ASTM C1562-10(2018)

This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the
Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.
Designation: C1562 − 10 (Reapproved 2018)
Standard Guide for
Evaluation of Materials Used in Extended Service of Interim
Spent Nuclear Fuel Dry Storage Systems
This standard is issued under the fixed designation C1562; the number immediately following the designation indicates the year of
original adoption or, in the case of revision, the year of last revision.Anumber in parentheses indicates the year of last reapproval.A
superscript epsilon (´) indicates an editorial change since the last revision or reapproval.
1. Scope 1.3 Thisguideaddressesthedeterminationoftheconditions
of the spent fuel and storage cask materials at the end of the
1.1 Part of the total inventory of commercial spent nuclear
initial20-yearlicenseperiodastheresultofnormaleventsand
fuel (SNF) is stored in dry cask storage systems (DCSS) under
conditions. However, the guide also addresses the analysis of
licenses granted by the U.S. Nuclear Regulatory Commission
potentialspentfuelandcaskmaterialsdegradationastheresult
(NRC). The purpose of this guide is to provide information to
of off-normal, and accident-level events and conditions that
assist in supporting the renewal of these licenses, safely and
may occur during any period.
without removal of the SNF from its licensed confinement, for
periods beyond those governed by the term of the original
1.4 This guide provides information on materials behavior
license.This guide provides information on materials behavior
to support continuing compliance with the safety criteria,
under conditions that may be important to safety evaluations
which are part of the regulatory basis, for licensed storage of
for the extended service of the renewal period. This guide is
SNFatanISFSI.Thesafetyfunctionsaddressedanddiscussed
written for DCSS containing light water reactor (LWR) fuel
in this standard guide include thermal performance, radiologi-
that is clad in zirconium alloy material and stored in accor-
cal protection, confinement, sub-criticality, and retrievability.
dance with the Code of Federal Regulations (CFR), at an
The regulatory basis includes 10 CFR Part 72 and supporting
independent spent-fuel storage installation (ISFSI). The com-
regulatoryguidesoftheU.S.NuclearRegulatoryCommission.
ponents of an ISFSI, addressed in this document, include the
Therequirementssetforthinthesedocumentsindicatethatthe
commercial SNF, canister, cask, and all parts of the storage
installationincludingtheISFSIpad.Thelanguageofthisguide following items were considered in the original licensing
is based, in part, on the requirements for a dry SNF storage
decisions: properties of materials, design considerations for
license that is granted, by the U.S. Nuclear Regulatory Com-
normal and off-normal service, operational and natural events,
mission (NRC), for up to 20 years. Although government
and the bases for the original calculations. These items may
regulations may differ for various nations, the guidance on
require reconsideration of the safety-related arguments that
materials properties and behavior given here is expected to
demonstratehowthesystemscontinuetosatisfytheregulatory
have broad applicability.
requirements. Further, to ensure continued safe operation, the
1.2 This guide addresses many of the factors affecting the
performance of materials must be justified in relation to the
time-dependent behavior of materials under ISFSI service [10
effects of time, temperature, radiation field, and environmental
CFR Part 72.42]. These factors are those regarded to be
conditions of normal and off-normal service. Arguments for
important to performance, in license extension, beyond the
long-term performance must account for materials alterations
currently licensed 20-year period. Examples of these factors
(especially degradations) that are expected during the service
aregiveninthisguideandtheyincludematerialsalterationsor
periods, which include the periods of the initial license and of
environmental conditions for components of an ISFSI system
the license renewal. This guide pertains only to structures,
that, over time, could have significance related to safety. For
systems, and components important to safety during extended
purposes of this guide, a license period of an additional 20 to
storage period and during retrieval functions, including trans-
80 years is assumed.
port and transfer operations. Materials information that per-
tains to safety functions, including retrieval functions, is
This guide is under the jurisdiction ofASTM Committee C26 on Nuclear Fuel
pertinent to current regulations and to license renewal process,
Cycle and is the direct responsibility of Subcommittee C26.13 on Spent Fuel and
High Level Waste.
andthisinformationisthefocusoftheguide.Thisguideisnot
Current edition approved Sept. 1, 2018. Published October 2018. Originally
intended to supplant the existing regulatory process.
approved in 2003. Last previous edition approved in 2010 as C1562–10. DOI:
10.1520/C1562-10R18.
