This “Technical Evaluation Report on the Content of the U.S. Department of Energy’s Yucca Mountain License Application; Postclosure Volume: Repository Safety After Permanent Closure” (TER Postclosure Volume) presents information on the NRC staff’s review of DOE’s Safety Analysis Report (SAR), provided on June 3, 2008, as updated by DOE on February 19, 2009. The NRC staff also reviewed information DOE provided in response to NRC staff’s requests for additional information and other information that DOE provided related to the SAR.

The "Summary Report of SNF Isotopic Comparisons for the Disposal Criticality Analysis Methodology" contains a summary of the analyses that compare SNF measured isotopic concentrations (radiochemical assays) to calculated SNF isotop~c concentrations (SAS2H module ·orScale4.3). The results of these analyses are used to support the validation of the isotopic models for spent commercial light water reactor (LWR) fuel.

NRC Waste Confidence Rulemaking, Federal Register, 1984, 1990, 1999, and 2008

In the existing U.S. Environmental Protection Agency (EPA) and Nuclear Regulatory Commission (NRC) regulations governing the spent nuclear fuel and high-level radioactive waste site at Yucca Mountain, Nevada, the time period of compliance was set at 10,000 years. Recently, a Court ordered that EPA and NRC either revise the regulation on this topic to be "based upon and consistent with" recommendations made by a panel of the National Academy of Sciences, who recommended a time period of compliance out to as long as one million years, or seek congressional relief.

The requirements of ANSI/ANS 8.1 specify that calculational methods for away-from-reactor
(AFR) criticality safety analyses be validated against experimental measurements. If credit for the
negative reactivity of the depleted (or spent) fuel isotopics is desired, it is necessary to benchmark
computational methods against spent fuel critical configurations. This report summarizes a portion
of the ongoing effort to benchmark AFR criticality analysis methods using selected critical
configurations from commercial pressurized-water reactors (PWR).

STARBUCS is a new prototypic analysis sequence for performing automated criticality safety analyses of spent fuel systems employing burnup credit. A depletion analysis calculation for each of the burnup-dependent regions of a spent fuel assembly, or other system containing spent fuel, is performed using the ORIGEN-ARP sequence of SCALE. The spent fuel compositions are then used to generate resonance self-shielded cross sections for each region of the problem, which are applied in a three-dimensional criticality safety calculation using the KENO V.a code.

The requirements of ANSI/ANS 8.1 specify that calculational methods for away-from-reactor
criticality safety analyses be validated against experimental measurements. If credit for the negative
reactivity of the depleted (or spent) fuel isotopics is desired, it is necessary to benchmark
computational methods against spent fuel critical configurations. This report summarizes a portion
of the ongoing effort to benchmark away-from-reactor criticality analysis methods using critical
configurations from commercial pressurized-water reactors.

Thirty spent fuel samples obtained from boiling-water-reactor (BWR) fuel pins have been
modeled at Oak Ridge National Laboratory using the SAS2H sequence of the SCALE code system.
The SAS2H sequence uses transport methods combined with the depletion and decay capabilities
of the ORIGEN-S code to estimate the isotopic composition of fuel as a function of its burnup
history. Results of these calculations are compared with chemical assay measurements of spent fuel
inventories for each sample. Results show reasonable agreement between measured and predicted

This paper provides a brief description of the United States Nuclear Regulatory Commission (NRC) and its regulatory process for the current nuclear fuel cycle for light water power reactors (LWRs). It focuses on the regulatory framework for the licensing of facilities in the fuel cycle. The first part of the paper provides an overview of the NRC and its regulatory program including a description of its organization, function, authority, and responsibilities.

Isotopic characterization of spent fuel via depletion and decay calculations is necessary for
determination of source terms for subsequent system analyses involving heat transfer, radiation
shielding, isotopic migration, etc. Unlike fresh fuel assumptions typically employed in the criticality
safety analysis of spent fuel configurations, burnup credit applications also rely on depletion and
decay calculations to predict the isotopic composition of spent fuel. These isotopics are used in

Five commercial reactor criticals (CRCs) for the LaSalle Unit 1 boiling-water reactor
have been analyzed using KENO V.a, the Monte Carlo criticality code of the SCALE 4 code
system. The irradiated fuel assembly isotopics for the criticality analyses were provided by the
Waste Package Design team at the Yucca Mountain Project in the United States, who performed
the depletion calculations using the SAS2H sequence of SCALE 4. The reactor critical
measurements involved two beginning-of-cycle and three middle-of-cycle configurations. The

