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.

Presented at the NEI Used Fuel Management Conference, St. Petersburg, FL, May 7-9, 2013

NRC/NEI, January 24, 2014 Public Meeting Presentations

Presented at the NEI Used Fuel Management Conference, St. Petersburg, FL, May 7-9, 2013

From 1998 to 2004, a series of critical experiments referred to as the fission product (FP) experimental program was performed at the Commissariat à l'Energie Atomique Valduc research facility. The experiments were designed by Institut de Radioprotection et de Sûreté Nucléaire (IRSN) and funded by AREVA NC and IRSN within the French program supporting development of a technical basis for burnup credit validation.

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

A methodology for performing and applying nuclear criticality safety calculations, for PWR spent nuclear fuel (SNF) packages with actinide-only burnup credit, is described. The changes in the U-234, U-235, U-236, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, and Am-241 concentration with burnup are used in burnup credit criticality analyses. No credit for fission product neutron absorbers is taken. The methodology consists of five major steps. (1) Validate a computer code system to calculate isotopic concentrations of SNF created during burnup in the reactor core and subsequent decay.

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.

In the 1980s a series of the Haut Taux de Combustion (HTC) critical experiments with fuel pins in a water-moderated lattice was conducted at the Apparatus B experimental facility in Valduc (Commissariat à l'Energie Atomique, France) with the support of the Institut de Radioprotection et de Sûreté Nucléaire and AREVA NC. Four series of experiments were designed to assess profit associated with actinide-only burnup credit in the criticality safety evaluation for fuel handling, pool storage, and spent-fuel cask conditions.

In the 1980s, a series of critical experiments referred to as the Haut Taux de Combustion (HTC)
experiments was conducted by the Institut de Radioprotection et de Sûreté Nucléaire (IRSN) at the
experimental criticality facility in Valduc, France. The plutonium-to- uranium ratio and the isotopic
compositions of both the uranium and plutonium used in the simulated fuel rods were designed to be
similar to what would be found in a typical pressurized-water reactor fuel assembly that initially had an

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.

Presented at the NEI Used Fuel Management Conference, St. Petersburg, FL, May 7-9, 2013

Presented at the NEI Used Fuel Management Conference, St. Petersburg, FL, May 7-9, 2013

The purpose of this engineering calculation is to document the benchmark range, over a variety of parameters, for the validation of the criticality calculations supporting the Monitored Geologic Repository (MGR). This engineering calculation accomplishes this by characterizing the Laboratory Critical Experiments (LCE) and the Pressurized Water Reactor (PWR) Commercial Reactor Criticals (CRC), and summarizing the significant parameters. This engineering calculation supports the Disposal Criticality Analysis Methodology program.

The purpose of this design analysis is to determine the accuracy of the SAS2H module of SCALE 4.3 in predicting isotopic concentrations of spent fuel assemblies. The objective is to develop a methodology for modeling assemblies similar to those evaluated within this analysis and to establish the consistency of SAS2H predictions. The results of this analysis may then be applied to future depletion calculations using SAS2H in which no measurements are available. The analytical model employed for this analysis was the SAS2H module of the SCALE sequence.

These slides were presented by Waste Control Specialists LLC (WCS) to the NRC at the June 16, 2015 pre-application public meeting at the NRC offices in Rockville, Maryland.

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