1.5 This international standard was developed in accor-
In general fuels of higher burnup (>45 MWd/kgU) and MOX fuels are not
dance with internationally recognized principles on standard-
included in this guide. Guidance for these fuels are expected to be included in an
Annex to be written later. ization established in the Decision on Principles for the
Copyright © ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959. United States
C1562 − 10 (2018)
Development of International Standards, Guides and Recom- ACI 349-00Code Requirements for Nuclear Safety Related
mendations issued by the World Trade Organization Technical Concrete Structures
Barriers to Trade (TBT) Committee. ACI 359-01 Code for Concrete Reactor Vessels and
Containments, also designated asASME Boiler and Pres-
2. Referenced Documents
sure Vessel Code, Section III, Div 2, Code for Concrete
Reactor Vessels and Containments
2.1 ASTM Standards:
2.5 ANSI Documents:
C33/C33MSpecification for Concrete Aggregates
ANSI/ANS-6.4-1985Guidelines on the Nuclear Analysis
C227 Test Method for Potential Alkali Reactivity of
and Design of Concrete Radiation Shielding for Nuclear
Cement-Aggregate Combinations (Mortar-Bar Method)
Power Plants
C295/C295MGuide for Petrographic Examination of Ag-
ANSI/ANS-57.9Design Criteria for an Independent Spent
gregates for Concrete
Fuel Storage Installation (Dry Storage Type)
C859Terminology Relating to Nuclear Materials
ANSI/ANS-57.10DesignCriteriaforConsolidationofLWR
C1174PracticeforEvaluationoftheLong-TermBehaviorof
Spent Fuel
Materials Used in Engineered Barrier Systems (EBS) for
2.6 Other Documents:
Geological Disposal of High-Level Radioactive Waste
ASME-B&PV Sect III-Div 2 (2001)Code for Concrete
2.2 Government Documents:
Reactor Vessels and Containments
10 CFR Part 50Domestic Licensing of Production and
EPRI-1994Class I Structures License Renewal Industry
Utilization Facilities
Report; Revision 1, TR-103842, July 1994
10 CFR Part 60Disposal of High Level Radioactive Wastes
in Geologic Repositories
3. Terminology
10 CFR Part 63Disposal of High Level Radioactive Wastes
3.1 The terminology of Terminology C859 applies to this
in a Proposed Geologic Repository in Yucca Mountain
document except as given below.
10 CFR Part 71Packaging and Transport of Radioactive
Materials
3.2 Definitions of Terms Specific to This Standard:
10CFRPart72LicensingRequirementsfortheIndependent
3.2.1 accident-level events or conditions—theextremelevel
Storage of Spent Nuclear Fuel and High-Level Radioac-
of an event or condition for which there is a specified
tive Waste
resistance, limit of response, and requirement for a given level
2.3 NUREG Standards:
of continuing capability, which exceed “off-normal” events or
NUREG-1536Standard Review Plan for Dry Storage Cask
conditions. They include both design basis accidents and
Systems, January 1997
design-basis for natural phenomena events and conditions.
NUREG-1567Standard Review Plan for Spent Fuel Dry
NUREG-1536, NUREG-1567
Storage Facilities, Report, January 1998
NOTE 1—Specific accident conditions to be addressed have been
NUREG-1571Information Handbook on Independent Spent
evaluated for each dry cask storage system (DCSS) and are documented
Fuel Storage Installations
in a Safety Analysis Report for that system.
NUREG/CR-6407Classification of Transportation Packag-
3.2.2 alteration mode—a particular form of alteration, for
ing and Dry Spent Fuel Storage System Components
example, general corrosion, passivation. C1174
According to Importance to Safety, February, 1996, INEL
3.2.3 ASTM guide—a compendium of information or series
Report 95/0551
ofoptionsthatdoesnotrecommendaspecificcourseofaction.
ISG-1Interim Staff Guidance Number 1, U.S. NRC, Spent
3.2.4 canister—in a dry cask storage system (DCSS) for
Fuel Project Office
spent nuclear fuel, a metal cylinder that is sealed at both ends
2.4 American Concrete Institute Standards:
and is used to perform the function of confinement, while a
ACI 201.2R-97Guide to Durable Concrete
separate overpack performs the functions of shielding and
ACI 209R-97Prediction of Creep, Shrinkage and Tempera-
protection of the canister from the effects of impact loading.
ture Effects in Concrete Structures
ACI 301-99Building Code Requirements for Reinforced
3.2.5 cask—in a dry cask storage system (DCSS) for spent
Concrete
nuclear fuel, a stand-alone device that performs the functions
ACI 318-02Building Code Requirements for Reinforced
of confinement, radiological shielding, and physical protection
Concrete
of spent fuel during normal, off-normal, and accident
conditions. NUREG-1571
3.2.6 certificate of compliance—in a dry cask storage sys-
For referenced ASTM standards, visit the ASTM website, www.astm.org, or
contact ASTM Customer Service at service@astm.org. For Annual Book of ASTM tem (DCSS) for spent nuclear fuel, a certificate issued by the
Standards volume information, refer to the standard’s Document Summary page on
the ASTM website.
4 7
AvailablefromU.S.GovernmentPrintingOfficeSuperintendentofDocuments, Available fromAmerican National Standards Institute (ANSI), 25 W. 43rd St.,
732 N. Capitol St., NW, Mail Stop: SDE, Washington, DC 20401, http:// 4th Floor, New York, NY 10036, http://www.ansi.org.
www.access.gpo.gov. Available from American Society of Mechanical Engineers (ASME), ASME
Available from National Technical Information Service (NTIS), 5285 Port International Headquarters, Two Park Ave., New York, NY 10016-5990, http://
Royal Rd., Springfield, VA 22161, http://www.ntis.gov. www.asme.org.
