121709 masters project in mechanical engineering

7/29/2019 121709 Masters Project in Mechanical Engineering

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Spent Nuclear Fuel Dry Cask Storage Requirements for

Yucca Mountain and Available Margin

by

Daniel Sadlon

A Project Submitted to the Graduate

Faculty of Rensselaer Polytechnic Institute

in Partial Fulfillment of the

Requirements for the degree of

MASTER OF ENGINEERING IN MECHANICAL ENGINEERING

Approved:

_________________________________________

Dushyanthi Hoole, Project Adviser

Rensselaer Polytechnic Institute

Hartford, CT

December, 2009


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CONTENTS

LIST OF FIGURES..........................................................................................................iv

LIST OF TABLES.............................................................................................................v

ACKNOWLEDGMENTS................................................................................................vi

ABSTRACT....................................................................................................................vii

ACRONYMS.................................................................................................................viii

NOMENCLATURE.......................................................................................................viii

1. Introduction.....................................................................................................................9

1.1Problem Description .........................................................................................11

1.2Problem Development.......................................................................................11

1.3Problem Benefit.................................................................................................11

2. Background...................................................................................................................12

2.1Classification of Spent Nuclear Fuel (SNF)......................................................12

2.2Spent Fuel Storage Descriptions........................................................................12

2.2.1Spent Fuel Pool (Wet Storage)...................................................................13

2.2.2Dry Cask Storage........................................................................................14

2.3NRC Spent Fuel Requirements..........................................................................18

2.3.1Storage Facilities........................................................................................22

2.3.2Dry Cask Requirements..............................................................................22

2.4Spent Fuel Requirements...................................................................................23

2.5Failed Dry Cask Impact.....................................................................................24

2.5.1Radiation Releases and Protection..............................................................26

2.5.2Public Impact..............................................................................................27

2.6Spent Fuel Failure Mechanisms.........................................................................29

2.6.1Delayed Hydride Cracking (DHC).............................................................292.6.2Stress Corrosion Cracking (SCC)...............................................................30

2.6.3Accelerated Creep Leading To Creep Rupture (CR) .................................30

2.6.4Diffusion Controlled Cavity Growth (DCCG)...........................................31

3. Temperature, Stress and Age Studies on Spent Nuclear Fuel Storage.........................32

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3.1Spent Fuel Safety Concerns...............................................................................32

3.2Spent Fuel Rod Temperature Requirements......................................................32

3.2.1PNL-10813 Study.......................................................................................34

3.2.2CANDU Reactors.......................................................................................38

3.2.3German Study.............................................................................................38

3.2.4PNL-6189 Study.........................................................................................39

3.2.5PNL-SA-13879 Study.................................................................................42

3.2.6Domestic Support.......................................................................................43

3.2.7Operating Locations....................................................................................43

3.2.8Hanford Engineering Development Laboratory.........................................45

4. Analysis........................................................................................................................47

4.1Spent Nuclear Fuel Margin Available...............................................................48

5. Conclusions...................................................................................................................50

6. Future Industry Proposals.............................................................................................53

7. References.....................................................................................................................55

Attachment A Dry Cask Models...................................................................................58

Attachment B Dry Storage Locations...........................................................................60

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LIST OF FIGURES

Figure 1 Typical Spent Fuel Pool Design and Location []............................................13

Figure 2 Spent Fuel Pool Capacity Prediction []...........................................................14

Figure 3 Typical Dry Cask Design []............................................................................15

Figure 4 Dry Cask Horizontal Orientation Design []....................................................15

Figure 5 Typical Dry Cask Dimensions [].....................................................................15

Figure 6 Dry Cask Radiation Shield []..........................................................................16

Figure 7 Dry Cask General Functions []........................................................................17

Figure 8 Dry Cask Safety Functions []..........................................................................17

Figure 9 Spent Fuel Storage Locations in the United States of America []..................18

Figure 10 DHC Formation [].........................................................................................30

Figure 11 Cladding Strain vs. Time at 150 MPa []........................................................33

Figure 12 Cladding Strain vs. Time at 200 MPa []........................................................33

Figure 13 Cladding Strain vs. Time at 380C []............................................................34

Figure 14 Cladding Stress vs. Temperature []...............................................................35

Figure 15 Cladding Stress vs. Temperature (Aged Fuel) []...........................................36

Figure 16 Cladding Stress vs. Temperature (Example) [].............................................37

Figure 17 Dry Casks in Storage at the Ahaus, Germany Site []....................................39

Figure 18 Temperature vs. Allowable Stress Limits []..................................................40

Figure 19 Temperature Decay Rate []...........................................................................41

41

Figure 20 Damage Accumulated in Dry Storage [].......................................................42

Figure 21 PWR Nuclear Fuel Rod Assembly []............................................................47

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LIST OF TABLES

Table 1: The Uranium-238 Decay Chain..........................................................................25

Table 2: The Uranium-235 Decay Chain..........................................................................26

Table 3: Typical Radiation Exposures..............................................................................27

Table 4: Temperature vs. Stress Limits for Varying Fuel Ages []...................................41

Table 5: Vault Storage Dry Cask Operating Experience []..............................................43

Table A-1 Dry Cask Models..........................................................................................58

Table B-1 Dry Cask Storage Locations.........................................................................60

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ACKNOWLEDGMENTS

Special thanks are extended to Dushyanthi Hoole, my project advisor. Without her

guidance and recommendations the completion of this project would not have been

achievable.

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ABSTRACT

Currently much focus in the energy industry has been placed on finding alternate

sources of energy. An efficient source of energy is nuclear energy. However the

accompanying problem of radioactive waste disposal, including storage of spent fuel

from nuclear industries, is also a major public and safety concern. The Yucca Mountain

storage facility was first proposed in 1986 to resolve this issue, however it was cancelled

by the current Presidential Administration []. Although details have not been released as

to the reasons for the cancellation, safety and cost concerns are certainly two of the main

reasons. The possibility of reducing storage costs by lowering current temperature

requirements appropriately while not breaching safety concerns formulated by the

Nuclear Regulatory Commission (NRC, 1986) is explored in this project. (Theserequirements have not been reviewed in many years, and are likely to be too restrictive

for current improved storage designs and conditions.) Towards this end:

1. The current industry requirements for spent nuclear fuel will be reviewed in light

of current best practices for temperature and age,

2. The reasons attributed to the requirements for Spent Nuclear Fuel (SNF) that

needs to be moved to repositories at present in the US and elsewhere will be

explored and

3. Due to the fact SNF is already stored at plant sites where its temperature drops

over the first five years (best practice), more relevant storage requirements will

be suggested by studying the appropriate literature citing conclusions of tests

performed for dry cask safety and NRC SNF temperature requirements.

Through the study of the above, it will be determined whether less demanding but

adequately safe temperature conditions are possible so as to make the Yucca Mountain

proposal more economically feasible.

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ACRONYMS

BNL Brookhaven National LaboratoryBWR Boiling Water Reactor CR Creep Rupture

DCCG Diffusion Controlled Cavity GrowthDHC Delayed Hydride CrackingDOE Department of EnergyEPA Environmental Protection AgencyEPRI Electric Power Research InstituteISFSI Independent Spent Fuel Storage InstillationLWR Light Water Reactor MRS Monitored Retrievable Storage InstillationMTU Metric Ton of UraniumNAS National Academy of SciencesNRC Nuclear Regulatory Commission

PNL Pacific Northwest National LaboratoryPWR Pressurized Water ReactorSCC Stress Corrosion CrackingSNF Spent Nuclear FuelSNL Sandia National Laboratory

NOMENCLATURE

MPa Mega PascalC Degrees Celsius

F Degrees Fahrenheitmhoop Dry storage cladding hoop stress, MPa

P Internal gas pressure of the rod, psiT1 Temperature at which P was determined, Kt Cladding wall thickness, in.Dmid Cladding midwall diameter, in. A factor, 0.95 for PWR rods or 0.90 for BWR rodsT2 Allowable storage temperature for mhoop , K

K KelvinE Youngs Modulus

Stressmrem millirem

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1. Introduction

Currently all nuclear power plants in the United States (as well as abroad) produce

spent nuclear fuel. The fuel is a byproduct of the nuclear fission reaction. A nuclear

fission reaction is one in which uranium atoms are bombarded by free neutrons and

nuclei, which in turn, will release more nuclei. This reaction produces the heat used in

the plants power production. For a more detailed description of the fission reaction see

Section 2.4. As time passes, the fuel ages, and must be replaced with new fuel. Many

foreign countries reprocess this fuel; however this type of reprocessing system is not

currently in place in the United States even though the installment of such policies is

achievable.

