 x 10-4

Ba 9.0 x 10-3 7.9 x 1-2 0.89 1.1 x 10-2 8.9 x 10-3 3.6 x 10-4 3.0 x 10-3

Tr 0 0 0 0.43 0.38 2.5 x 10-2 0.18

5-21

locational distribution of each fission product group at the end of the

scenario.

The predicted releases to the environment for S3HF3 fall between

those for S3HF1 and S3HF2 for I and Cs and are slightly higher for the other

fission product groups. This is expected because the cavity is dry before the

end of the ex-vessel release. The majority of the increase in fission

products leaving the reactor cavity has been dposited in either the lower

compartment or the ice condenser. In this scenario 19 percent of the I,

15 percent of the Cs, and 6 percent of the Te re retained in the ice

condenser. The ICEDF code indicates that there is very little difference in

the integrated decontamination factors predicted for the in-vessel release

period and the ex-vessel release period, the frmer being somewhat greater

than four and the latter a little less than fr.

5.2 TB Sequence

5.2.1 Release and Transport in the RCS

The TB sequence does not differ significantly from the SOF sequence

in the evolution of the RCS events because of he similar thermal-hydraulic

conditions. Retention of fission products in he RCS is therefore expected to

be similar for these scenarios as well. This is indeed the case as can be

seen from selected results of calculations in Tables 513 and Table 514 and

Figure 56 through Figure 59 which mirror the results for the S3HF1 scenario.

5.2.2 Release and Transport in Containment

In this scenario the fission products from both the primary system

and the reactor cavity are released into the lower compartment. The fission

product flow path is from the lower compartment to ice condenser, ice

condenser to upper compartment, and, after containment fails, upper

compartment to environment. The calculational procedure shown in Figure 5.5

was used for this scenario also.

Table 5.15 summarizes the release of radionuclides from the RCS

during core meltdown, the puff at the time of essel failure and the release

during core-concrete interactions.

TABLE 513. MASSES OF DOMINANT SPECIES RELEASED FROM FUEL (TOTAL)AND RETAINED ON RCS STRUCTURES (RET) AS A FUNCTION OFTIME -- TB SEQUENCE

cSI CSOH TE AEROSOLTIME RET TOTAL RET TOTAL RET TOTAL RET TOTAL(S) (KG) (KG) (KG) (KG) (KG) (KG) (KG) (KG)

19782. .2 1.7 3.2 i3.4 .2 .7 4.5 31.8

19923. 1.2 3.8 12.0 26.0 .7 1.4 17.7 46.9

20063. 2.8 6.4 23.5 4i.5 1.4 2.4 31.9 64.0

20198. 4.6 9.2 38.3 57.8 2.2 3.5 46.9 84.9

20337. 6.7 12.3 50.5 75.7 3.3 5.0 65.8 iO7.7

20475. 8.7 15.3 65.1 93.7 4.7 7.0 85.9 129.7

20814. We le.i 79.0 110.3 6.5 9.8 106.2 152.7

20753. 13.4 20.5 90.8 124.3 8.9 12.6 127.3 i76.9

20891. 16.0 22.6 iO6.2 i36.6 11.7 15.3 i5l.4 200.5

21030. ig.3 24.4 122.4 i47.7 is.0 i8.8 178.8 224.9

21167. 20.0 28.0 133.2 157.1 18.7 22.5 200.4 250.2

21306. 21.1 28.1 141.9 171.0 22.4 24.9 232.3 309.1

21449. 21.4 29.9 i43.5 i82.3 23.9 25.9 250.9 329.4

2iS84. 21.1 30.0 142.7 183.4 24.3 26.2 254.7 342.6

21721. 20.9 30.1 i4i.5 183.8 24.6 26.5 256.0 351.4

21863. 20.7 30.1 140.7 184.1 24.6 26.6 256.4 355.2

21999. 20.7 30.2 140.2 i84.2 24.4 26.6 256.5 355.7

22i3g. 20.6 30.2 139.9 184.2 24.4 26.6 256.6 356.0

22279. 20.6 30.2 i39.7 i84.3 24.3 26.6 256.6 356.3

22419. 2O.'6 30.2 139.6 i84.3 24.3 26.6 256.6 356.3

TABLE 514. MASSES OF RADIONUCLIDE RELEASED FROM FUELAND RETAINED ON RCS (BY GROUP) -- TB SEQUENCE


1 14.7 lo.i

Cs 178.9 134.4

TE 26.6 24.3 C"

SR .0 .0

RU .0 .0

LA .0 .0

NG 336.9 .0

CE .0 .0

BA .9 .7

TABLE 5.15. SUMMARY OF RELEASE TO CONTAINMENTFOR THE TB SEQUENCE

DURING DURING DURINGIN-VESSEL PUFF CORE-CONCRETE

GROUP RELEASE RELEASE ATTACK

1 .3070 i.2532E-04 5.322SE-03

Cs .2405 2.3820E-04 7.83SOE-03

pi 7.3313E-04 3.198SE-07

TE 7.SIBOE-02 2.226SE-04 3.5622E-02

SR 1.5758E-04 i.8207E-09 3.2427E-02

RU 2.680SE-07 3.77i6E-i3 3.4072E-08

LA 3.50i7E-00 4.319SE-12 1.5930E-03

CE 0. 0. 1.246iE-03

BA 2.8929E-03 9.5601E-oe 2.003SE-02

35

30

25

20

15 Legend ulFUEL

10 CORE

UPPER PLENUM

HOT LEG

STEAM GENERATOR

019500 20600 20500 21�00 21900 22600 22500

TIME (sec)

FIGURE 56. MASS OF CsI RELEASED FROM INDICATED RCS COMPONENT AS A FUNCTION OFTIME - TB SEQUENCE

200-

150

loo

LegendFUEL

CORE50 UPPER PLENUM

HOT LEG

STEAM GENERATOR

01 -19500 20000 20;00 21000 21400 22600 22500

TIME (sec)

FIGURE 5.7. MASS OF CsOH RELEASED FROM INDICATED RCS COMPONENT AS A FUNCTION OFTIME - TB SEQUENCE

30

25

20

15

Legend 14

FUEL10

Wn R E

UPPER PLENUM

5 HOT LEG

STEAM GENERATOR

04500 20000 20500 21000 21900 22600 22500

TIME (sec)

FIGURE 5.8. MASS OF TE RELEASED FROM INDICATED RCS COMPONENT AS A FUNCTION OFTIME - TB SEQUENCE

400

350

300

250

200

00Legend150 FUEL

CORE

100 "-Z UPPER PLENUM

HOT LEG

50 STEAM GENERATOR

019500 20000 20;00 21600 21500 22600 22500

TIME (sec)

FIGURE 59. 14ASS OF AEROSOL RELEASED FROM INDICATED RCS COMPONENT AS A FUNCTION OFTIME - TB SEQUENCE

5-29

The size distribution of airborne particles in the uper compartment

is shown in Table 516, and the fraction of the ore inventory released from

the upper compartment is listed in Table 517. 'Table 5.18 presents the

distribution of each fission product group in various locations within the

plant at the end of the accident.