6 9
AvailablefromAmericanConcreteInstitute(ACI),P.O.Box9094,Farmington Available from Electric Power Research Institute (EPRI), 3512 HillviewAve.,
Hills, MI 48333-9094, http://www.aci-int.org. Palo Alto, CA 94304–1395.
C1562 − 10 (2018)
U.S. Nuclear Regulatory Commission (NRC) to the designer/ thereisacorrespondingmaximumspecifiedresistance,limitof
vendorofaspecificcaskmodelthatmeetstherequirementsset response, or requirement for a given level of continuing
forth in 10 CFR Part 72.236. capability. NUREG-1536
3.2.7 confinement—in a dry cask storage system (DCSS) for
NOTE 3—Specific off-normal conditions to be addressed have been
evaluated for each licensed DCSS and are documented in a Safety
spent nuclear fuel, the ability to prevent the release of
Analysis Report for that system.
radioactive substances into the environment. NUREG-1571
3.2.18 radiation shielding—in a dry cask storage system
3.2.8 confinement systems—in a dry cask storage system
(DCSS) for spent nuclear fuel, barriers to radiation, which are
(DCSS) for spent nuclear fuel, the assembly of components of
designed to meet the requirements of 10 CFR Parts 72.104(a),
thepackagingintendedtoretaintheradioactivematerialduring
and 72.106(b), and 72.128(a.2).
storage. These may include the cladding, storage system shell,
3.2.19 retrievability—in a dry cask storage system (DCSS)
bottom and lid, penetration covers, the closure welds or seals
for spent nuclear fuel, the ability to remove spent nuclear fuel
and bolts and other components. NUREG-1536
from storage for further processing or disposal. 10 CFR Part
3.2.9 criticality—in a dry cask storage system (DCSS) for
72.122 (1)
spent nuclear fuel, the condition wherein a system or medium
3.2.20 safety analysis report (SAR)—in a dry cask storage
is capable of sustaining a nuclear chain reaction. C859
system (DCSS) for spent nuclear fuel, the document that is
3.2.10 degradation—any change in the properties of a
supplied by a DCSS vendor or site-specific independent spent
material that adversely affects the behavior of that material;
fuel storage installation (ISFSI) applicant to the U.S. Nuclear
adverse alteration. C1174
Regulatory Commission (NRC) for analysis and confirming
3.2.11 degraded cladding—in spent nuclear fuel, cladding
calculations (review and approval). NUREG-1571
material that by visual inspection appears to be structurally
3.2.21 safetyevaluationreport(SER)—inadrycaskstorage
deformed or damaged to an extent that special handling is
system (DCSS) for spent nuclear fuel, the document that the
expected to be required. See ISG-1 for damaged fuel.
U.S. Nuclear Regulatory Commission (NRC) publishes after
3.2.12 dry cask storage system (DCSS)—in nuclear waste review of a Safety Analysis Report (SAR). NUREG-1571
management, a set of components that performs the functions
3.2.22 service conditions—in a dry cask storage system
of confinement, radiological shielding, and physical protection
(DCSS) for spent nuclear fuel, the time of service,
of spent nuclear fuel during normal, off-normal, and accident
temperatures, environmental conditions, radiation, and
conditions. Examples would include canister-based systems
loading, etc. that a component experiences during storage.
withtheirmetalorconcreteoverpackorvault,oranintegrated
3.2.23 spent nuclear fuel (SNF), spent fuel—nuclear fuel
cask.
that has undergone at least one year of decay since being used
3.2.13 dry storage—in nuclear waste management,thestor-
as a source of energy in a power reactor, and has not been
age of spent nuclear fuel after removal of the water from the
separated into its constituent elements by reprocessing.
fuel,claddingandallcomponentsofadrycaskstoragesystem,
NUREG-1571
and after the atmosphere has been replaced with an inert
NOTE 4—In this guide, only commercial light water reactor SNF that is
atmosphere.
clad in zirconium alloy material and has been removed from service is
considered.
3.2.14 independent spent fuel storage installation (ISFSI)—
any complex designed and constructed for interim dry storage
3.2.24 sub-criticality margin—in a dry cask storage system
ofspentnuclearfuelandotherradioactivematerialsassociated
(DCSS) for spent nuclear fuel, the difference between one and
with spent fuel storage. It must meet the requirements in 10
the allowed calculated effective neutron multiplication factor
CFR Part 72. In this guide a Monitored Retrievable Storage
(k ), which is maintained at or below 0.95 in accordance with
eff
(MRS) site is also considered an ISFSI. NUREG-1571
NUREG-1536 and NUREG-1567.
3.2.15 monitoring—in a dry cask storage system (DCSS) for
NOTE 5—An adequate margin of sub-criticality is regarded to be 0.05.
spent nuclear fuel, testing and data collection to determine the
3.2.25 thermal performance—in a dry cask storage system
status of a DCSS and to verify the continued efficacy of the
(DCSS) for spent nuclear fuel, heat-removal capability having
system, on the basis of measurements of specified parameters
testability and reliability consistent with its importance to
including temperature, radiation, functionality and/or charac-
safety. 10 CFR Part 72.128(a)(4)
teristics of components of the system.