Nuclear power plants in the United States are currently required to store the spentfuel rods either on site, at other plants sites, or at dedicated spent fuel sites. The fuel can

be stored in a spent fuel pool (wet storage) or in dry casks (dry storage). A spent fuel

pool keeps the fuel rods fully covered in water and removes the heat energy remaining in

the fuel. Dry casks are used only after sufficient heat energy is removed from the spent

fuel rods, and can be maintained sufficiently cool in an encapsulated dry cask. Spent

fuel is also often times transported in such casks from one site to another.

When plants were originally licensed they were not required to build with the

capability to store their life time supply of spent fuel on site, in a spent fuel pool. While

not all plants fully utilize their spent fuel pool (some send their spent fuel elsewhere, or

use dry casks) many exclusively do, and are currently running into capacity issues due to

license extensions (i.e., plants can re-apply for operating licenses and therefore extend

the storage requirements). This being the case, there is an industry need for additional

storage facilities.

Yucca Mountain was proposed in the early 1980s as a dedicated spent fuel storage

facility for a significant portion of the United States spent fuel. It was proposed, as the

site was deemed to be sufficiently safe, from a design and location perspective. The site

would also be large enough to store many plants excess spent fuel.

The current Presidential Administration cancelled the proposal for the Yucca

Mountain site []. The reasons behind the project cancellation are not fully clear;

however safety and financial considerations are certainly the primary culprits.

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A leading concern at the nuclear power plant site is radiation released from the spent

fuel rods. Thermal and Irradiation creep can cause radiation damage in metals and

alloys in reactors by affecting their mechanical properties. This can lead to spent fuel

being released to the environment, which is known as fuel cladding failure. In dry

storage methods, creep strain is an important spent fuel rod failure mechanism. For an

example of thermal and irradiation creep behavior in metal alloys, see Journal Article

Thermal and Irradiation Creep Behavior of a Titanium Aluminide in Advanced Nuclear

Plant Environments []. As the following report will discuss, the leading contributor to

spent fuel rod failure has been demonstrated to be creep rupture by many sources. Creep

Rupture is impacted by fuel storage temperature. This safety parameter will be

investigated to encompass the requirements and proposed design requirements present in

the Yucca Mountain proposal [] meet temperature and fuel ageing requirements with

those required by the NRC.

Determination of the temperature requirements and design criteria will help

determine whether sufficient safety margin is present in the design such that spent fuel

temperature and age requirements may be relaxed. Doing so would result in a cost

reduction in the site design as well as a reduction in the design requirements of the spent

fuel dry casks. Additionally, the effects of fuel age will be analyzed to determine when

the most appropriate time to move spent fuel to dry casks would be.

This report:

Identified a spent fuel storage location that was canceled (Yucca Mountain)

Determined the cause for the cancellation (safety/finances)

Determined the leading contributor to risk (fuel rod failure)

Determined leading factor for the failure mode (temperature)

Analyzed data which determined temperature limits

Compared temperature limits from data to NRC required temperature limits Identified deltas between the two temperature limits

Explained how industry could take advantage of the margin available

This list can be used to provide the reader with a reference point to the overall

purpose of the project.

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1.1 Problem Description

The specific problem to be assessed in this project is to determine whether the NRC

required fuel storage limits for spent nuclear fuel in dry cask storage is appropriate (interms of temperature requirements and spent fuel age requirements). It is believed by

many research organizations as well as utility personnel that these limits are too

conservative, and that the fuel can meet the current safety expectations, after calculating

the amount and the relationship of, the storage requirements recommended by the NRC.

1.2 Problem Development

The problem will be solved by examining industry and experimental data from a

number of nuclear research facilities and utilities. Through data investigation, it will be

determined how much margin is available for the storage of spent fuel. The margin will

be defined in terms of available temperature margin and what the most appropriate age

of the fuel should be when transferring to dry storage, while maintaining NRC defined

radiation and safety regulations.

1.3 Problem Benefit

If NRC spent fuel dry cask storage requirements are too stringent it will allow

utilities to save money by relaxing dry cask storage designs, storing the fuel at more sites

and allowing the fuel to be stored in the same dry cask for longer periods of time. The

realization of margin will also assist with public perception, which is a significant hurdle

in any nuclear licensed application.

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2. Background

Nuclear power plants produce energy through the nuclear fission process, which

produces waste as a byproduct. The waste comes in the form of spent nuclear fuel rods.

Every 12-18 months plants refuel and replace some portion of the reactors used fuel

supply with new fuel. The spent (or used) fuel is then placed into a spent fuel pool, on

site, at the plant. The fuel remains in the pool, under about 20 feet of water for a number

of years, depending on a few variables. Those variables are listed below, and described

in the following sections:

Amount of residual heat removal required

Storage capacity on site

Regulatory requirements

Depending on what variables are in play at a particular plant, the necessity for

keeping the spent fuel in the spent fuel pool may no longer be present. In cases in which

wet storage is no longer an option the fuel can be placed into dry cask storage containers

and either stored on site, or shipped to another storage site. As with spent fuel pool

requirements, there are many Nuclear Regulatory Commission (NRC) requirements

dealing with the storage of fuel in such a manner. It should be noted that there are

regulations on all types of nuclear waste, and the regulations and storage methods

described in the following sections deal specifically with spent nuclear fuel.

2.1 Classification of Spent Nuclear Fuel (SNF)

SNF from a nuclear reactor is classified as high-level waste. High activity leads to

dose rates greater than 50 mrem/hr. This being the case, the regulations discussed focus

on the laws dealing with high-level waste. It may at a future date be reprocessed to

lower activity wastes.

2.2 Spent Fuel Storage Descriptions

Currently, most plants do not recycle spent fuel, and either store the spent fuel rods

in wet or dry conditions. The wet conditions exist in spent fuel pools on the plant site,

while dry cask storage can exist either on site, or at an alternate location. The Yucca

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Mountain facility aimed to meet this dedicated dry cask storage facility need for the

industry.

2.2.1 Spent Fuel Pool (Wet Storage)

Spent fuel pools are large pools of water located within the nuclear power plant site

boundaries, within restricted areas in the fuel handling building. The pools are often

strategically placed to avoid threat of terrorist attacks. The pools are typically 45 feet

deep and cover the spent fuel rods with more than 20 feet of water. The spent fuel in the

pool is organized to maximize cooling mechanisms and to minimize safety concerns.

Additionally the plants are designed to provide a number of backup water supplies to the

spent fuel pool. Figure 1 shows a typical spent fuel pool design and plant location.

Figure 1 Typical Spent Fuel Pool Design and Location []

When plants were originally designed, utilities only intended to store a few years of

spent nuclear fuel in the spent fuel pools. The intent was to send sufficiently cooled fuel

to reprocessing facilities, however this never occurred domestically. Although analyses

have since been performed to verify that more fuel than the original designs intent can

be stored in the pool, (i.e., re-racking [], increasing cooling supplies, etc.) the need for

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alternate fuel storage is apparent. Figure 2 demonstrates the dire state the industry will

be in if an alternate storage facility is not developed.

Figure2 Spent Fuel Pool Capacity Prediction []

The figure clearly demonstrates that a large number of spent fuel pools will have

been filled to capacity in the very near future. Although more spent fuel pools can be

built, the time, money and resources to do so far exceed the benefit. Rather, a more

attractive approach is likely utilizing the existing technology of dry cask storage, as this

is a much more affordable option.

2.2.2 Dry Cask Storage

Typical dry casks are metal cylinders, sealed through welding techniques or simply

bolted closed. The casks are surrounded again by a metal or concrete outer shell. The

casks are built to be air tight. The casks are designed to rest with the fuel sitting either

vertically or horizontally in the enclosure. Figure 3 and Figure 4, show typical storage

orientations. Figure 5 displays some typical dry storage cask dimensions. Attachment A

contains a list of NRC approved dry cask designs.