As can be seen in Table 5.18, relatively little of the fission

product inventory is released to the environment. The maximum releases are

for the I and Cs groups at about 2 percent each. As seen in the previous

scenarios, the majority of the fission products eleased from the core are

found to remain in either the primary system or the reactor cavity water. Of

the fission products released into the containment, there is a significant

amount retained in the lower compartment and captured by the ice condenser.

The integrated ice condenser decontamination factors during the in-vessel phase

of the accident are between and 7 during the ex-vessel phase the overall

decontamination factors vary between 3 and 8. Te higher ice bed

decontamination factors seen in this case are the result of higher steam

fractions for the flow through the ice condenser.

TABLE 516. SIZE DISTRIBUTION OF AEROSOLS IN CONTAINMENT - TB SCENARIO

TIME 010 7.003 7.5of 0.500 9.000 9. SW 10.000 10.500 11.000 15.500 18.004

DENSITY (ra/CN3) 3001!+00 3.09E+00 3.40E+00 3.$OE+00 3. 6SE+OO 3.GBE+00 3.71E+00 3.73E+00 3. 89E+00 4. 05E+00


5.00E-02 0. 0. 2.41E-24 5.57E-23 8.55E-21 8.692-20 1.50E-18 5.08E-18 8.20E-17 0.5.209-03 0. D. 1.4st-io 1.32E-18 6.80E-17 4.40E-16 4.28E-15 i-17E-14 7.57E-i4 0.1.382-02 2."1-23 0. 2.029-15 9.37E-15 1.82E-13 7.951-i3 4.65E-12 I-OGE-il 3.82E-ii 0. Ln2.219-02 1.20L-14 0.579-18 7.25E-12 1.999-11 1.03E-10 5.12E-10 1.86E-09 3.68E-09 8.55E-09 0. ILI3.92E-02 2.51E-10 1.07E-11 6.29E-09 1.31E-08 4.50E-08 l.iOE-07 2.62E-07 4.63E-07 8.26E-07 1.77E-19 CD5.95L-02 9.62E-08 2.20E-Ol 1.29E-06 2.41E-06 4.OGE-OG 7.57E-08 1.33E-05 2.05E-05 3.37E-05 7.15E-ii9.762-02 6-3OE-OG 2.94E-06 6.51E-05 1.23E-04 1.32E-04 i.SBE-04 2.61E-04 3.59E-04 6.15E-04 7.88E-07I.GOE-01 1.54E-04 1.55E-04 9.toff-04 2.07E-03 1.79E-03 2.01E-03 2.439-03 3.ME-03 5.92E-03 1.20E-042.43E-01 1.79E-03 4.341-03 6.55E-03 i.45E-02 1.26E-02 1.25E-02 1.35E-02 1.56E-02 3.69E-02 3.OOE-034.31E-01 1.18E-02 4.09E-02 4.42E-02 5.46E-02 5.23E-02 5.15E-02 5.17E-02 5.49E-02 1.55E-01 3.36E-027.071-01 9.229-02 1.13E-01 1.51E-01 1.41E-01 1.41E-01 1.42E-01 1.41E-01 1.41E-01 3.46E-01 1.95E-01I.IQE400 1.512-01 1.61E-Ol 2."E-01 2.81E-Ol 2.79E-01 2.78E-01 2.77E-01 2.75E-01 3.10E-Ol 4.35E-011. 900+00 2-49E-Ol 2.34E-01 3.30E-01 3.33E-01 3.3GE-O1 3.36E-01 3.36E-01 3.359-01 1.18E-01 2.82E-012.12a+oo 2.98E-Oi 2.54E-01 1.43E-01 1.49E-01 1.83E-01 1.54E-01 1.64E-01 1.53E-01 2.45E-02 4.66E-029.129+00 1.89E-Ol 1.54E-01 2.58E-02 2.412-02 2.38E-02 2.32E-02 2.25E-02 2.15E-02 1.72E-03 2.32E-038.41E+00 5.35E-02 3.74E-02 1.949-03 1.40E-03 1.27E-03 1.iff-03 1.01E-03 8.83E-04 2.t7E-05 3.04E-05i.S$E+01 3-59E-03 1.89E-03 4.87E-05 2.29E-06 1.80E-05 1.429-06 1.10E-09 8.33E-06 4.28E-08 I-OiE-072.269+01 3.43E-05 1.25E-06 3.089-07 8.65E-08 5.02E-08 3.80E-06 2.529-08 I.GOE-08 1.87E-11 7.99E-li2.719+01 8.699-08 1.63E-os 4.951-10 0.63E-11 3.73E-11 2.iSE-11 i.20E-ii 6.38E-12 3.20E-15 1.43E-i4G.O9E+O1 1.90E-11 4.61E-12 1.72E-13 i.07E-14 5.35E-15 2.64E-15 3.38E-11 4.02E-11 7.17E-ii 4.53E-11I.O*E+02 9.49E-11 8.65E-li 3.53E-11 4.70E-11 5.1SE-11 5-98E-11 7.OSE-11 8.40E-11 1.50E-10 9.46E-11

TABLE 517. FRACTION OF CORE INVENTORY RELEASED FROM CONTAINMENT - TB SCENARIO

T114E FISSION PRODUCT an"(M) I Cs pi TE SR pti LA CE OA PE TR

7.003 I-IIE-02 9.83E-03 2.72E-05 2.98E-02 8.94E-06 1.01E-Os 1.33E-09 3.91E-M 1.09E-04 1.16E-03 2.21E-b27.501 I-ISE-02 9.39E-03 2.8911-05 2.86E-03 1.38E-05 1.07E-08 0.37E-07 3.57E-07 1.20E-04 2.GiE-01 2.12 02:.600 ;.:4:-02 1.51E-02 4.41E-05 6.28E-03 4.37E-03 2.14E-00 2.03E-04 1.66E-04 2.73E-03 2.71E+02 1. O:-Ol

, 000 . 5 02 1.62E-02 4.06E-05 6.22E-03 5.92E-03 2.98E-Ol 2.63E-04 2.21E-04 3.70E-03 4.OiE+02 1.24E-019.500 1.96E-02 1.931-02 4.67E-05 0.321-03 6.02E-03 3.08E-08 2.67E-04 2.24E-04 3.76E-03 4.11E+02 1.252-01

10.000 1.969-02 1.04E-02 4.98E-Of 6.41E-03 O.OGE-03 3.17E-Ol 2.69E-04 2.27E-04 3.80E-03 4.19E+02 1.26E-Oi10.500 1.971-02 1.64E-02 4.68E-05 6.491-03 6.14E-03 3.30E-08 2.71E-04 2.28E-04 3.84E-03 4.25E+02 1.26E-0111.000 1.97E-02 i.ME-02 4.69E-05 9.672-03 G.ISE-03 3.53E-08 2.73E-04 2.30E-04 3.8GE-O3 4.31E+02 1.27E-0118.500 1.98E-02 I.ME-02 4.70E-05 7.OOZ-O3 6.26E-03 7.89E-08 2.76E-04 2.33E-04 3.93E-03 4.49E+02 1.27E-0118.004 1.98E-02 1.66E-02 4.70E-05 7.80E-03 6.28E-03 2.17E-07 2.76E-04 2.33E-04 3.97E-03 4.75E+02 1.27E-01