3.2.26 time limited aging analysis (TLAA)—a calculation or
3.2.16 normal events and conditions—the maximum level
analysis that addresses the effects of time and environmental
of an event or condition expected to routinely occur.
conditions on the performance of a system or component.
NOTE 2—Specific normal conditions to be addressed have been evalu-
4. Summary of Guide
ated for each licensed DCSS and are documented in a Safety Analysis
Report for that system.
4.1 Informationinthisguidedealswithmaterialsaspectsof
3.2.17 off-normal events or conditions—in a dry cask stor- spent nuclear fuel dry storage facilities that relate to license
age system (DCSS) for spent nuclear fuel, the maximum level extension beyond the original twenty-year license. Safety and
of an event that, although not occurring regularly, can be retrievability of the spent nuclear fuel are to be maintained
expected to occur with moderate frequency, and for which throughout the licensed period.
C1562 − 10 (2018)
4.2 Topics addressed in this guide all relate to materials active barrier to radionuclide release during transfer and
performance including regulations, design, environmental transport operations. It is possible that cladding could be
conditions, materials behavior under various conditions and altered during dry storage. The hydrogen effects, fracture
circumstances, monitoring, evaluation, etc. References are toughnessofthecladdingandthecreepbehaviorareimportant
provided to guide the user of this document to find additional
parameters to be evaluated and controlled during the dry
information and analyses, if needed. The structure of the storage phase of the life cycle. These degradation mechanisms
document is presented here to show the contents and annexes
are discussed in Annex A2 and Annex A4.
and appendixes of this guide.
6. Performance Requirements Related to the Design of a
Contents Section
Scope 1
DCSS
Referenced Documents 2
Terminology 3
6.1 Materialsforextendedservicemustmeetthedesignand
Summary of Guide 4
performance requirements given in 10 CFR 72.The DCSS has
Significance and Use 5
Performance Requirements in Relation to DCSS 6
been designed to store spent fuel safely for a minimum of 20
Design
years and to permit maintenance as required in the original
The Materials Evaluation Process for Dry Storage 7
licensed term. Structures, systems and components important
License Renewal
Factors That Affect Materials Performance in Annex A1 to safety have been designed, fabricated, erected and tested to
an ISFSI
meet standards commensurate with their function and their
Potential Degradation Mechanisms and Behavior of Annex A2
importance to the safety of the overall system. The service
Spent Nuclear Fuel Cladding Under Normal
Conditions
conditionsfortherenewalperiodmaybelessseverethanthose
Potential Degradation Mechanisms and Behavior of Annex A3
of the initial licensing period. If the cask contains its original
DCSS Materials Under Normal Conditions
SNF, then the demands on materials properties for an addi-
Consequences and Potential Degradation Mecha- Annex A4
nisms Under Off-Normal/Accident Conditions
tional 20 to 80 years of storage may be reduced due to lower
Durability and Properties of Concrete Structures Annex A5
temperatures and radiation levels. The general assumption put
and Components
Background on Licenses, Materials, and Criticality Appendix X1 forth here regarding decreases in thermal and radiation condi-
Polymer Annex for C1562 Annex A6
tions are based on the expectation that reloading of SNF does
not occur. It is assumed that at the time of license renewal, the
5. Significance and Use
reloading of casks (with SNF different from that originally
5.1 Information is provided in this document and other
stored in a cask) is very unlikely. If new (replacement) SNF is
referenced documents to assist the licensee and the licensor in
putinthecask,thentherequirementsonthematerialproperties
analyzing the materials aspects of performance of SNF and
andtheabilitytomeetthemwouldhavetobedeterminedusing
DCSS components during extended storage. The effects of the
the conditions established by the properties of the new SNF.
service conditions of the first licensing period are reviewed in
6.2 Structures, Systems and Components (SSC):
the license renewal process. These service conditions are
highlighted and discussed in Annex A1 as factors that affect 6.2.1 ThefunctionsimportanttosafetyofDCSSStructures,
materials performance in an ISFSI. Emphasis is on the effects
systems and components (SSC) are [NUREG-1536] to main-
of time, temperature, radiation, and the environment on the tain:
condition of the SNF and the performance of components of
6.2.1.1 Thermal performance,
ISFSI storage systems.