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Figure3 Typical Dry Cask

Design []

Figure4 Dry Cask HorizontalOrientation Design []

Figure5 Typical Dry Cask Dimensions []

Figure 6, is another depiction of a dry cask container, where the radiation shields

have been identified [].

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Figure6 Dry Cask Radiation Shield []

The spent fuel rods are loaded into the dry casks after sufficient heat has been

removed from the rods in the spent fuel pool. When they are transferred to the dry casks

the water and air are removed from the canisters and they are filled with an inert gas.

The casks rest on concrete pads or in storage bunkers, as shown in the above figures.

The dry casks are designed to withstand extreme conditions, both due to weather and

terrorism threats. The casks typically store between two and six fuel assemblies

(depending on the assembly design). The NRC reports that typically, the maximum

heat generated in an hour from 24 fuel assemblies stored in a cask is less than that given

off by a typical home heating system for the same amount of time [ ]. This provides an

indication of the magnitude of the heat production from the fuel rods in the casks.

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Figure 7 identifies general functions of a Mitsubishi spent fuel dry storage cask.

Figure 8 depicts radiation protection mechanisms present in the dry cask and dry cask

heat removal mechanisms as well as cask safety functions [].

Figure7 Dry Cask General Functions []

Figure8 Dry Cask Safety Functions []

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The casks can also often be used for transportation, and sometimes plants place the

spent fuel into casks so that they may be shipped to dedicated dry casks storage

locations. Figure 9 identifies spent fuel storage locations, although not all are dry

storage locations.

Figure 9 Spent Fuel Storage Locations in the United States of America []

The following analysis will specifically focus on dry cask storage, as opposed to wet

fuel storage due to cost concerns. It is much more financially feasible to develop an

onsite (or offsite) dry cask storage facility, as the safety regulations are much more

relaxed for the dry cask storage facilities. Spent fuel pools must also be built inside the

reactor facilities boundaries, while the dry cask storage facility may be located outside

the facility walls.

2.3 NRC Spent Fuel Requirements

Given the severity of an accident associated with nuclear waste, the NRC is diligent

with the licensing and operational processes dealing with spent nuclear fuel storage and

treatment. At the highest level, the regulation begins with the definition and regulation

of the spent fuel storage sites, whether the site is for a spent fuel pool or a dedicated site

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for dry casks. This analysis will specifically focus on Independent Spent Fuel Storage

Installations (ISFSIs) as they relate to the storage of spent fuel in dry casks in offsite

locations.

Licensing requirements for spent fuel sites are endorsed through 10 CFR 72,

Licensing Requirements for the Independent Storage of Spent Nuclear Fuel, High-

Level Radioactive Waste, and Reactor-Related Greater than Class C Waste. This

document contains 12 subparts, each subpart with a number of subsections. Certain

subsections apply more specifically to the storage of spent fuel rods offsite in dry casks.

Some of the most applicable sections will be described in the following paragraphs as

they provide necessary background describing design constraints and operational

procedures.

Section 10 CFR 72.106 [] Controlled area of an ISFSI or MRS defines the area

required for an ISFSI. The section sets dosage limits as well as public accessibility

limits. Section 10 CFR 72.106 Criteria for radioactive materials in effluents and direct

radiation from an ISFSI or MRS, also sets radiation limits for the spent fuel storage

locations. Section 10 CFR 72.108 [] Spent fuel of high-level radioactive waste

transportation, deals specifically with waste transportation.

Section 10 CFR 72.236 [] Specific requirements for spent fuel storage cask

approval and fabrication, identifies the specific regulations that the technical analysts

must consider when proposing a dry casks storage facility. Those specific requirements

are shown below as they appear in 10 CFR 72.236 []:

The certificate holder and applicant shall ensure that the requirements of this section

are met.

(a) Specifications must be provided for the spent fuel to be stored in the spent

fuel storage cask, such as, but not limited to, type of spent fuel ( i.e., BWR,

PWR, both), maximum allowable enrichment of the fuel prior to any

irradiation, burn-up (i.e., megawatt-days/MTU), minimum acceptable cooling

time of the spent fuel prior to storage in the spent fuel storage cask, maximum

heat designed to be dissipated, maximum spent fuel loading limit, condition

of the spent fuel (i.e., intact assembly or consolidated fuel rods), the inerting

atmosphere requirements.

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(b) Design bases and design criteria must be provided for structures, systems, and

components important to safety.

(c) The spent fuel storage cask must be designed and fabricated so that the spent

fuel is maintained in a subcritical condition under credible conditions.

(d) Radiation shielding and confinement features must be provided sufficient to

meet the requirements in 72.104 [] and 72.106 [].

(e) The spent fuel storage cask must be designed to provide redundant sealing of

confinement systems.

(f) The spent fuel storage cask must be designed to provide adequate heat

removal capacity without active cooling systems.

(g) The spent fuel storage cask must be designed to store the spent fuel safely for

a minimum of 20 years and permit maintenance as required.

(h) The spent fuel storage cask must be compatible with wet or dry spent fuel

loading and unloading facilities.

(i) The spent fuel storage cask must be designed to facilitate decontamination to

the extent practicable.

(j) The spent fuel storage cask must be inspected to ascertain that there are no

cracks, pinholes, uncontrolled voids, or other defects that could significantly

reduce its confinement effectiveness.

(k) The spent fuel storage cask must be conspicuously and durably marked with--

(1) A model number;

(2) A unique identification number; and

(3) An empty weight.

(l) The spent fuel storage cask and its systems important to safety must be

evaluated, by appropriate tests or by other means acceptable to the NRC, to

demonstrate that they will reasonably maintain confinement of radioactive

material under normal, off-normal, and credible accident conditions.

(m) To the extent practicable in the design of spent fuel storage casks,

consideration should be given to compatibility with removal of the stored

spent fuel from a reactor site, transportation, and ultimate disposition by the

Department of Energy.

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(n) Safeguards Information shall be protected against unauthorized disclosure in

accordance with the requirements of 73.21 and the requirements of 73.22 or

73.23 of this chapter, as applicable.

NUREG-1567 [] defines the spent fuel requirements for spent fuel at dry storage

facilities. NUREG-1567 states that:

Fuel cladding temperature during dry storage shall be maintained below the

expected damage-threshold temperatures for normal conditions and a minimum of 20

years dry storage for ISFSI or MRS design and environmental conditions. The fuel

cladding temperature should also generally be maintained below 570C (1058F) for

short-term off-normal, short-term accident, and fuel transfer operations (e.g., vacuum

drying of the cask or dry transfer) [].

This temperature requirement is more applicable to immediate dry storage, and not

long term sites such as Yucca Mountain.

Section 4.4.1.1 of NUREG-1567 [] identifies more spent fuel requirements, and

describes how a utility must document their intentions for fuel storage. Additional

guidance for this is provided in 10 CFR 72.120. These documents also describedifferences in treatment for different types of spent fuel (fuel classes, damaged fuel,

etc.).

Section 5.4.2 of NUREG-1567 illustrates that the ISFSI must be designed to operate

safely for all design functions. This implies that if a spent fuel storage facility is used

for only the storage of fuel, and not fuel transfer (as a spent fuel pool would be used) the

facility needs to only be designed as such. Therefore, the requirements for an off site

dry cask storage facility may be slightly relaxed if the fuel being stored there is safer

(i.e., lower temperatures, better condition) than fuel stored on site. Yucca Mountain

would specifically fit into this category (the fuel stored at Yucca Mountain would be

only spent fuel that has already been sufficiently cooled, and doesnt store any damaged

fuel).

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2.3.1 Storage Facilities

NUREG-1567 [] defines and identifies requirements for spent fuel storage facilities.

NS-G-1.4 [] defines storage facility requirements for spent fuel storage. Section 2.6

identifies that the storage facility must:

provide for the safe and secure storage of fuel from the time of its removal from the

reactor until such time as it is transferred and loaded into a spent fuel cask for

transport away from the reactor site for disposal as radioactive waste or for

reprocessing. The facility will therefore include systems for the handling, storage,

transfer and retrieval of the spent fuel. The primary safety functions of these systems

should be to ensure that the fuel is maintained subcritical at all times, that the

integrity of the fuel cladding is preserved, that the fuel is adequately cooled to

remove residual heat, that radioactive material is contained, and that there is no

undue risk to health and safety or hazard to the environment.