TABLE 5.18. DISTRIBUTION OF FISSION PRODUCTS BY GROUP - TB SCENARIO


1 0.66 2.5 x 1-2 0 8.0 12 0.21 2.9 x 10-3 2.0 x 1-2

Cs 0.73 2.4 x 1-2 0 5.8 x 10-2 0.18 2.3 x 10-3 1.7 x 10-2-2 -2 -2 7.9 x 1-2 9.4 x 10-4

Te 0.76 3.9 x 1 8.6 x 10 2.2 x 10 7.8 x 10-

Sr 4.8 x 10-4 0.14 0.83 4.8 x 10-3 2.2 x 10-2 2.1 x 10-4 6.3 x 10-3

Ru 8.0 x10-7 6.5 x 10-7 1.0 4.1 x 10-7 2.6 x 10-6 2.7 x 10-8 2.2 x 10-7

La 7.6 x10-8 7.2 x 10-3 0.99 2.9 x 10-4 1.0 x 10-3 1.1 x 10-5 2.8 x 10-4

Ce 0 5.4 x 10-3 0.99 1.9 x 10-4 8.2 x 10-4 8.0 x 10-6 2.3 x 10-4

Ba 8.6 x10-3 7.8 x 1-2 0.89 3.5 x 10-3 1.5 x 10-2 1.6 x 10-4 4.0 x 10-3

Tr 0 0 0 0.35 0.49 1.7 x 10-2 0.13

5-33

5.3 TMLU-SGTR Sequence

The analysis of the TMLU-SGTR sequence is focused on the behavior of

fission products in the reactor coolant system and in the secondary side of

the steam generator; the containment aspects were not addressed for this

sequence. For the present purposes it was assumed that the events associated

with core slumping lead to the rupture of the steam generator tubes, with the

release of steam, hydrogen, as well as fission roduct aerosols to the

secondary side of the steam generators. The steam generator secondary was

assumed to be maintained at 1100 psia by the operation of the atmospheric

steam dump valves. The flow through the steam enerators was assumed to cease

with the failure of the reactor vessel head and depressurization of the primary

system.

The analysis of this sequence involved two separate TRAP-MERGE runs.

The first evaluated the transport and deposition of the fission products

within the primary coolant system and defined te releases to the secondary

side of the steam generator. The second TRAP-MERGE analysis evaluated

transport and deposition in the secondary side f the steam generator. It was

not possible to treat both parts of the problem in a single run since the

pressures in the primary system and the steam gnerator secondary are not the

same. The initial TRAP-MERGE analysis was essentially the same as a typical

analysis of this type, except for the need to distinguish between the releases

to the containment and those to the steam generator secondary. The analysis

of fission product transport in the steam generator secondary side required

some additional manipulation of the MARCH thermal-hydraulic data to take into

account the depressurization of the flow through the ruptured steam generator

tubes. (Normally the MARCH data is used directly as input into TRAP-MERGE.)

Table 519 summarizes the results of the TRAP-MERGE analysis for the

reactor coolant system which defined the input to the steam generator

secondary. The behavior of the fission products in the steam generator

secondary is summarized in Tables 520 and 521 and Figures 5.10 through 513.

The deposition of all the species considered is seen to increase monotonically

with time in the steam generator. The principal deposition is predicted to

take place on the surface of the steam generator tubes and the steam dryers.

The releases to the environment from the secondary side of the steam generator,

5 34

TABLE 519. SUMMARY OF PRIMARY COOLANT SYSTEM FISSION PRODUCTBEHAVIOR FOR TLU-SGTR

Total Core Suspended Deposited Released to Released toRelease in RCS in RCS Sump SG Secondary

Species (kg) (kg) (kg) (kg) (kg)

CSI 29.8 1.29 19.35 0.618 8.52

CsOH 181. 8.73 117. 3.74 51.2

Te 10.5 1.03 3.20 0.377 5.87

Sr 3.38(-2) 6.41(-4) 2.12(-2) 6.89(-4) 1.13(-2)

Ru 3.59(-4) 3.03(-6) 2.28(-4) 7.23(-6) 1.21(-4)

La 5.89(-5) 1.05(-6) 3.67(-5) 1.20(-6) 1.99(-5)

Ce 0 0 0 0 0

Ba 0.813 0.0197 0.509 0.0164 0.268

5 35

TABLE 520. TIME DEPENDENT FISSION PRODUCT BEHAVIOR IN STEAMGENERATOR SECONDARY SIDE

CS1 CSOH TE AEROSOLTIME RET TOTAL RET TOTAL RET TOTAL RET TOTAL(S) (KG) (KG) (KG) (KG) (KG) (KG) (KG) (KG)

9127. .0 .5 .1 3.1 .0 .3 i 5.7

9180. .1 2.1 .4 12.5 .0 1.1 .7 22.9

9234. .2 4.3 1.3 26.0 .1 2.4 2.3 47.8

9288. .9 5.9 5.13 35.7 .5 3.5 10.4 67.0

9340. 1.5 7.0 9.1 41.7 .9 4.5 17.4 80.8

9395. 2.3 7.6 13.9 45.5 1.6 5.2 27.5 90.5

9446. 3.0 8.0 i7.8 47.7 2.1 5.5 36.3 96.5

9502. 3.4 8.2 20.3 48.9 2.5 5.6 42.0 100.2

9552. 3.0 8.3 21.5 49.7 2.6 5.7 45.0 102.7

9607. 3.7 8.3 22.3 50.1 2.7 5.7 46.0 104.5

9664. 3.8 8.4 22.8 50.4 2.8 5.6 48.4 105.0

9713. 3.9 8.4 23.1 50.6 2.8 5.6 49.2 106.1

9771. 3.9 8.4 23.3 50.7 2.9 5.8 49.8 106.4

9823. 3.9 8.4 23.4 50.7 2.9 5.8 50.1 106.7

9872. 3.9 8.4 23.5 50.8 2.9 5.8 50.3 106.9

9926. 3.9 8.5 23.5 50.9 2.9 5.8 50.4 107.2

9981. 3.9 8.5 23.5 51.0 2.0 5.8 50.5 107.4

10030. 3.9 8.5 23.5 51.1 2.9 5.0 50.5 107.5

10084. 3.9 8.5 23.5 51.1 2.9 5.8 50.5 107.7

10138. 3.9 8.5 23.6 51.2 2.9 5.9 50.6 107.9

5 36

TABLE 521. CUMULATIVE FISSION PRODUCT DEPOSITION IN STEAMGENERATOR SECONDARY SIDE

Group Released to Secondary Retained on Secondary Surfaces

1 4.2 1.9

Cs 49.8 22.9

TE 5.9 2.9

SR .0 .0

RU .0 .0

LA . 0 .0

NG 200.2 .0

CE .0 .0

BA 3

Csl (qtss)

10

.........