6.2.1.2 Radiological protection,
5.2 The storage of SNF that is irradiated under the regula-
6.2.1.3 Confinement,
tions of 10 CFR Part 50 is governed by regulations in 10 CFR
6.2.1.4 Sub-criticality, and
Part 72. Regulatory requirements for the subsequent geologic
6.2.1.5 Retrievability.
disposal of this SNF are presently given in 10 CFR Part 60,
6.2.2 Systems,structuresandcomponentsthatareimportant
with specific requirements for the use of Yucca Mountain as a
to safety must be designed to accommodate the load combina-
repository being given in the regulatory requirements of 10
tionsapplicabletonormal,off-normalandaccidenteventswith
CFR Part 63. Between the life-cycle phases of storage and
anadequatemarginofsafetyper10CFRPart72-,122b,122c,
disposal,SNFmaybetransportedundertherequirementsof10
and 24c, 10 CFR Part 100, and 10 CFR Part 72.102(l). The
CFR Part 71.Therefore, in storage, it is important to acknowl-
DCSS must reasonably maintain confinement of radioactive
edge the transport and disposal phases of the life cycle. In
material under normal, off-normal and credible accident con-
doing this, the materials properties that are important to these
ditions [10 CFR Part 72.236(l)]. The cask must be designed
subsequent phases are to be considered in order to promote
and fabricated so that the spent fuel is maintained in a
successful completion of these subsequent phases in the life
sub-critical condition under credible conditions [10 CFR Part
cycle of SNF. Retrievability of SNF (or high-level radioactive
72 Part 72.236 § C; 10 CFR Part 72.124 (a)].
waste) is set as a requirement in 10 CFR Part 72.122(g)(5) and
10 CFR Part 72.122(l). Care should be taken in operations 6.2.3 For a license renewal, a DCSS should be analyzed to
conducted prior to disposal, for example, storage, transfer, and demonstrate that the SSC will continue to perform so as to
transport, to ensure that the SNF is not abused and that SNF ensure that SNF is maintained under conditions that meet
assemblies will be retrievable, the protective value of the safety requirements under design basis conditions, even for an
cladding is not degraded and remains capable of serving as an extended storage period (up to 80 additional years).
C1562 − 10 (2018)
6.2.4 The requirements of 10 CFR 72.122 (h)(1) seek to licensee must be able to determine when corrective action
ensure safe fuel storage and handling and to minimize post- needs to be taken to maintain safe storage conditions. Instru-
operational safety problems with respect to the removal of the mentation and control systems deemed to be important to
fuel from storage. In accordance with this regulation, the spent safety shall also remain operational during the license renewal
fuel cladding must be protected during storage against degra- period. Radiation exposure and dose rates to workers and the
dation that leads to gross ruptures, or the fuel must be public must not exceed acceptable levels and remain as low as
otherwise confined such that degradation of the fuel during reasonably achievable (ALARA).
storage will not pose operational problems with respect to its
6.5 Sub-Criticality:
removal from storage. Additionally, 10 CFR 72.122(l) and
6.5.1 Subcriticality must be maintained [10 CFR Part
72.236(m)requirethatthestoragesystembedesignedtoallow
72.124]. The neutron multiplication factor, k , must be main-
eff
ready retrieval of the spent fuel from the storage system for
tained at or below 0.95 so as to obtain an adequate sub-
further processing or disposal.
criticality margin. The DCSS must be designed to ensure that
6.3 Thermal Behavior: this limit on the computed k is not exceeded, under all
eff
6.3.1 The spent fuel cladding must be protected against credible conditions. Spent fuel handling, packaging, transfer,
and storage systems must be designed to be maintained
degradation by thermally activated processes. This is done by
subcritical and to ensure that, before a nuclear criticality
maintainingthetemperaturebelowallowablelimits.Spentfuel
accident is possible, at least two unlikely, independent, and
storage or handling systems must be designed with a heat-
concurrent or sequential changes have occurred in the condi-
removal capability having testability and reliability consistent
tions essential to nuclear criticality safety. In an extended
with its importance to safety [10 CFR Part 72.128(a)(4)]. The
license period, special attention should be given to all material
DCSS must be designed to provide adequate heat removal
and components which may undergo thermal or corrosive
capacity without active cooling systems [10 CFR Part
alteration or any actions that would result in geometric
72.236(f)].The conditions in the second storage period will be
rearrangement of either the boron (or other poison/neutron
lessseverethanintheoriginallicensetermsincethedecayheat
absorber) or the SNF.
(as well as the radiation source term) decreases with time.
6.5.2 Boron is the element usually added inside a DCSS to
Hence, the decreasing decay heat requires less heat removal
absorb thermal neutrons and to maintain neutron flux at a
capacity during the extended licensing period. The safety
function related to thermal performance is a requirement to moderately low level. Other absorbers (Hf, Gd or Cd) may be
considered for absorber applications. The level of boron is
protect the fuel, that is, to ensure against the type of cladding
damagementionedin6.2.3and6.2.4.Attheinitiallicensingof customarily specified as an areal density (which is the thick-
ness times the volume density) for solids, such as metal alloys
a DCSS, the temperature of the fuel is limited and the cask
design is important to the thermal performance requirement of andpolymersusedformixedneutronabsorbers.Thegeometry
or physical configuration of the fixed neutron absorbers in the
the DCSS. Due to heat decay and the significant decrease in
temperatures of the fuel and cask over time, this safety system is important, and the matrix materials must not fail,
corrodeordegrade,soastoensurethattheabsorberremainsin
requirement will be met for extended licensing periods pro-
place. If redistribution of SNF rods occurs within the canister,
vided that the thermal properties of the cask have not been
orifthereareanysignificantchangesorredistributionofeither
significantly degraded and the geometry of its contents have
the absorbing material plates or the moderators of the SNF, it
not been significantly altered.
must be shown that the k will remain at or below 0.95.