In the context of the previous paragraph, subcritical (mass of active nuclides) refers

to the fact that the spent fuel can no longer sustain a nuclear chain reaction.

2.3.2 Dry Cask Requirements

NS-G-1.4 [] also identifies the requirements for the dry cask storage devices.

Section 2.7 identifies the requirements as follows:

Irradiated fuel should be transported in shielded and adequately cooled casks that

are either internally dry or partially filled with coolant. The casks should have an

internal structure to keep the fuel in a well defined geometric arrangement during

transport. The casks should be loaded either under water in a specific area at thestorage pool or in a separate cask loading pool, or they should be loaded dry. The

fuel may first be placed in a basket which may then be loaded into the cask. The

systems for cask handling should be such as to ensure that the casks can be received,

loaded and prepared for transport, either on the site or off the site, in such a manner

that they meet the applicable requirements.

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Section 6.0 of NS-G-1.4 [] also provides further guidance for both the transportation

and handling of the dry casks.

2.4 Spent Fuel Requirements

To further characterize the spent fuel temperature requirements at Yucca Mountain,

the EPA requirements for dosage limits allowable at that site were defined. On

September 30, 2008 the EPA issued the following radiation standards for the Yucca

Mountain site:

The EPA is required to set standards consistent with the findings and

recommendations of the National Academy of Sciences (NAS) and satisfy a July

2004 court decision to extend the standards' duration. The Yucca Mountain

standards are in line with approaches used in the international radioactive waste

management community. The final standards will:

Retain the dose limit of 15 millirem per year for the first

10,000 years after disposal;

Establish a dose limit of 100 millirem annual exposure per

year between 10,000 years and 1 million years;

Require the Department of Energy (DOE) to consider the

effects of climate change, earthquakes, volcanoes, and corrosion of

the waste packages to safely contain the waste during the 1 million-

year period; and,

Be consistent with the recommendations of the NAS by

establishing a radiological protection standard for this facility at the

time of peak dose up to 1 million years after disposal.Human exposure to radiation varies from natural sources, such as radon and

ultraviolet radiation from the sun, and other sources, such as medical X-rays. The

average annual radiation exposure from both naturally occurring and manmade

sources for a person living in the United States has been estimated to be 360

millirem per year.

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With this information, analysis will determined whether or not these requirements

are appropriate (in terms of safety) for the Yucca Mountain proposal. These

requirements however, while specific to Yucca Mountain, are only used for site

licensing.

The generic requirement for spent fuel storage defined by the NRC comes from

NUREG-1536 [], which states that the fuel in the dry casks may only be allowed a

cladding failure probability of less than 0.5%. Therefore, further analyses and data

utilized in this project will adhere to this probability limit.

2.5 Failed Dry Cask Impact

Rules and regulations instilled on the storage of spent fuel in dry casks are put in

place for good reason. The impacts of failed fuel and failed storage casks have

significant impacts, and can potentially release very dangerous materials. Spent fuel rod

failure involves failure of the fuel rods cladding, which surrounds the fuel. This

cladding is made from metal alloy tubes. In the cases discussed in this report, the fuel is

stored in Zircaloy tubes. The following sections will discuss the impact of failed dry

cask storage devices, however it should be mentioned that this document focuses only on

the failure of the spent fuel rods, and their cladding failure. The cladding failure serves

as the primary barrier of the fissile materials to the environment, while the dry casks

themselves serve as the secondary barrier. For analyses on the failure rates and

mechanisms of the dry casks themselves, refer to technical publication paper M04-6,

presented at Transactions of the 17th International Conference on Structural Mechanics in

Reactor Technology [].

The radiation released from a nuclear reaction can be predicted and defined when the

nuclear fission reaction is examined. A general governing equation for a nuclear

reaction in a PWR is shown below in Equation 1:

Uranium-235Fission Product A + Fission Product B + Energy + Neutron (1)

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This same equation can also be represented in terms of the actual fission products

released. This is shown in Equation 2:

1 Neutron + Uranium-235Barium-141 + Krypton-92 + Energy + 2 Neutrons (2)

The energy released will include the different radiation types, which can be

measured and will be discussed in the following sections []. The neutron particle can

further subdivide and react in the nuclear reaction, either bombarding another element to

further subdivide, or bring different elements and particles in the reaction to steady state.

Equation 3 identifies the neutrons makeup. The neutrino will represent much of the

energy released from the nuclear reaction.

NeutronProton + Electron (Beta Particle) + Neutrino (3)

The previous equations specifically discuss the nuclear fission reaction however, and

this is not the leading cause for radiation released from failed spent nuclear fuel. At this

point in the fuels life cycle (spent fuel), the fuels elements are decaying, and as a result

releasing heat and radiation. Any nuclide (element) that is produced from the fission

reaction and is unstable, will continue to decay until a stable state is reached. Two of the

most abundant elements present in the spent fuel are Uranium-235 and Uranium-238.

Table 1 and Table 2 depict the decay chains for these two nuclides, and describe the

radiation type. (Note, every decay state emits gamma radiation.) In both cases Pb (lead)

is the stable state for the decay chain []. Due to the fact that nuclide decay produces

three different types of radiation (alpha, beta and gamma), the spent fuel dry casks must

be equipped with multiple levels of radiation protection. These levels of radiation

protection include the fuel cladding, the steel cask wall and the concrete cask wall.

Table 1: The Uranium-238 Decay Chain

Nuclide Half-Life (t) Radiation Type

U-238 4.468x109 years alpha,

Th-234 24.1 days beta,

Pa-234 1.17 minutes beta,

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Table 1: The Uranium-238 Decay Chain

Nuclide Half-Life (t) Radiation Type

U-234 244,500 years alpha,

Th-230 77,000 years alpha,

Ra-226 1,600 years alpha,

Rn-222 3.8235 days alpha, Po-218 3.05 minutes alpha,

Pb-214 26.8 minutes beta,

Bi-214 19.9 minutes beta,

Po-214 63.7 microseconds alpha,

Pb-210 22.26 years beta,

Bi-210 5.013 days beta,

Po-210 138.378 days alpha,

Pb-206 Stable N/A

Table 2: The Uranium-235 Decay ChainNuclide Half-Life (t) Radiation Type

U-235 7.038x108 years alpha,

Th-231 25.52 hours beta,

Pa-231 32,760 years alpha,

Ac-227 21.773 years beta,

Th-227 18.718 days alpha,

Ra-223 11.434 days alpha,

Rn-219 3.96 seconds alpha,

Po-215 778 microseconds alpha,

Pb-211 36.1 minutes beta,

Bi-211 2.13 minutes alpha,

Tl-207 4.77 minutes beta,

Pb-207 Stable N/A

The half-life (t) indicates the time required for the concentration of an active

nuclide to become half its original abundance.

2.5.1 Radiation Releases and Protection

Failure of the dry cask storage devices, coupled with fuel cladding failure will result

in radiation releases. The two main types of radiation releases resulting from spent fuel

failures are gamma ray radiation and neutron radiation []. Gamma rays are

electromagnetic radiation which can be produced from a number of different atomic

reactions. With respect to this analysis they occur as a byproduct of the nuclear fission

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reaction. The alpha (), beta () and gamma () rays can be blocked by thick structures

such as concrete or lead, as well as depleted uranium. A safe distance from these

radiation types would depend on the abundance of the radiation types, and the protection

between the source and the individual in question.

Neutron radiation occurs from similar types of reactions as gamma rays, however it

consists simply of released neutrons from the fission reaction. Materials most effective

in blocking neutron radiation include concrete type substances, as well as water.

2.5.2 Public Impact

Both types of radiation can be severely hazardous, with neutron radiation being the

more severe of the two most major types. They can cause cancer, with the severity and

occurrence of the disease being proportional to the amount of radiation an individual is

exposed to. The amount of radiation released would be a function of the number of fuel

casks damaged, the age of waste present in the casks and the methods by which the

fissile materials were dispersed throughout the surrounding communities. These

dispersion mechanisms would include ground water, river water, sea currents and wind

patterns. The EPA requirements for radiation released from the Yucca Mountain site

(see Section 2.3) can shed some light as to what is anticipated from the Yucca Mountain

site.