IL4- Legend

SOURCE

TUBE BUNDLE2

SEPARATORS

DRYERS-

09000 9200 9460 9600 98,00 lo�oo 10200

TIME (sec)

FIGURE 5.10. CsI BEHAVIOR IN STEAM GENERATOR SECONDARY

CsOH (qtss)

60

50-

40-

V)

Ln

IEl 00(L LegendQ SOURCEL 20-

TUBE BUNDLE

10- SEPARATORSDRYERS-

0 4-9000 9200 9400 96'00 9800 lo6oo loi0o

TIME (sec)

FIGURE 5.11. CsOH BEHAVIOR IN STEAM GENERATOR SECONDARY

Te (qtss)6

0) 4-

3

IL LegendSOURCE2i

TUBE BUNDLE

SEPARATORS

DRYERS-

09000 9200 9460 9�00 9�00 10000 10�00

TIME (sec)

FTGURE 512. TE BEHAVIOR IN STEAM GENERATOR SECONDARY

PI (qtss)

12

100-

80-

60

Legend CD

SOURCE40

TUBE BUNDLE

20- SEPARATORSDRYERS-

09000 9200 9400 9600 9800 10000 10�00

TIME (sec)

FIGURE 5.13. PARTICULATE BEHAVIOR IN STEAM GENERATOR SECONDARY

5-41

expressed as fractions of the initial core inventory, are summarized in

Table 522. The releases of the volatile species are seen to be considerable.

Approximately 38 percent of the noble gases were calculated to be

released to the environment through the steam generator relief valves. The

sensible energy release associated with the fission product releases was

calculated to be 529 x 106 Btu.

5.4 TBA Sequence

As has been previously described, in -the TBA sequence as considered

here, initial core uncovery and heatup take place under high primary system

pressure. An accident-induced primary system break is assumed to occur after

the onset of core melting, but prior to fuel smping out of the core region.

The depressurization of the primary system through the induced break leads to

accumulator discharge with recovery and quenching of the core. The accumulator

water eventually boils off and the core remelts..

5.4.1 Release and Transport in the RCS

The analysis of the fission product tansport within the reactor

primary system was accomplished by two separate TRAP-MERGE runs. The first

TRAP-MERGE anaysis treated the initial high pressure phase of the accident,

with release to the containment out of the pressurizer safety/relief valve.

The second TRAP-MERGE analysis was applied to the later remelting phase of the

accident when the primary system was depressurized; the releases to the

containment for this phase were through the break in the hot leg piping.

The results of the primary system fission product transport during

the initial heatup and melting phase of the accident are summarized in

Tables 523 and 524 as well as Figures 514 through 517. Relatively little

primary system deposition is predicted for this phase of the accident. The

depressurization of the primary system in response to the induced break is seen

to rapidly sweep the airborne species out of the primary system.

The primary system fission product transport analyses for the

remelting phase of this sequence are summarized in Tables 525 and 526 and

illustrated in Figures 5.18 through 521. More primary system retention of

5-42

TABLE 522. ENVIRONMENTAL RELEASES FOR TMI-U-SGTR

GROUP

1 .1449

Cs .1433

pi 4.139SE-04

TE 9.1717E-02

SR 9.8275E-05

RU 1.346SE-07

LA 1.5382E-08

Ha 0.3835

CE 0.

BA 1.8183E-03

TABLE 523. TIME DEPENDENT AND FISSION PRODUCT RELEASE AND DEPOSITION IN THEPRIMARY SYSTEM FOR THE INITIAL PHASE OF THE TBA SEQUENCE

CS1 cSom TE AEROSOLTIME RET TOTAL RET TOTAL RET TOTAL RET TOTAL(S) (KG) (KG) (KG) (KG) (KG) (KG) (KG) (KG)

32759. .0 .2 .1 4.0 .0 .0 .0 .0

32874. .0 .3 .3 4.9 .0 .0 .0 4.0

32990. .0 .4 .5 6.0 .0 .0 .1 17.3

33107. .1 i.5 1.1 12.6 .0 .0 1.3 30.8

33227. .2 3.7 2.3 24.8 .0 .2 4.1 39.6

33339. .4 5.7 3.9 36.1 .0 .3 7.0 48.0

33452. .7 7.1 5.6 44.1 .0 .5 9.5 57.3

33572. 1.0 8.6 7.5 52.6 .1 .8 12.0 65.8

33884. 1.3 9.6 9.6 58.3 .1 1.0 14.5 72.6 CA)

33797. 1.4 14.9 10.2 91.3 .4 5.0 24.3 150.9

33913. .5 19.6 5.4 120.2 1.8 13.9 24.9 174.1

34028. .3 20.8 3.4 i27.3 1.9 16.4 25.4 186.9

34144. .5 21.2 5.3 129.2 1.9 17.2 26.3 200.3

34260. .5 21.2 5.0 129.7 1.9 17.4 28.1 212.6

34375. .4 21.3 4.7 129.8 2.0 17.6 29.1 223.3

34490. .4 21.3 4.3 129.8 2.0 17.6 29.4 228.3

34608. .4 21.3 3.8 i29.8 2.0 i7.8 29.5 229.8

34722. .3 V.3 3.8 129.9 2.0 17.6 29.5 230.4

34837. .3 2i.3 3.8 129.8 2.0 17.6 29.5 230.3

34933. .2 21.3 3.6 i29.8 2.0 17.8 29.5 230.8

TABLE 524. CUMULATIVE FISSION PRODUCT RELEASES FORTHE VARIOUS GROUPS DURING INITIAL PASEOF TBA SEQUENCE


1 10.4 .1

Cs 126.1 3.3

TE 17.6 2.0

SP .0 .0

RU .0 .0

LA .0 .0

NG 237.9 .0

CE .0 .0

OA .5 .0

QTBA Cs!25 -

20 -

-y

1 -V)

LLI LegendCL 10 < FUELuV) c(pp-1-1Li-i

5- UPPER PLENUM

HL + SL

0 -J31600 32000 33000 34000 35000

TIME (sec)

FIGURE 514. Csl BEHAVIOR IN THE PRIMARY SYSTEM DURING INITIAL PHASEOF TBA SEQUENCE

QTBA CsOH150

100

V)

(A

LegendLAJ

cl-

< FUELu 5 -V) CORELi UPPER PLENUM

HL + SL + P

031000 32000 33000 34000 35000

TIME (sec)

FIGURE 5.15. CsOH BEHAVIOR IN THE PRIMARY SYSTEM DURING INITIALPHASE OF TBA SEQUENCE

QTBA Te20

all, 150)

c/)

lo

Legend< FUEL

CORE

UPPER PLENUM

HL + SL A- P

0

31000 32000 33000 34000 35000

TIME (sec)