6.3.2 Examples of components used to meet the thermal
eff
6.5.3 Neutron absorbing materials must continue to be
performance criteria are (1) cooling fins, which, for metal
effective. The license renewal application should evaluate the
casks, are usually fabricated from carbon steel (SA283 or SA
durability of the neutron absorbing material in its radiation,
285 GradeA), copper, or stainless steel (SA240Type 304), so
thermal,stress,andchemicalenvironmentinthecask.Itshould
astoincreaseheattransfer,and (2)penetrationsintheconcrete
demonstrate that the material remains in place at the end of 20
shielding that allow air to cool the canister.
years,andwillremaininplaceforthelicenseextensionperiod.
6.4 Shielding/Radiation Protection and Confinement—
Consumption of neutron-absorbing materials during dry stor-
Radiological protection and confinement features that are
age period is generally not a matter of concern because the
sufficienttomeetallnecessaryrequirementsof10CFRPart72
neutron fluxes are low, and are almost entirely fast. Boron
should continue to be provided. The confinement canister of a
consumed in storage usually represents only a tiny fraction of
DCSS provides a redundant seal. This feature is one that aids
the available boron in the system.
in ensuring that the confinement systems perform their safety-
6.6 Retrievability—Storage systems must be designed to
related functions in a reliable manner that is predictable over
allow ready retrieval of spent fuel for further processing or
time. In some sub-systems the performance must be under
disposal [10 CFR Part 72.122(l)]. System conditions are set so
intermittentorcontinuousmonitoringusingappropriateinstru-
that materials alteration does not compromise retrievability.
mentation and control systems. These sub-systems are ex-
pected to experience material property changes as they age
7. The Materials Evaluation Process for Dry Storage
under the combined influences of radiation and temperature
License Renewal
(and in some instances chemical environment) associated with
dry cask storage. Typical examples are polymer-based 7.1 Materialsrequirementsthatareimportanttosafetymust
materials,elastomers,andorganicbasedmaterials.Inshort,the be considered for license renewal of an ISFSI. The following
C1562 − 10 (2018)
types of service are to be considered: normal events and special consideration appropriate to those events Clearly, any
conditions, off-normal events and conditions, and accident- actiontakenduetotheseeventsistakenatthetimeoftheevent
level events and conditions. Fig. 1 illustrates an analysis logic
and in Annex A1 the principal factors that affect materials
that might be considered (in accounting for alterations of
performance in ISFSI service are briefly described under the
materials) during a license renewal. It begins by asking
headings of Temperature, Radiation and Chemical Environ-
whether conditions have been other than normal, and if they
ment. The effects of these overall environmental conditions,
have not, the user establishes the new initial conditions, which
over time, on the properties of the materials may be important
result from normal service conditions. When either off-normal
to the performance of the materials. For a license renewal, the
or accident conditions had been experienced for a given cask
materials alterations and operational events during the first
system, the user is referred to appropriate Annex materials in
20-year storage period are considered along with the original
this guide to cover the selected conditions that may require
FIG. 1 Analysis Logic and Identification of Materials Conditions to be Considered in ISFSI License Renewal
C1562 − 10 (2018)
design bases, and future materials requirements for the service cask and all interior components of the storage container.
conditions of the renewal period. Neutronabsorbersmustcontinuetobeadequatelyeffectiveand
structural components (for example, baskets, supports, weld
7.2 Evaluation of Materials Capabilities in Relation to
closures, lifting lugs and all other components) of the DCSS
Service Requirements:
musthavesufficientstrengthtomeettherequiredperformance.
7.2.1 Assess the service conditions: normal, off-normal and
Seals must be maintained in accordance with requirements of
accident that occurred during the initial storage period.
the safety evaluation and safety analysis reports. If during the
7.2.2 Determine the profile or service history (time/
first licensing period for the ISFSI only normal events or
temperature, radiation, chemical environment) of the compo-
conditionshaveoccurred,thenitcanbeassumedthattherehas
nents to be analyzed.
been no air or water ingress into the storage casks, thermal
7.2.3 Establish the relevant properties of SNF and DCSS
conditions have been within specified limits and no significant
materials at the start of the license renewal period based on
thermal or mechanical damage to the spent fuel has occurred.
materials alterations that may have occurred during the initial
Thus, only those material degradation mechanisms discussed
storage period.
for normal conditions need be considered. These include the
7.2.4 Assess the capability of the materials to perform their
spent fuel in AnnexA2, the DCSS materials in AnnexA3, and
functional and safety requirements during the renewal period.
the concrete in Annex A5. All require analyses needed to
7.2.5 Methodologies for life prediction, under the scope of
establish the initial conditions for a license renewal.
this guide, are concerned with the alteration of the materials
7.3.2 Off-Normal Events and Conditions—Any off-normal
usedinthesub-systems,structuresandcomponentsofaDCSS.