Although the amount of radiation dosage encountered from an accident depends on

the severity of the accident, and the proximity of the victim, there can be health effects.

Table 3 lists some radiation exposure examples.

Table 3: Typical Radiation Exposures

Dosage Level Duration Description

360 mrem Annual Average USA dosage from all sources.

5000 mrem Annual USA NRC occupational limit.10000 mrem Acute SNL analyzed study exposure amount.

17500 mrem Annual Guarapari, Brazil natural radiation exposure.

25000 mrem Annual Significant increase in health effects.

500000 mrem Annual Probable death expected.

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Clearly, in all cases, the more concentrated the exposure to the same dosage level the

more damaging the exposure would be. Sandia National Laboratories performed a study

in which it was theorized that a spent fuel cask failed open while being transported, and

all the spent fuel inside had been oxidized (fully failed), thus resulting in the worst

possible accident. Although this example only resulted in a failure of one cask, this still

provides a good example of the potential for spent fuel dry cask failure. The

consequences of the analysis are summarized below: [] and []

The maximally exposed individual receives a dose of about 10.2 rem, primarily

from inhalation of radionuclides. The maximally exposed individual is an

emergency responder located 70 meters directly downwind from the point of

release for a period of "a few hours" ("...no protective equipment is worn and no

attempt is made to avoid inhalation of radionuclides in the atmosphere.") The

10.2 rem dose "is considered to have no consequence other than a possible small

increase in the probability of incurring cancer in later years."

The radionuclides released by the accident are carried downwind ( in the plume

smoke from a petroleum fire) and contaminate an area of about 110 square

kilometers (42.5 square miles) at levels of 0.2 microcuries per square meter or

greater. Three radionuclides - Co-60, Cs-134, and Cs-137 - account for over 99percent of the activity deposited on the ground. [Kr-85 and other radioactive

noble gases are assumed to dissipate harmless.] Within the contaminated region,

an area of about 2.2 square kilometers (0.9 square miles) is contaminated at

levels of 10 microcuries per square meter; an area of 4.3 square kilometers (1.7

square miles) is contaminated above 5 microcuries per square meter.

The release from the worst case rail accident in a typical urban area results in "22

latent health effects" (about half cancers, and half genetic disorders). In a rural

area, "the same accident could result in about 0.035 latent health effects." The

exposed urban and rural populations would be expected to "experience about

470,000 and 730 cancer fatalities, respectively, from all other causes in the same

time period. Clearly, the severe but credible rail cask accident does not contribute

significantly to the number of cancer fatalities in the region."

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Cleanup of the contaminated area, assuming that the accident occurs in a rural

area, could cost as much as $620 million and require 460 days. Sandquist

assumes cleanup "to a level that reduces individual dose rates from deposited

radionuclides down to a maximum value of 500 mrem/yr." In an urban area, the

cleanup cost is estimated to exceed $2 billion.

This study provides evidence to the severity such an accident could have on the

surroundings. Although the described accident is a worst case situation, it should be

noted that this type of release would be severely unlikely, and even less likely if failure

occurred within the storage facility. This being the case, the fact that the fuel must be

transported creates the possibility for a similar occurrence of the previously described

scenario.

Radiation release of this magnitude would also have an impact on the surrounding

environment. The radiation would travel through wind currents as well as any present

water sources, including ground water. Low levels of radiation would have an impact on

vegetation in the surrounding area, with impact following onto all affected in the food

chain. Animals and other wildlife would experience similar type of reactions as humans,

with the most significant impacts occurring with high dosage exposure during short

amounts of time.

2.6 Spent Fuel Failure Mechanisms

The following sections identify the main mechanisms by which the spent fuel may

fail. The four main fuel cladding damage mechanisms are Delayed Hydride Cracking

(DHC), Stress Corrosion Cracking (SCC), Accelerated Creep leading to Creep Rupture

(CR) and thermally activated micro-cavitation leading to failure by Diffusion Controlled

Cavity Growth (DCCG).

2.6.1 Delayed Hydride Cracking (DHC)

DHC is a small crack growth mechanism, where a brittle hydride forms at the top of

the crack, and forces the crack to grow larger. A brittle hydride is essentially the

formation of cracking due to the presence of hydrogen. This is a common phenomenon

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in many metallic structures []. Figure 10, from IAEA-TECDOC-1410, depicts the

formation of DHC.

Figure 10 DHC Formation []

The top figure shows a formed crack at the base of the metal. The middle figure

highlights a hydride forming at the top of the crack, and the bottom figure demonstrates

the growth of the crack due to the presence of the hydride.

2.6.2 Stress Corrosion Cracking (SCC)

SCC is a failure of normally ductile metals due to tensile stresses. This failure

mechanism occurs suddenly and is accompanied by a sudden temperature increase. Thisis a highly chemically dependent failure mode, and is not an especially significant

contributor to spent fuel rod damage.

2.6.3 Accelerated Creep Leading To Creep Rupture (CR)

Creep rupture in spent nuclear fuel can occur as a result of two different types of

failure mechanisms. The first mechanism is simply the materials tendency to slowly

deform over a period of time, followed by a short period of transient creep. The secondmechanism is the spent fuel rods response to decreasing temperature and pressure. The

second mechanism is the leading contributor to spent fuel rod rupture [] and [].

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2.6.4 Diffusion Controlled Cavity Growth (DCCG)

DCCG is failure caused by thermally activated micro-cavitation []. This failure

mechanism is not believed to be a significant contributor to spent fuel rod failures. This

is mostly due to the fact that the mechanism is unlikely to occur if fuel temperatures arebelow 400C (the NRCs dry cask storage temperature limit) [].

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3. Temperature, Stress and Age Studies on Spent Nuclear Fuel Storage

3.1 Spent Fuel Safety Concerns

The preceding sections identified requirements outlined by the NRC for spent fuel

storage, specifically focusing on dry cask storage at an off site location (Yucca

Mountain). While the NRC does give general temperature recommendations, the only

quantifiable requirement comes from the EPA, which recommends the dosage release

requirements for that specific site.

The following analysis will investigate the reasonableness of the temperature and

fuel age storage requirements for the spent nuclear fuel when stored in dry casks. This

will be done by identifying the spent fuel storage safety concerns, and the mechanisms

by which these safety concerns are mitigated. This analysis will focus on the failure of

spent fuel being stored within the dry casks.

3.2 Spent Fuel Rod Temperature Requirements

BTM 03-126 [] investigates each of the previously discussed damage mechanisms

and has determined (through testing) that the most significant damage mechanism is

creep rupture. BTM 03-126 has also determined (through testing) that creep strain is a

function of fuel rod design, the irradiation history, and the dry storage conditions(specifically temperature).

Therefore, temperature analyses will be investigated to determine whether or not the

requirements identified for the Yucca Mountain proposal contain excess margin.

According to BNL-52235 the NRC temperature requirements for dry storage of spent

fuel are set at a temperature limit of 400C (752F) [].

Significant investigations have been performed to determine what the pressure vs.

temperatures limits which cause creep strain are (note, creep strain is the leading cause

for cladding failure in dry cask storage). Note, EPRI-1001207 determined via testing the

creep strain limits most appropriate for varying temperature and pressure limits. These

strain limits define what percentage of the individual fuel rod can fail due to creep strain.

It is with these limits that the temperature and pressure relationships can be derived [ ].


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Figure 11, Figure 12, and Figure 13 from EPRI-1001207, which were used to determine

creep strain limits, are shown below:

Figure 11 Cladding Strain vs. Time at 150 MPa []

Figure 12 Cladding Strain vs. Time at 200 MPa []


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Figure 13 Cladding Strain vs. Time at 380C []

There are a number of data sources that will be analyzed to determine what current

industry experience and scientific testing demonstrates as the need for spent fuel dry

cask storage temperatures and storage conditions. The following sections will discuss

each of the various data sources, the type testing or data gathering that occurred, and

what the applicable results were.