FIGURE 516. TE BEHAVIOR IN THE PRIMARY SYSTEM DURING INITIAL PHASEOF TBA SEQUENCE

QTBA PI250

200 -

15 -U)

LegendbiCL 100 < FUEL

V) CORE coLi

5 - UPPER- PLENUM

HL + S-+-P-

0-1

31000 32000 33000 34000 35000

TIME (sec)

FIGURE 517. PARTICULATE BEHAVIOR IN THE PRIMARY SYSTEM DURING INITIALPHASE OF TBA SEQUENCE

TABLE 525. T114E DEPENDENT AND FISSION PRODUCT RELEASE AND DEPOSITIONIN THE PRIMARY SYSTEM FOR THE SECOND PHASE OF THE TBASEQUENCE

CST Cso" TE AEROSOLTIME RET TOTAL RET TOTAL RET TOTAL RET TOTAL(s) (KG) (KG) (KG) (KG) (KG) (KG) (KG) (KG)

46426. .0 .0 .0 .1 .0 .0 .0 .0

47098. .0 .4 .4 2.7 .0 .0 .5 7.1

47766. .3 2.8 2.1 15.4 .4 .8 2.6 25.3

48436. .8 4.5 3.7 26.9 l.i 1.9 9.5 71.7

49104. .9 6.1 6.2 36.7 1.6 2.6 24.4 137.1

49774. 1.2 7.4 8.9 45.0 2.2 3.4 45.4 205.5

50443. i.4 8.2 10.3 49.9 2.0 3.8 57.7 249.9

Bills. i.4 9.2 10.3 50.0 2.0 3.8 57.7 248.9

51784. 1.4 8.2 10.3 50.0 2.0 3.8 57.7 248.9

5245i. i.4 8.2 10.3 50.0 2.0 3.8 57.7 248.9

53120. 1.4 8.2 10.3 50.0 2.0 3.8 57.7 248.9

53789. 1.4 8.2 10.3 50.2 2.0 3.8 57.7 248.9

54458. 1.4 8.3 10.3 50.7 2.0 3.8 57.8 249.1

55132. 1.4 8.6 10.4 52.5 2.0 3.9 57.8 250.1

55797. 1.5 9.0 10.6 55.1 2.0 4.0 57.9 252.7

56466. 1.5 9.5 10.8 58.6 2.1 4.3 58.1 262.5

57135. 1.5 9.7 11.0 59.6 2.i 4.8 58.4 281.9

57805. 1.6 9.7 11.1 59.6 2.1 5.5 58.7 325.5

58473. 1.8 9.7 11.2 59.5 2.1 6.3 99.1 388.5

59i4s. 1.6 9.3 11.2 58.8 2.1 7.6 59.5 505.9

TABLE 526. CUMULATIVE FISSION PRODUCT RELEASES FOR THEVARIOUS GROUPS DURING SECOND PHASE OF TBASEQUENCE


1 4.5 .8

Cs 55.1 10.7

TE 7.6 2.1

SR i .0

RU .0 .0 c-n

LA .0 .0 CD

NG 109.7 .0

CE .0 .0

BA 2.6 .3

QTBA Csl10

-

V) 6-V)

Li 4-Legend

V) FUELLu

2- CORE

UPPER PLENUM

044000 46000 48000 50000 52000 54000 56000 58000 0000

TIME (sec)

FIGURE 5.18. CsI BEHAVIOR IN THE PRIMARY SYSTEM DURING SECONDPHASE OF TBA SEQUENCE

QTBA CsOH60

-y%%-." 40 - ---V)

LiCL< Legend0 20 -V) FUEL

CORE

UPPER PLENUM

044000 46000 48000 50000 52000 54000 56000 58000 bUUUU

TIME (sec)

FIGURE 519. CsOH BEHAVIOR IN THE PRIMARY SYSTEM DURING SECONDPHASE OF TBA SEQUENCE

QTBA Te8

6

-y

V)V)

4-Ln

< LegenduV) r7 I l

U-1 F uc-i-

CORE

UPPER PLENUM

044000 46000 48000 50000 52000 54000 56000 58000 60000

TIME (sec)

FIGURE 520. TE BEHAVIOR IN THE PRIMARY SYSTEM DURING SECONDPHASE OF TBA SEQUENCE

QTBA Pi600

50 -

400 -V)

300 -

Q-< Legend0 200 -V) FUELLi

CORE-100-

UPPER PLENUM

044000 46000 48000 50000 52000 54000 56000 58000 60600

TIME (sec)

FIGURE 521. PARTICULATE BEHAVIOR IN THE PRIMARY SYSTEM DURINGSECOND PHASE OF TBA SEQUENCE

5- 55

all the species considered is seen during the econd phase of the accident.

The indicated decrease of the CsI and CsOH releases from the fuel in

Figures 5.18 and 519 corresponds to the predicted settling of these species

back into the core region at late times in the sequence. It is also

interesting to note the significant releases o Te and particulates,

Figure 520 and 521, near the end of the in-vessel phase of the accident.

Since the primary system is depressurized, it takes considerable time for the

debris to melt through the vessel head; during this time the debris reheat and

release the above species.

5.4.2 Release and Transport in Containment

in this scenario the fission products from both the primary system

and the reactor cavity are released to the lower compartment of the ice

condenser containment. The fission product flowpath is from the lower

compartment to the ice condenser, from there to the upper compartment, and

after containment failure, to the environment. The calculational procedure

shown in Figure 5.5 is applicable to this sequence also. During the initial

core heatup phase of the accident, there is ample ice in the ice condenser;

during the later core remelt phase most of the ice is gone. The reactor

cavity is essentially dry throughout this accident sequence.

Table 527 summarizes the calculated fission product source terms to

the containment. The size distribution of the airborne particles in the upper

compartment is given in Table 528, and the fractions of core inventory

released to the environment for each of the fission product groups are given

in Table 529. The locational distribution of the various fission product

groups at the end of the calculation are given in Table 530.

The ice was predicted to be completely melted in this sequence prior

to the time of reactor vessel failure. To gain further insight on the

effectiveness of the ice in removing fission poducts, the above analyses was

repeated assuming no more fission product removal by the ice after 90 percent

of it had melted. This would correspond to nonuniform melting of the ice

with some flow breakthrough before all the ice had melted. The results of

this case are summarized in Table 531. From the changes in the releases to

TABLE 527. FISSION PRODUCT SOURCE TERMS RELEASED TO THECONTAINMENT FOR TM SEQUENCE

DURING DURING DURINGIN-VESSEL PUFF CORE-CONCRETE

GROUP RELEASE RELEASE ATTACK

1 .8977 2.529SE-02 8.7607E-04

Cs .8777 2.830SE-02 8.0673E-04

pi 2.8879E-03 1.8729E-03 0.