events that occur during the original license term must be
Guidance is provided on evaluating the most significant
evaluated for their impact on materials behavior and capabili-
material alterations that have been observed or predicted to
ties during the extended term. Under off-normal events and
occur under dry storage conditions during the initial and
conditionsduringtheinitiallicenseperiod,theISFSImayhave
renewal license periods.To insure system performance in each
experienced no permanent deformation or design related faults
of these periods, an acceptable methodology for life prediction
associated with a degradation of capability to perform its full
should (1) identify alteration mechanisms, (2) quantify the
function over the full license period, although operations may
alterations, (3) evaluate the effects (on materials properties) of
be suspended or curtailed. If during the initial 20-year license
thealterations, (4)determineifthealterationscompromiseany
period, an off-normal event or condition has occurred, and that
safety functions of the system, and (5) determine the conse-
event has potentially allowed air or water ingress into the
quences of compromising the performance of the component,
storage cask, then those material degradation mechanisms,
the sub-system or the system (safety, operational, economic).
discussed in Annex A4, must be addressed in addition to the
Theuseofanacceptablemethodologywillhelptoestablishthe
mechanismsdiscussedinAnnexA2,AnnexA3,andAnnexA5
requirements for materials data and testing, monitoring and
toestablishtheinitialmaterialconditionforthelicenserenewal
surveillance, preventive maintenance and operations manage-
analysis.
ment.Inaddition,itisnotedthatPracticeC1174isexpectedto
7.3.3 AccidentLevelEventsandConditions—Ifanaccident-
beausefulreferenceforevaluationsofmaterialsissuesrelated
leveleventorconditionhasoccurredduringtheinitial(20-year
to license renewals for spent fuel dry storage. This ASTM
licensed)periodoftheISFSI,thereisapossibilityofalteration
practice includes the prediction of long-term behavior, as well
or damage to the spent fuel due to air/water ingress and/or
as methods and criteria for accelerated testing and the use of
mechanical trauma to the SNF or components of, the storage
modelsandmechanisticunderstandingsofalterationprocesses.
cask. The material degradation mechanisms discussed in An-
7.3 Establishing Initial Conditions for License Renewal—
nex A4, therefore, should be addressed in addition to the
Almost all components of an ISFSI that are subject to license
mechanismsdiscussedinAnnexA2,AnnexA3,andAnnexA5
renewal will have undergone only normal service conditions
in establishing the initial material condition for the license
during the initial license period. If off-normal or accident
renewal analysis.
conditions occurred in a manner that had adverse effects on
7.4 Consideration for Future DCSS Usage—During the
some components of an ISFSI during the initial license period
license renewal process, the applicant should assess the radia-
the components would have been required to be restored to
tion and thermal load for which license renewal is sought. If
their original design and licensing bases. However, the effects,
the SNF is anticipated to remain throughout the renewal
if any, of the off-normal or accident conditions on the ISFSI
componentsaretobeincludedinformingtheinitialconditions
TABLE 1 Guide to Use of Annex A1 through Annex A5 for
for materials evaluations for ISFSI license renewal and this is
Material Evaluations
shown in Fig. 1.Aging mechanisms evaluated for the original
Normal Off Normal/Accident
licensed term must again be evaluated for the license renewal
Factors Affecting Performance Annex A1 Annex A1
term. Similarly, any evidence gained through intermittent or
A
Fuel (UO ) Annex A2 Annex A4
routine monitoring of ISFSI components during the initial A
Cladding Annex A2 Annex A4
A
Cask Components Annex A3 Annex A4
storage term that suggest accelerated or unanticipated aging
Pad/Concrete Annex A5 Annex A5
must be evaluated for the license renewal term.
A
Only corrosion is discussed in Annex A4. There is no discussion of mechanical
7.3.1 Normal Events and Conditions—Normal conditions
disruption as these are unique to a given event and reports are developed to
includeadryandinertprotectiveenvironmentfor (1)theSNF,
describe their relevance to safety.
and inner and outer surfaces of the cladding, and (2) inside the
C1562 − 10 (2018)
period, then the service conditions as a result of the SNF will 7.5.1 Afterdecadesofstorage,SNFinextendeddrystorage
belessseverethanthoseoftheinitiallicenseperiod.However, is expected to undergo little, if any, further alterations. For
if credit is taken in the license renewal application for the less
continued safe operation and to protect the SNF, the cask,
severe conditions, then the SNF permitted to be stored in the neutron absorbers, shielding materials baskets, supports,
DCSS may be limited by those less severe conditions. Consid-
closures, lifting lugs as well as other (including minor)
eration should be given to the unlikely event that the applicant
components of the DCSS, must retain sufficient strength and
may need to reload the DCSS during the renewal period with
other physical and mechanical properties to meet the required
SNF having different thermal and radiological properties than
performance criteria, specifications, etc.
the SNF stored at the time of the initial license period. To
7.5.2 Alteration modes that could lead to degradation or
preclude SNF loading penalties in a DCSS, the applicant may
failure of cladding in extended dry storage under normal
considerstoringSNFhavingdesignbasispropertiesduringthe
storage conditions are discussed in Annex A2. Alteration
renewal period. The original design basis may differ from that
modes that could lead to degradation or failure in the other
requiredintherenewalperiod.Therefore,thedesignbasisused
DCSS components under normal conditions are discussed in
in a renewal application should correspond with the type of
Annex A3. Other degradation mechanisms that could become
SNF to be stored during the renewal period.
important inoff-normal or accident conditions arediscussedin
7.5 Degradation of SNF and DCSS Components During Annex A4. Mechanisms by which concrete can be altered are
Extended Storage: presented in Annex A5.