3.2.1 PNL-10813 Study

PNL-10813 [] determined (via testing) that typically, fuel temperature upper bounds

are at about 380C (approximately the NRC fuel temperature limit requirement). (The

temperature limit and temperature analyses were performed on spent fuel with an

expected maximum burnup of 60 GWd/MTU and subjected to gamma fields of about

105 R/h with a neutron flux of about 104 to 106 n/cm2-s. The fuel is enriched Uranium

235, Zircaloy tube bundles, which are typical characteristics for Light Water Reactors

(LWRs)) PNL-10813 demonstrates, using the heat decay curve from PNL-6364, [] thatthis temperature decreases to around 100C in ten years, at which point this temperature

slowly decreases over the following 100 years. Figure 14 from PNL-6364 [] depicts

cladding temperature limits for spent fuel at five years (earliest time period at which

spent fuel is typically moved from spent fuel pool storage to dry storage).


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PNL-6364 developed a fuel cladding failure model, and determined the relationship

between fuel temperature and stress levels.

Figure 14 Cladding Stress vs. Temperature []

PNL-6364 demonstrates that for fuel at 400C the cladding stress limit is 50 MPa.

This value rises significantly as the fuel cools, and rises to over 100 MPa when the fuel

is cooled to 380C, which PNL-6364 points out, is a more typical starting temperature

for spent fuel in dry storage conditions. (Note, these stress limits were determined using

the fuel rod failure limit identified by the EPA and NRC, which is less than five fuel

rods in 1000, or 5%.)Figure 15 also identifies that this trend remains true for fuel aged more than five

years.


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Figure 15 Cladding Stress vs. Temperature (Aged Fuel) []

The following illustrates some sample calculations which are representative of the

type of analysis performed in developing the detailed modeling found in PNL-6364.

( ) ( )

000,10

684.69

2 1

2=

T

T

t

DP midmhoop (4)

( ) ( )

( ) ( ) ( ) 684.69000,1021

2 =

mid

mhoop

DP

tTT (5)

Where,

mhoop = dry storage cladding hoop stress, MPa

P = internal gas pressure of the rod, psi

T1 = temperature at which P was determined, K

t = cladding wall thickness, in.

Dmid = cladding midwall diameter, in.

= a factor, 0.95 for PWR rods or 0.90 for BWR rods

10,000/69.684 = conversion factor

T2 = allowable storage temperature for mhoop , K


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Plugging in the following typical PWR spent fuel rod generic characteristics yields a

cladding temperature limit of 393C and a stress limit of 69.2 MPa.

mhoop = dry storage cladding hoop stress, MPa

P = 942.721 psi

T1 =298 K

t = .0394 in.

Dmid = .394 in.

= 0.95

10,000/69.684 = conversion factor

T2 = allowable storage temperature for mhoop , K

Figure 16, from PNL-6364, identifies this point on their modeled chart. (It should be

noted that other studies utilize these methods, or methods very similar for their

temperature vs. stress relationship modeling.)

Figure 16 Cladding Stress vs. Temperature (Example) []


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PNL has a cladding stress vs. temperature relationship for spent fuel failing at a 5%

failure rate. PNL determined that although the NRC required that the fuel be stored at

less than 400C, fuel was typically below this temperature limit when it exited the spent

fuel pool (typically 380C). Therefore, fuel subjected to the same cladding stress would

exhibit less cladding failure than previously believed.

3.2.2 CANDU Reactors

IAEA-TECHDOC-1100 reports that over 200 dry cask containers have been stored

for 18 years without any degradation reported. This fuel was stored at 150C however,

as the CANDU fuel can be held at much lower temperatures and is required to be stored

below 160C []. Although these fuel types are not exactly the same, the operating

experience with no degradation reported supports the argument that this is an acceptable

storage technique.

3.2.3 German Study

German plants have stored 305 casks (storing both PWR and BWR type fuel) at a

storage facility in Ahaus Germany. To date, no problems have occurred. This fuel is

only required to be held to a temperature limit of 410C and has had no reported

incidents [] and []. This provides further indication that margin exists in the current

regulatory environment. PNL-10813 reports that Germany has been conducting full

scale dry storage cask demonstrations with irradiated PWR and BWR fuel since 1982.

Tests lasted up to two years, at which time cladding creep stopped due to pressure,

temperature, and decay heat drops in the fuel rod. This provides further indication of

the importance placed on the safety of this fuel and further validates their results. The

German studies also demonstrated that spent fuel could be safely stored in a dry cask of

at least 40 to 50 years [].


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Figure 17 Dry Casks in Storage at the Ahaus, Germany Site []

Figure 17 is a representation of the dry casks held in storage at the Ahaus, Germany site.

3.2.4 PNL-6189 Study

PNL also performed a study where they investigated the relationship between fuel

age and temperature. The study determined that fuel cools rapidly initially,

demonstrating (when compared with PNL-6364 pressure limits), that storing fuel at high

initial temperatures, will allow for safe storage. PNL-6189 determined that the spent

fuel cools about 280C the first five years the fuel is stored in a dry cask, which puts the

stress levels of the fuel extremely low when compared to the PNL-6364 stress limits [].

The main point of the analysis however, was to identify that temperature alone could not

be used as the only grounds for regulating the spent fuel storage. The failure criteria

needed to be developed as a combination of temperature and fuel age. Simply stated,

five year old fuel at 350C is very safe to store in a dry cask while 20 year old fuel at

350C may not be nearly as safe. The conclusion of the study was that such generic

restriction of temperature limits are not appropriate and may be too restrictive (i.e., the

fuel temperature required may be too low) for new fuel. PNL-6189 states:


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The variance in fuel age that exists in the population of spent fuel leads to the

strong indication that a single-valued dry storage temperature limit, which would

need to be conservative to account for old fuel, imposes unnecessarily conservative

temperature limits on fresher fuel, [].

depicts the temperature vs. allowable stress levels determined through testing.

Figure 19 is a representation of the temperature decay vs. time curves. It demonstrates

the effectiveness of their model as it compares their model to collected data for a LWR

spent fuel cask type.

Figure 18 Temperature vs. Allowable Stress Limits []


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Figure 19 Temperature Decay Rate []

Table 4: Temperature vs. Stress Limits for Varying Fuel Ages []

From Table 4 it is clear that the higher the age of the spent nuclear fuel, the lower the

allowable storage temperature.


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3.2.5 PNL-SA-13879 Study

PNL-SA-13879 also performed similar analyses as the previously discussed studies.

PNL-SA-13879 determined (via testing and modeling) that fuel which had only spent ayear in a spent fuel pool could be stored at initially higher temperatures than fuel which

had been stored in a pool for five years. Figure 20 is a representation of this. Figure 20

shows that although cladding failure for fuel spending less time in wet storage occurs

earlier than older fuel, the amount of damaged fuel is much lower for the newer fuel [].

Figure 20 Damage Accumulated in Dry Storage []

This alludes to the fact that it may not be appropriate to require one single spent fuel

dry cask storage temperature for SNF of different ages. This analysis also does not takeinto account the fact that even if spent fuel is kept in wet storage longer, it will likely be

at a lower temperature when it is removed, rather than being maintained at the same

temperature as fuel that has only been in wet storage for one year. Both of these factors

(i.e., stress vs. temperature, fuel damage vs. storage age) support the argument that the

current restrictions on spent fuel dry storage may not be appropriate.


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3.2.6 Domestic Support

The United States has also demonstrated successful operating experience with spent

fuel dry casks storage. Two such sites, Surry and HB Robinson, have stored spent fueldry casks since 1985, with 29 casks stored at the Surry site and eight at the HB Robinson

site [] and []. These sites reported no operating failures while abiding by the same

regulations and environment as analyzed in this document.

3.2.7 Operating Locations

The following two tables summarize industry operating history, in which no casks

have been reported to have failed []. The UK site, has been operational for more than 20

years, which is the NRC license duration.

Table 5: Vault Storage Dry Cask Operating Experience []

This further supports the stance that the current industry requirements for spent fuel

are too stringent, and that there is available margin in the current regulatory

environment.


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Table 6: Outdoor Storage Dry Cask Operating Experience []


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Table 5 and Table 6 serve to highlight the extensive number of dry cask storage

facilities, located both domestically and abroad.