TE .350 .1188 8.3143E-02

SR i.0074E-03 8.734BE-04 isgo

RU 1.458SE-08 1.475BE-06 3.7599E-08

LA 1.483SE-07 i.2132E-07 8.Oi79E-03

me .9543 4.470SE-02 0.

CE 0. 0. 5.6207E-03

OA 1.893SE-02 1.637SE-02 iOlo

TABLE 528. SIZE DISTRIKJTION OF AEROSOLS IN CONTAIN14ENT - TBA SCENARIO

TIME 9.501 10.001 10.501 il.501 12.501 14.000 16.000 19.000 23.000 28.003

DENSITY 3.OOE+00 3.OOE+00 3.OOE+00 3.OOE+00 3.OOE+00 3.OOE+00 3.OOE+00 3.52E+00 3.90E+00 3.78E+00


5.00E-03 0. 0. 0. 0. 0. 0. 1.91E-25 0. 2.88E-la 0.8.20E-03 6.36E-25 0. 0. 0. 0. 2.37E-24 4.11E-20 2.35E-16 1.87E-13 0.1.35E-02 2.01E-18 3.59E-19 i.25E-22 0. 0. 3.65E-19 2.44E-16 1.44E-13 4.0GE-11 0.2.21E-02 1.97E-14 3.28E-13 1.94E-16 G.78E-23 1.16E-25 I 3E-14 1.91E-13 3.91E-11 4.36E-09 0.3.62E-02 1.39E-11 9.84E-10 4.13E-11 1.12E-13 5.68E-IG 8.89E-11 4.09E-11 5.26E-09 2.39E-07 1.02E-225.95E-02 2.52E-09 1.58E-07 4.26E-08 3.74E-09 4.27E-10 9.07E-08 1.81E-08 3.39E-07 0.57E-06 1.47E-12 Ln9.76E-02 1.97E-07 4.02E-06 2.299-06 8.01E-07 3.15E-07 1.88E-OS 1.33E-08 1.01E-05 9.36E-05 i.ISE-07 Ln1.60E-Ol 8.14E-00 2.68E-05 2.08E-05 1.29E-05 8.51E-00 6.49E-04 6.38E-05 1.46E-04 7.68E-04 3.78E-052.63E-01 1.92E-04 1.22E-04 1.08E-04 8.69E-05 7.1GE-O5 6.89E-03 1.89E-03 1.15E-03 4.23E-03 1.18E-034.31E-01 2.77E-03 1.38E-03 1.31E-03 1.48E-03 1.09E-03 2.80E-02 2.02E-02 5.61E-03 1.79E-02 1.30E-027.07E-01 2.43E-02 1.32E-02 1.30E-02 1.27E-02 1.24E-02 5.36E-02 9.62E-02 1.97E-02 6.12E-02 7.92E-021.16E+00 1.24E-01 7.349-02 7.45E-02 7.63E-02 7.79E-02 7.89E-02 1.80E-Ol 6.86E-02 1.67E-01 2.62E-014.9,E+00 2.78E�01 2.119-01 2.18E-01 2.31E-Oi 2.43E-01 1.94E-01 2.25E-01 2.igE-Oi 3.09E-oi 3.83E-eii.M+00 2.76E-01 3.40E-01 3.501-01 3.68E-01 3.82E-Oi 3.19E-01 2.81E-Ol 3.54E-Oi 2.99E-Ol 2.12E-015.12E+00 2.031-01 2.70E-01 2.65E-Ol 2.53E-Ol 2.39E-01 2.23E-Ol 1.63E-01 2.23E-Ol 1.17E-01 4.58E-028.41E+00 7.48E-02 8.08E-02 7.06E-02 5.41E-02 4.20E-02 S.OOE-02 3.17E-02 8.59E-02 2.21E-02 3.97E-031.38Ei-Ol 1.69E-02 1.01E-02 6.79E-03 3.55E-03 2.12E-03 1.52E-02 1.14E-03 2.08E-02 1.63E-03 1.03E-042.26E+01 6.09E-04 3.75E-04 1.8SE-04 5.57E-05 2.62E-05 7.35E-04 1.33E-05 2.49E-03 3.35E-05 8.37E-073.71E+01 0.299-06 2.48E-00 7.59E-07 1.95E-07 7.27E-08 1.03E-05 2.27E-07 1.15E-04 1.63E-07 8.71E-108.09E+01 6.385-08 3.19E-09 7.81E-10 1.52E-10 4.50E-11 2.84E-08 2.22E-09 1.57E-06 1.79E-10 2.62E-131.00E+02 4.299-08 8.72E-13 1.77E-13 2.65E-i4 6.24E-15 1.2iE-11 4.28E-i2 5.50E-09 4.44E-14 4.OGE-11

TABLE 529. FRACTION OF CORE INVENTORY RELEASED FROM CONTAINMENT -TBA SCENARIO

TIME FISSION PRODUCT GROUP(HR) I Cs pi TE SR RU LA CE BA PE TR

1:.:Ol 1.86E-07 1.83E-07 3.29E-10 1.21E-07 4.38E-11 6.45E-14 6.53E-15 0. 8.24E-10 0. D.01 1.41E-02 1.40E-02 3.41E-05 1.59E-02 1.21E-05 1.91E-08 1.81E-09 0. 2.28E-04 0. 0.

10,501 1.44E-02 1.43E-02 3.50E-05 1.63E-02 1.25E-05 1.97E-08 1.88E-09 D. 2.34E-04 0. 0.11,501 1,4 E02 1.44E-02 3.51E-05 1.64E-02 1.26E-05 1.9se-08 1.87E-09 0. 2.36E-04 0. 0.' 1 -09 2.36E-04 0. 0.12.90 1.4:E-02 1.44E-02 3.51E-05 1.84E-02 1.26E-05 1.98E-os 1.87E 0.14.000 1.50E-02 1.49E-02 3.74E-05 1.69E-02 1.36E-05 2.13E-08 2.02E-09 0. 2.55E-04 0. 0.