The requirements of ANSI/ANS-8.1 specify that calculational methods for away-from-reactor
criticality safety analyses be validated against experimental measurements. If credit is to be taken for
the reduced reactivity of burned or spent fuel relative to its original "fresh" composition, it is
necessary to benchmark computational methods used in determining such reactivity worth against
spent fuel reactivity measurements. This report summarizes a portion of the ongoing effort to

This report is part of a report series designed to document benchmark-quality radiochemical assay data
against which computer code predictions of isotopic composition for spent nuclear fuel can be validated
to establish the uncertainty and bias associated with the code predictions. The experimental data analyzed
in the present report were acquired from two international programs: (1) ARIANE and (2) REBUS, both
coordinated by Belgonucleaire. All measurements include extensive actinide and fission product data of

The requirements of ANSI/ANS 8.1 specify that calculational methods for away-from-reactor
criticality safety analyses be validated against experimental measurements. If credit is to be taken for
the reduced reactivity of burned or spent fuel relative to its original $fresh# composition, it is
necessary to benchmark computational methods used in determining such reactivity worth against
spent fuel reactivity measurements. This report summarizes a portion of the ongoing effort to

The validity of the computation of pressurized-water-reactor (PWR) spent fuel isotopic
composition by the SCALE system depletion analysis was assessed using data presented in the report.
Radiochemical measurements and SCALE/SAS2H computations of depleted fuel isotopics were
compared with 19 benchmark-problem samples from Calvert Cliffs Unit 1, H. B. Robinson Unit 2,
and Obrigheim PWRs. Even though not exhaustive in scope, the validation included comparison of
predicted and measured concentrations for 14 actinides and 37 fission and activation products.

This Interim Staff Guidance (ISG) provides guidance to the staff for determining if
storage systems to be licensed under 10 CFR Part 72 allow ready retrieval of spent fuel.
This guidance is not a regulation or a requirement.

This Interim Staff Guidance (ISG) provides guidance to the staff on classifying spent nuclear
fuel as either (1) damaged, (2) undamaged, or (3) intact, before interim storage or
transportation. This is not a regulation or requirement and can be modified or superseded by
an applicant with supportable technical arguments.

Revision 2

Compliance with 10 CFR 72.122(l) has been interpreted to mean that a licensee, during any
point in the storage cycle, must have a means of retrieving and repackaging individual fuel
assemblies even after an accident. The staff has reevaluated this interpretation.

The closure weld for the outer cover plate for austenitic stainless steel designs may be
inspected using either volumetric or multiple pass dye penetrant techniques subject to the
following conditions:
• Dye penetrant (PT) examination may only be used in lieu of volumetric
examination only on austenitic stainless steels. PT examination should be done
in accordance with ASME Section V, Article 6, “Liquid Penetrant Examination.”
• For either ultrasonic examination (UT) or PT examination, the minimum

Several changes have occurred since the issuance of NUREG-1536, “Standard Review Plan
(SRP) for Dry Cask Storage Systems,” that affect the staff’s approach to confinement
evaluation. The attachment to this ISG integrates the current staff approach into a revision of
ISG-5. The highlights of the changes include:
• Reflects October 1998 revisions to 10 CFR 72.104 and 10 CFR 72.106.
• Expands and clarifies acceptance criteria associated with confinement analysis and
acceptance of “leak tight” testing instead of detailed confinement analysis.

The Standard Review Plan, NUREG-1536, Chapter 5, Section V, 2 recommends that “the
applicant calculate the source term on the basis of the fuel that will actually provide the
bounding source term,” and states that the applicant should, “either specify the minimum initial
enrichment or establish the specific source terms as operating controls and limits for cask use.”
A specified source term is difficult for most cask users to determine and for inspectors to verify.

Staff raised two major issues concerning the adverse effects of fission gases to the gas-mixture
thermal conductivity in a spent fuel canister in a post accident environment. The two major
concerns were: (1) the reduction of the thermal conductivity of the canister gas by the mixing of
fission gases expelled from failed fuel pins and (2) the resultant temperature and pressure rise
within the canister. Since the fission gas is typically of a lower conductivity than the cover gas,

Title 10 of the Code of Federal Regulations (10 CFR) Part 71, Packaging and Transportation of
Radioactive Material, and 10 CFR Part 72, Licensing Requirements for the Independent
Storage of Spent Nuclear Fuel, High-Level Radioactive Waste, and Reactor-Related Greater
Than Class C Waste, require that spent nuclear fuel (SNF) remain subcritical in transportation
and storage, respectively. Unirradiated reactor fuel has a well-specified nuclide composition
that provides a straightforward and bounding approach to the criticality safety analysis of