3.2.8 Hanford Engineering Development Laboratory

Hanford Engineering performed similar evaluations to the various PNL studies. The

Hanford studies resulted in a maximum cladding temperature recommendation of 380C,

and 396C for short term storage applications. Although these temperature limits are

lower than the other studies conclusions, these conclusions are due to conservative


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assumptions imbedded in the analysis, which if considered appropriately would produce

results more consistent with the previously discussed studies.

Hanford identifies that the most conservative inputs available are used in their

evaluation, and no rod failures are acceptable in the analyses (0.5% is typically

acceptable by NRC regulations). Additionally the studies do not account for the rapid

temperature decrease of the fuel rods, which is evident, based on the claim that the

higher temperature limit may only be used for short term storage. It should also be noted

that this study was performed in 1978, and has been included to demonstrate the

evolution this topic has taken in recent years [].


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4. Analysis

To support the industry need for storage of SNF, the Yucca Mountain storage facility

site was proposed. A concern surrounding this issue however, was the safety associated

with the storage conditions of the SNF. The Yucca Mountain proposal was cancelled,

and although specific details were not released, safety was certainly a major concern. As

this analysis has examined current SNF storage conditions, it has been determined that

there is available margin present in current industry regulator requirements.

The current SNF storage conditions specifically examined in this analysis were as

follows:

1. SNF storage temperature (currently the NRC requires a storage temperature

of less than 400C)2. High-level waste with active Uranium-235 and Uranium-238 decay

3. Temperature vs. stress relationships (which leads to creep rupture)

4. Affect of fuel age on temperature vs. stress relationship

Figure 21 shown below, depicts a cross section representative of the SNF examined

in this study.

Figure 21 PWR Nuclear Fuel Rod Assembly []

A study of the literature suggests that there is margin present in the current NRC

regulatory environment for the storage of spent nuclear fuel in dry casks. The preceding


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data has supported this claim, and the margin available will be further discussed in the

following section.

4.1 Spent Nuclear Fuel Margin Available

Current NRC requirements do not consider the age of the SNF in their licensing

requirements. The requirements only state that all fuel must be stored at less than

400C, regardless of SNF age.

To determine more appropriate regulatory requirements, industry tests and nuclear

fuel storage facilities operating experiences will be studied. By examining these data

sources, more appropriate storage requirements will be identified. Industry temperature

vs. stress tests performed on SNF by PNL, German researchers and Hanford Engineering

were investigated. Additionally, operating experience from Canada, Germany and the

US demonstrated impeccable operating histories under current industry requirements and

were examined through this study.

Through the study of these sources, it has been determined that current requirements

are too restrictive, and do not appropriately consider SNF age. These studies have

shown that the SNF can be safely stored at temperatures below the current NRC

licensing regulations, and that the age of the fuel plays a significant role in determining a

safe storage temperature.

This analysis identifies the leading mechanism causing the failure of a spent fuel rod

(temperature induced creep) and demonstrates that there is significant margin available

between NRC requirements and typical operating levels. The analyses investigated the

possibility of significantly reducing risk for spent fuel rod rupture (as cladding stress due

to temperature conditions is the most significant damage mechanism). The NRC

requires that the spent fuel temperature be analyzed at 400C; however this temperature

is quickly reduced by spending more time cooling in the dry casks (or in wet storage).

More time cooling, as illustrated in Figure 15, would result in more relaxed cladding

stress limits. Additionally, the NRC is dictating requirements based on spent fuel at

initial conditions, analyzed at 400C, which, as PNL-6364 demonstrated, can be rapidly

reduced and is realistically at a lower starting point initially regardless.


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Furthermore, numerous utilities and storage facilities have demonstrated flawless

storage histories. Canada, Germany and many US sites, support impeccable fuel storage

histories. The demonstration of this successful spent fuel dry cask storage history is

further evidence alluding to the success that this storage method has had.

It was further demonstrated that fuel which had been moved from wet storage to dry

storage early in its lifespan can withstand much greater stress limits at similar

temperature limits than its aged counterparts. It has been recommended that rather than

setting one standard temperature requirement, it would be more appropriate to have

different requirements for different ages of fuel, as the current regulation is too

restrictive for relatively new spent fuel.


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5. Conclusions

The primary focus of this analysis was to identify the possible existence of margin

available for spent nuclear fuel rod dry cask storage. The failure mechanisms were

herein identified, and their leading causes were investigated. Creep rupture due to

temperature was determined to be the leading cause.

Through testing and modeling performed by PNL it was shown that spent fuel rod

temperatures decrease extremely quickly while the pressures required to force the creep

rupture increases. Those analyses also identified that although the NRC required that the

spent fuel be analyzed for conditions at 400C, testing has shown that fuel is more often

around 380C when it transitions from the spent fuel pool to dry storage. The analyses

also demonstrated that while the limit for the spent fuel is around 400C, the spent fueltemperature drops rapidly the first few years of dry cask storage. This is also true for

fuel stored in wet storage, and studies have shown that the reliability of the casks can be

greatly improved if the casks enter dry storage at a much lower temperature. This works

well with current industry practice, as much of the current spent fuel inventory has

already been held on site in spent fuel pools for the duration of the spent fuels life.

The analyses have also demonstrated, through widespread industry operating

experience, that the current generic standards have demonstrated impeccable safety. It is

through a combination of these data sources, that it can be concluded that storing spent

fuel in dry casks is a safe endeavor. Therefore, it can also be proposed that the storage

of spent fuel in dry casks be licensed for more than the current 20 year licensing period

in the United States. Other countries have already demonstrated a high degree of safety,

with operating histories longer than 20 years.

This analysis aimed to identify potential improvements for the Yucca mountain

proposal. The proposal was rejected for undisclosed reasons, but these reasons likely

dealt with the safety and reliability of the spent fuel storage. Therefore, as this document

identifies, it is simply proposed that the government acknowledges that the fuel storage

should be made safer by allowing the fuel to be reduced further in temperature prior to

storage at the Yucca Mountain site. This is currently already being done, and the

analyses additionally show that the fuel is typically already below the 400C, as required

by the NRC. Additionally the Yucca Mountain site could support storage of spent fuel


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far beyond 20 years of storage. This analysis has also shown, that given the rapid

temperature decrease of the spent fuel, the temperature vs. stress limits will be lower

than the predicted models limits.

Review of licensing requirements for spent fuel dry storage facilities identifies that

the licensing requirements need only reflect the intent of that specific site. That is, if a

site will not contain wet storage facilities, it isnt required to maintain safety measures

typical of those types of facilities. Therefore, the Yucca Mountain site could be housing

spent fuel far below the typical storage temperature limits and spent fuel age

requirements.

It is for these reasons, that the Yucca mountain proposal and more generally, spent

fuel dry storage, should be reexamined more thoroughly with respect to the temperature

and fuel age safety requirements, so that it may be fully utilized. Dry cask storage of

spent nuclear fuel has been shown to be a sufficiently safe practice, and given the current

industry environment will be even safer than previously believed. Additionally, the

industry has a great need for the storage of these materials as the current storage

practices will no longer be feasible as the industry ages and fuel storage limits are being

gradually met.

Furthermore, this analysis has demonstrated that sufficient margin is available for a

redesign of dry cask storage equipment. A number of changes to the design could be

made, and would need to be analyzed, but could greatly benefit the industry. Changes

such as these may lead to either the resurrection of the Yucca Mountain storage facility,

or allow for new storage facilities to be proposed. These storage facilities are severely

needed across the industry, and as this project has demonstrated, dry storage facilities are

a very reasonable and feasible alternative to the wet storage facilities.

This project can further recommend safer spent fuel dry storage practices. The

analyses have shown that spent fuel moved to dry storage early in life, will decrease in

temperature rapidly, and will maintain safe temperature vs. stress relationships. Industry

experience has also shown that the storage of this fuel can likely be extended well

beyond the current 20 year licensing period, as the fuel cools so significantly early in its

lifespan, and the fact that there has been such an impeccable operating history.


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Ideally, this project will shed some light on the current practices of spent fuel dry

storage practices. Through identifying some of the industries tests and applicable

nuclear power plants operating experiences, the level of safety associated with this

practice has been clarified further. Many alternatives to current practices are propsed in

Section 6, and the basis for these recommended changes are discussed and explained. It

would be necessary to perform extensive testing to justify any of these recommended

changes, but the benefit would be immense. The future of the nuclear industry requires

adequate spent fuel storage options, such as those which have been highlighted in this

project as for the need for energy ever increases.