2.27E-02 8.59E-05 2.15E-02 3.69E-05 5.47E-08 5.42E-09 0. 6.90E-04 0. 0.16.000 2.31E-02 +03 9.97E-0219.000 2.93E-02 2.88E-02 2.87E-04 4.429-02 6.12E-02 2.33E-07 3.51E-03 2.21E-03 3.72E-02 2.74E23.000 2.94E-02 2.90E-02 2.89E-04 6.47E-02 9.26E-02 5.38E-07 4.51E-03 3.11E-03 5.70E-02 4.98E403 1.00E-0128.003 2.94E-02 2.90E-02 2.89E-04 8.35E-02 9.43E-02 2.30E-06 4.58E-03 3.16E-03 5.87E-02 5.S5f+03 1.01E-Ol 00

TABLE 5.30. DISTRIBUTION OF FISSION PRODUCTS BY GROUP - TBA SCENARIO


I 5.8 12 3.4 x 1-5 0 0.19 0.66 5.1 x 1-2 2.9 x 10-2

Cs 7.6 x 12 1.1 x 10-5 0 0.19 0.64 4.9 x 10-2 2.9 x 10-2

Te 0.13 4.0 x 10-4 0.11 0.23 0.39 4.5 x 10-2 8.3 x 10-2

Sr 2.6 x 10-4 2.6 x 10-3 0.83 6.4 x 10-2 5.4 x 10-4 1.2 x 10-2 9.4 x 10-2Ln

Ru 3.7 x 10-7 1.5 x 10-9 1.0 1.5 x 10-6 7.8 x 10-7 4.3 x 10-7 2.3 x 10-6 to

La 3.8 x 10-8 1.7 x 10-4 0.99 1.5 x 10-3 7.8 x 10-8 5.1 x 10-4 4.6 x 10-3

Ce 0 9.7 x 10-5 0.99 1.2 x 10-3 0 4.4 x 10-4 3.2 x 10-3

Ba 4.8 x 10-3 1.5 x 10-3 0.86 4.4 x 12 1.0 x 12 1.0 x 10-2 5.9 x 10-2

Tr 0 0 0 0.89 0 7.7 x 10-3 0.10

TABLE 531. DISTRIBUTION OF FISSION PRODUCTS BY GROUP - TBAI SCENARIO*


1 5.8 x 12 3.4 x 1-5 0 0.19 0.57 0.12 4.5 x 12

Cs 7.6 x 12 1.1 x10-5 0 0.19 0.56 0.12 4.4 x 10-2

Te 0.13 4.0 x10-4 0.11 0.23 0.36 6.5 x 12 8.8 x 10-2

Sr 2.6 x 10-4 2.6 x10-3 0.83 3.8 x 12 2.4 x 10-4 1.3 x 10-2 9.4 x 10-2

Ru 3.7 x10-7 1.5 x10-9 1.0 2.2 x 10-6 3.7 x 10-7 8.7 x 10-7 2.4 x 10-6

La 3.8 x10-8 1.7 x10-4 0.99 1.5 x 10-3 3.6 x 10-8 5.1 x 10-4 4.5 x 10-3

Ce 0 9.7 x10-5 0.99 1.1 x 10-3 0 3.9 x 10-4 3.1 x 10-3

Ba 4.8 x10-3 1.5 x10-3 0.86 4.2 x 12 4.6 x 10-3 1.5 x 10-2 6.0 x 10-2

Tr 0 0 0 0.88 0 7.7 x 10-3 0.10

Ice bed DF = I with less than 10% of ice remaining

5 61

the environment, it can be seen that the availability of the ice can have a

significant impact on predicted environmental eleases.

5.5 Noble Gas and Energy Release to Environment

The release of noble gases to the environment is calculated by the

MARCH 3 code rather than the fission product tansport codes TRAP and NAUA

since the noble gases are assumed to be transported with the bulk flow of

gases without attenuation. The energy associated with gases escaping the

containment is also calculated in the MARCH 3 code. The environmental

releases of noble gases and energy are tabulated in Table 532. The heat of

vaporization of the steam in the escaping gases is not included in the table.

For the TMLU-SGTR scenario the release is essentially a puff 20 minutes

duration) containing 38 percent of the noble gses and 529 x 106 Btu of

energy.

5.6 Icebed Decontamination

One of the important factors affecting the magnitude of source terms

for the ice-condenser plant is the availability of ice at the time of release

and the effectiveness with which decontamination occurs. The ICEDF computer

code was used in the Source Term Code Package to predict the amount of

decontamination of aerosols in the icebed. The principal mechanism for removal

is diffusiophoresis, the flow of aerosols to te ice with condensing steam.

As a result the decontamination factor is seen to be very sensitive to the

fraction of steam in the gases flowing through the bed. Table 533 provides.a

comparison of decontamination factors for two equences, S3HF1 and TB(S3B).

The decontamination factor is applied to the flow passing through the ice-bed.

If there is significant recirculation flow from the upper compartment to the

lower compartment, as in the S3HF1 sequence before head failure, the effective

decontamination factor can be higher than the ingle pass valve because the

aerosols have more than one opportunity to be captured in the ice. Up to the

time of bottom head failure 410 min) in the SHF1 sequence, the single pass

DF is quite low because the fans have circulated non-condensible gases back

into the lower compartment and the partial pressure of steam is low (-2.5 psi

5 62

TABLE 532. NOBLE GAS AND ENERGY RELEASE TO THE ENVIRONMENT

Scenario S3HF1 Scenario S3HF2

Time Noble Energy Time Noble EnergyHr Gas (Btu) Hr Gas (Btu)

7.0 0.351 5.54(6) 7.0 0.366 5.42(6)7.5 0.350 5.54(6) 7.5 0.364 5.42(6)8.5 0.382 5.68(6) 8.5 0.395 5.57(6)9.0 0.439 5.89(6) 9.0 0.454 5.81(6)9.5 0.597 7.62(6) 9.5 0.576 7.41(6)

10.0 0.667 9.10(6) 10.0 0.666 9.02(6)10.5 0.690 9.27(6) 10.5 0.700 9.30(6)11.0 0.718 9.50(6) 11.0 0.728 9.54(6)15.5 0.822 1.04(7) 15.5 0.859 1.30(7)20.0 20.0

Scenario S3B Scenario S3HF3

Time Noble Energy Time Noble EnergyHr Gas (Btu) Hr Gas (Btu)

7.0 0.384 6.88(6) 7.0 0.366 5.42(6)7.5 0.411 6.98(6) 7.5 0.365 5.42(6)8.5 0.667 1.15(7) 8.5 0.431 5.71(6)9.0 0.707 1.19(7) 9.0 0.502 6.00(6)9.5 0.733 1.22(7) 9.5 0.668 1.03(7)

10.0 0.749 1.23(7) 10.0 0.699 1.06(7)10.5 0.763 1.25(7) 10.5 0.726 1.09(7)11.0 0.771 1.26(7) 11.0 0.758 1.13(7)15.5 0.875 1.40(7) 15.5 0.836 1.22(7)20.0 20.0

5-63

TABLE 532. NOBLE GAS AND ENERGY RELEASETO THE ENVIRONMENT

(continued)

Time Noble Energy(Hr) Gas (Btu)

Scenario TBA

9.5 0.139 5.30(6)

10.0 0.253 6.61(6)

10.5 0.253 6.61(6)

11.5 0.253 6.61(6)

12.5 0.253 6.61(6)

14.0 0.262 6.64(6)

16.0 0.571 7.21(6)

19.0 0.965 1.24(7)

23.0 0.994 1.58(7)

28.0 --

5 64

TABLE 533. ICEBED DECONTMINATION FACTOR

Decontamination FactorTime S3HF1 Time TB(S3B)(Min) (Min)

380 1.6 340 7.1

400 1.4 360 5.2

420 2.2 380 --

440 1.0 400 --

460 1.0 450 3.9

500 -- 500 2.7

550 6.8

550 1.4 600 7.3

600 3.9 700 7.1

650 4.2 800 7.1

700 4.2 900 4.0

800 5.3 1000 3.6

900 5.4

5 65

steam relative to a total pressure of 17 psi). In comparison, the DF during

this time period for the TB sequence (vessel failure occurs at 380 min in the

TS sequence) is substantially higher. In this sequence steam released from

the reactor coolant system has depleted the amount of non-condensible gases in

the lower plenum 18 psi steam out of a total of 21 psi). Throughout the

accident the predicted decontamination factors are closely correlated with the

fraction of steam in the lower compartment.