ANNEXES
(Mandatory Information)
A1. FACTORS THAT AFFECT MATERIALS PERFORMANCE IN AN ISFSI
A1.1 Introduction A1.2.3 One methodology for assessing the effect of tem-
perature on material performance was given by Peehs et al.
A1.1.1 Factors that affect the behavior of SNF (and other
(5-7) who suggests four phases (modes) that define the
components) in ISFSI service include (a) temperature, (b)
temperature range over a given period for evaluation of
radiation, and (c) the environment. The values of these factors
expected degradation mechanisms. Rates of temperature
change as a function of time.
change are principally a function of the age of the SNF.
Duration and temperature of these phases are functions of the
A1.2 Temperature
specific fuel and cask conditions. Peehs et al. (5-7) dealt with
A1.2.1 Temperature is an important factor for material
commercial light water reactor fuel with Zircaloy™ cladding.
performance since many degradation mechanisms are ther-
A1.2.3.1 Phase I—In the Peehs’ example, temperatures
mally activated. Over time in the DCSS, the temperature will
above 300°C are characterized by a rapid decrease in tempera-
decreaseduetodecreasingdecayheat.Temperaturesdiscussed
ture. Phase I is a short term stage, typical of the first two years
in this section are fuel-cladding temperatures. While the heat
in dry storage for SNF out of the reactor for less than seven
decay characteristics of the fuel govern the cladding
years. The duration of this stage in dry storage is, of course, a
temperature, in a DCSS various factors affect the temperature
function of the initial time in wet storage.
over the time that fuel remains in dry storage.
A1.2.3.2 Phase II—Temperatures between 175 and 300°C
A1.2.2 The temperature of the various components of a
are characterized by a medium rate of decrease in temperature
particularDCSSdependsontheburnup,initialenrichment,and
occurring later in interim storage (usually from two to five
decaytimeofthespentfuelandthedesign(thatis,orientation,
years in dry storage).
heat removal capability) of the DCSS. The temperature profile
A1.2.3.3 Phase III—Temperatures between 120 and 175°C
of the DCSS varies both radially and axially, with the maxi-
are characterized by a low rate of decrease in temperature.
mum temperature occurring over the center 50% of the cask
A1.2.3.4 Phase IV—Temperatures below 120°C, character-
(1, 2) and falling away sharply at the outer edges. Tempera-
ized by a negligible decrease in temperature.
turedropovertimehasbeencalculatedusingcaskheattransfer
A1.2.3.5 PhasesIIIandIVarecharacteristicofthetempera-
codes and decay heats from ORIGEN (2-4). In general a
temperature drop from 380°C to 100°C was calculated (for a tures expected for extended dry storage, in this example.
typical 5 year, 30 GWD/MTU SNF) for the first 10 years, with
A1.2.4 Thermal conditions external to the cask can be
thetemperatureremainingatabout100°Cforthenext90years
important to the alteration of properties to the concrete
(2).
components. When concrete is used as shielding the design
temperature range is given in the Safety Analysis Reports for
the DCSS system. A general discussion of the effect of
The boldface numbers in parentheses refer to the list of references at the end
of this standard. temperature on concrete is found in Annex A5.
C1562 − 10 (2018)
A1.3 Radiation A1.3.2 While these levels of gamma radiation (10 rad) are
not significant for materials used inside a cask system, their
A1.3.1 After 20 years of dry storage, the fast neutron
absorption by the shielding materials is important to the
fluence at the interior of the DCSS is typically on the order of
14 2
radiation protection afforded to people.
10 n/cm and the cumulative gamma dose is on the order of
10 rad. The radiation shielding within a DCSS absorbs and
A1.3.3 For discussion of materials used in seals, see
attenuates neutrons and decreases the exposure levels and the
A3.3.3.3.
potential damage to the materials of the exterior components
but, in general, at this fluence level the effects on materials of
A1.4 Chemical Environment
interest are small. These levels of neutron fluence could
A1.4.1 The potential chemical environments to be consid-
potentially have some effects on mechanical properties of
ered are: backfill gases which may be air, nitrogen, helium, or
steels. The ferritic materials would require at least several
argon, residual water remaining in the cask after drying, zinc
orders of magnitude greater neutron fluence to have any
vapor if internal components are galvanized, and (potentially)
significant effect on mechanical properties (8) and the effects
fissionproducts.Theeffectsofradiolysisonthecompositionof
would be limited to those on impact properties, that is, on
the internal atmosphere should be assessed in a license
either on the upper-shelf energy absorption or on the transition
temperature behaviors. On the other hand, for any austenitic renewal,wheneverconcernexistsoverthepresenceofnitrogen
materials, the effect are nil for the fluence levels of interest. or water.
A2. POTENTIAL DEGRADATION MECHANISMS AND BEHAVIOR OF SPENT NUCLEAR F
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