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6. Future Industry Proposals

Given the results drawn from the above analysis it appears that the current industry

requirements are too stringent for the storage of spent nuclear fuel in dry casks. The

demonstrated temperature margin can be taken advantage of as follows:

1. To obtain NRC approval: To use the demonstrated temperature margin present

and the fact that the casks have been demonstrated to be hardier, to demonstrate

that an unwarranted safety margin is present in current dry cask storage

proposals. This increases likelihood of a proposal for a lower storage

temperature being approved by the current Presidential Administration.

2. Cost reduction: Reducing the design requirements would maintain the same

radiation protection, but would allow for less expensive gases for cooling, andwould also allow for relaxed physical requirements. Currently the casks are built

with inner steel and outer concrete walls. The inner steel walls are typically 24

thick while the inner steel walls are typically 4 thick. Given the analyses here, it

can be shown that the cask wall thicknesses can be reduced. This would save

money from a design, transportation and materials standpoint.

These proposed changes however would not significantly change the current

industry environment, and would not significantly reduce costs. The most

significant method to reduce storage costs would not be to reduce the design

requirements of the current dry casks types, but rather to propose a new storage

design.

3. Increased cask storage capacities: Currently the most significant restrictor for the

dry storage costs is the fact that only a certain number of spent fuel rods can be

stored in each dry cask. The current analysis has demonstrated that the fuel rods

can withstand higher temperatures than previously believed (these temperatures

fall extremely quickly as well). Therefore, it would still be effective to produce

slightly larger casks, which store more fuel rods in the same casks. This could be

done by very slightly increasing cask size, but significantly increasing the

amount of fuel that could be stored in the casks.

Clearly further analyses would need to be performed on the impact of this, but

based on the analyses this would be a reasonable endeavor to undertake. The


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added temperature and pressure impacts could be analyzed, and an appropriate

fuel amount per storage container could be determined. By simply increasing the

amount of fuel rods in a single cask the costs of spent fuel storage could be

significantly reduced, and although increasing the number of fuel rods in the dry

casks may raise the temperatures, the fuel temperature can be further reduced in

wet storage as is current practice, prior to this step.

4. Increasing the allowed storage time at the site: A final method for capitalizing on

the apparent safety margin demonstrated in the previous analysis would be to

extend the duration for spent fuel dry cask storage. The industry experiences and

analyses performed have demonstrated that spent fuel dry cask storage is safe

enough that the casks can be stored for an extended amount of time. This is

further supported by the fact that the fuels temperature decreases as time goes by,

which creates an even safer (less risk of failure) condition within the cask.

Therefore, it can be proposed that dry cask storage be licensed for longer than the

20 years currently recommended by the NRC. This view is also supported by

numerous research institutes, as demonstrated in Section 3.2 and has already

been demonstrated as being successful by other nations. By extending the

licensing period a significant amount of money can be saved by utilities and

storage facilities, which can be redirected to more appropriate measures.


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7. References

1. Bedard, P., Reid Celebrates Obamas Yucca Mountain Decision, U.S. News and

World Report, February 26, 2009.

2. Magnusson, P., Chen, J., Hoffelner, W., Thermal and Irradiation Creep Behavior of

a Titanium Aluminide in Advanced Nuclear Power Plants, Metallurgical and Metals

Transactions, Vol. 40, Number 12, ISSN-1073-5623, pages 2837-2842, December

2009.

3. NUREG-1804, Rev. 2, Yucca Mountain Review Plan, July 2003.

4. Fact Sheet on Storage of Spent Nuclear Fuel, NRC Website. 2 December 2009, .5. Ashar, H., Degrassi, G., Design and Analysis of Free-Standing Spent Fuel Racks in

Nuclear Power Plants, U.S. Nuclear Regulatory Commission and Brookhaven

National Laboratory.

6. Dry Cask Storage, Entergy Website. 6 December 2009, .

7. Yamamoto, T., Hode, S., et al., Developments in Spent Fuel Transport and Storage

Casks, Mitsubishi Heavy Industries, Ltd., December 2006.

8. 10 CFR 72.106, Controlled Area of an ISFSI or MRS.

9. 10 CFR 72.108, Spent Fuel Of High-Level Radioactive Waste Transportation.

10. 10 CFR 72.236, Specific Requirements for Spent Fuel Storage Cask Approval and

Fabrication.

11. 10 CFR 72.104, Criteria for Radioactive Materials in Effluents and Direct Radiation

from an ISFSI or MRS.

12. Johnson, A., Gilbert, E., PNL-4835, Technical Basis for Storage of Zircalloy-Clad

Spent Fuel in Inert Gases, Pacific Northwest Laboratory, September 1983.

13. NUREG-1567, Standard Review Plan for Spent Fuel Dry Storage Facilities, March

2000.

14. NS-G-1.4, Design of Fuel Handling and Storage Systems for Nuclear Power

Plants, International Atomic Energy Agency, August 2003.


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15. NUREG-1536, Standard Review Plan for Dry Cask Storage Systems, January

1997.

16. Santos, C., Kalinousky, D., Ryder, C., et al., M04-6, A Probabilistic Failure

Assessment of a Dry Cask Storage System, Transactions of the 17th International

Conference on Structural Mechanics in Reactor Technology (SMiRT 17), August

2003.

17. Glasstone, S., Sesonske, A., Nuclear Reactor Engineering, Van Nostrand Reinhold

Company, 1981.

18. Sandquist, G.M., et al., Exposures and Health Effects from Spent Fuel

Transportation, RAE-8339/12-1, Prepared for U.S. DOE, Office of Civilian

Radioactive Waste Management, Salt Lake City: Rogers and Associates Engineering

Corporation, November 29, 1985.

19. Halstead, R., Radiation Exposures from Spent Nuclear Fuel and

High-Level Nuclear Waste Transportation to a Geologic Repository or Interim

Storage Facility in Nevada, State of Nevada Nuclear Waste Project Office. 14

December 2009, < http://www.state.nv.us/nucwaste/trans/radexp.htm#notes>.

20. IAEA-TECDOC-1410, Delayed Hydride Cracking in Zirconium Alloys in Pressure

Tube Nuclear Reactors, International Atomic Energy Agency, October 2004.

21. Machiels, A., EPRI-1001207, Creep as the Limiting Mechanism for Spent Fuel Dry

Storage, Electric Power Research Institute, December 2000.

22. Schreiner, R., ANL-WIS-MD-000008, Rev. 2, Clad Degradation FEPs Screening

Arguments, Bechtel SAID, October 2004.

23. Linderoth, W., BTM 03-126, Rev. 0, Fuel Rod Behavior Under Spent Fuel Dry

Storage Conditions, Westinghouse Electric Sweden. June 2004.

24. Pescatore, C., Cowgill, M., Sullivan, T., BNL-52235, Zircaloy Cladding

Performance Under Spent Fuel Disposal Conditions, Brookhaven National

Labrator, October 1989.

25. Einziger, R., McKinnon, M., Machiels, A., PNL-10813, Extending Dry Storage Of

Spent LWR Fuel For Up To 100 Years, International Symposium on Storage of

Spent Fuel from Power Reactors. Vienna, Austria. September 1999.


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26. PNL-6364, Control of Degradation of Spent LWR Fuel During Dry Storage in an

Inert Atmosphere, Pacific Northwest Laboratory, October 1987.

27. IAEA-TECHDOC-1100, Survey of Wet and Dry Spent Fuel Storage, International

Atomic Energy Agency, July 1999.

28. Nechaev, A., Onufriev, V., Thomas, K., Long-Term Storage and Disposal of Spent

Fuel,International Atomic Energy Agency, Spring 1986.

29. PNL-6189, Recommended Temperature Limits for Dry Storage of Spent Light

Water Reactor Zircaloy-Clad Fuel Rods in Inert Gas, Pacific Northwest Laboratory,

May 1987.

30. Tarn, J., Madsen, N., Chin, B., PNL-SA-13879, Predictions of Dry Storage

Behavior of Zircaloy Clad Spent F

121709 masters project in mechanical engineering

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