6-1

6. SUMMARY AND CONCLUSIONS

This report presents the results of Source Term Code Package

analyses for a number of postulated accident sequences in the Sequoyah ice

condenser containment PWR. The present results supplement the earlier

analyses reported in Vol. IV of BMI-2104. In addition to utilizing state-of-

the-art source term methodology, the work reported here reflects the latest

thinking on accident sequence identification and definition. Both of these

factors contribute to the nature of the results reported here.

As a result of the current view of the most likely modes of safety

system failure several accident sequences previously considered to be risk

significant no longer appear to be so; these include the S3H and S2HF

sequences. At the time of the most probable recirculation system failure in

these sequences there are still ample quantities of ice available to mitigate

the accident consequences. Also, under the current definition of these

sequences and the containment failure pressure aopted by SARRP, there appears

to be relatively low probability of early containment failure in these

sequences.

For several of the accident scenarios eamined there are substantial

quantities of water in the reactor cavity. This water may offer the potential

for termination of these accident scenarios if coolable debris beds are

formed; or, if the debris are uncoolable, the presence of this water has been

found to result in significant fission product rtention as well as reduction

of driving forces for leakage out of the containment.

In some of the accident sequences considered in this report the

water in the reactor cavity is predicted to be cmpletely evaporated; the re-

release of fission products retained by this water has not been included in

the present analyses.

For the accident sequences involving pump seal failures core over-

heating is predicted to occur at high primary system pressures and significant

retention of the volatile fission product species on primary system surfaces

is predicted. The potential long-term revaporization of these species and the

impact of such revaporization on the predicted evironmental releases has not

been addressed in the present study.

6-2

Included in this effort is the assessment of the environmental

source terms for an accident-induced steam generator tube rupture scenario.

The results presented here should also be applicable to other PWRs of similar

primary system design.

7- 1

7. REFERENCES

1. Gieseke, J. A., et al., "Source Term Code Pckage: A User's Guide",NUREG/CR-4587, Draft, April, 1986.

2. Gieseke, J. A., et al., "Radionuclide Release Under Specific LWR AccidentConditions", BMI-2104, Volume IV, Draft, July, 1984.

3.. Silberberg, M., et al., "Reassessment of the Technical Bases forEstimating Source Terms", NUREG-0956, Draft, July, 1985.

4. Wooton, R. O., Cybulskis, P., and Quayle, S. F., 11MARCH2 (MeltdownAccident Response Characteristics) Code Description and Usir's Manual",'�attelle'_s ColumbuTLaboratories, NUREG/CR-:3988, BMI-2115,September, 1984.

5. Kuhlman, M. R., Lehmicke, D. J., and Meyer, R. O., 1CORSOR User'sManual", Battelle's Columbus Laboratories, NUREG/CR-4173, BMI-2122,March, 1985.

6. Cole, R. K., Kelly, D. P., and Ellis, M. A., "CORCON MD2: A ComputerProgram for Analysis of Molten Core Concrete Interactions",NUREG/CR-3920, August, 1984.

7. Powers, D. A., Brockman, J. E., and Shiver, A. W., 11VANESA, A MechanisticModel of Radionuclide Release and Aerosol Gneration During Core DebrisInteractions with Concrete", Sandia National Laboratories, NUREG/CR-4308,SAND85-1370, Draft.

8. Muir, J. F., et al., 11CORCON-Mod 1: An Improved Model for Molten-Core/Concrete Interactions", Sandia National Laboratories, NUREG/CR-2142,SAND80-2415, July, 1981.

9. Freeman-Kelly, R., and Jung, R. G., "A User's Guide for MERGE",Battelle's Columbus Laboratories, NUREG/CR-4172, BMI-2121, March, 1985.

10. Jordan, H., and Kuhlman, M. R., TRAP-MELT2 User's Manual", Battelle'sColumbus Laboratories, NUREG/CR-4205, BMI-2124, May, 1985.

11. U.S. Nuclear Regulatory Commission, NReactor Safety Study - An Assessmentof Accident Risks in U.S. Commercial Nuclear Power Plants", WASH-1400(NUREG 75-104), October, 1975.

NFIC FORM 335 U S. NUCLEAR REGULATORY COMMISSION I REPORT NUMBER (Assg�edby TIDC, add Vol No, da�yj

(2-841NRCM 1102,3201,3202 BIBLIOGRAPHIC DATA SHEET NUREG/CR-4624, Vol. 2SEE INSTRUCTIONS ON THE REVERSE

2 TITLE AND SUBTITLE 3 LEAVE BLANK

Radionuclide Release Calculations fr Selected SevereAccident Scenarios, Volume 2 PWR, Ice CondenserDesign 4 DATE REPORT COMPLETED

MONTH YEAR

5 AUTHOR(S) Ma y IR S. Denning, J. A. Gieseke, P. Cybulskis, K. W. Lee, 6 DATE REPORT ISSUED

H Jordan, L. A. Curtis, R. F. Kelly,, V. Kogan, and MONTH YEAR

P. M. Schumacher July 19867 PERFORMING ORGANIZATION NAME AND MAILING ADDRESS fl-cl.deZp Cape; 8 PROJECT/TASK/WORK UNIT NUMBER

Battelle's Columbus Division 9 FIN OR GRANT NUMBER

505 King AvenueColumbus, Ohio 43201 A1322

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Division of Risk Analysis and Operations Technical ReportOffice of Nuclear Regulatory Research b PERIOD COVERED (1-1.s- dtm)

U.S. Nuclear Regulatory CommissionWashington, D.C. 20555 6/85-7/86

12 SUPPLEMENTARY NOTES

13 ABSTRACT 200 wods o!e,,)

This report presents results of analyses of the environmental releases offission products (source terms) for severe accident scenarios in a pressurizedwater reactor with an ice-condenser containment. The analyses were performedto support the Severe Accident Risk Reduction/Risk Rebaselining Program (SARRP)which is being undertaken for the U.S. Nuclear Regulatory Commission bySandia National Laboratories. In the SARRP program, risk estimates are beinggenerated for a number of reference plant designs. The Sequoyah Plant hasbeen used in this study as an example of a PWR ice-condenser plant.

14 DOCUMENT ANALYSIS - KEYWORDS/DESCRiPTORS 15 AVAILABILITYSTATEMENT

Source TermsSevere Accidents UnlimitedFission Products 16 SECURITY CASSIFICATION

(Tha page)

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