The Reactor and Fuel Cycle Technology Subcommittee was formed to respond to the charge—set forth in the charter of the Blue Ribbon Commission—to evaluate existing fuel cycle technologies and R&D programs in terms of multiple criteria.

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.

This analysis provides information necessary for total system performance assessment (TSPA) for the license application (LA) to include the excess U.S. Department of Energy (DOE) plutonium in the form of mixed oxide (MOX) spent nuclear fuel and lanthanide borosilicate (LaBS) glass. This information includes the additional radionuclide inventory due to MOX spent nuclear fuel and LaBS glass and the analysis that shows that the TSPA models for commercial spent nuclear fuel (CSNF) and high-level waste (HLW) degradation are appropriate for MOX spent nuclear fuel and LaBS glass, respectively.

As part of the plutonium waste form development and down-select process, repository analyses have been conducted to evaluate the long-term performance of these forms for repository acceptance. Intact and degraded mode criticality analysis of the mixed oxide (MOX) spent fuel is presented in Volume I, while Volume II presents the evaluations of the waste form containing plutonium immobilized in a ceramic matrix.

The purpose of this analysis is to provide an initial radionuclide inventory (in grams per waste package) and associated uncertainty distributions for use in the Total System Performance Assessment for the License Application (TSPA-LA) in support of the license application for the repository at Yucca Mountain, Nevada. This document is intended for use in postclosure analysis only.

The means to prevent and control criticality must be addressed as part of the Preclosure Safety Analysis (PCSA) required for compliance with 10 CFR Part 63 [DIRS 180319], where the preclosure period covers the time prior to permanent closure activities. This technical report presents the nuclear criticality safety evaluation that documents the achievement of this objective.

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.

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.

At the request of the U.S. Congress, the National Academies assessed the safety and
security of spent nuclear fuel stored in pools and dry casks at commercial nuclear power
plants in the United States. The public report can be viewed on the National Academies
Press website at http://books.nap.edu/catalog/11263.html.

The purpose of this report is to document calculations of the number of waste packages that could be damaged in a potential future igneous event intersecting a repository at YuccaMountain. The analyses include disruption from an igneous intrusion and from an igneous eruption. The analyses also support the evaluation of the potential consequences from a future event as part of the total system performance assessment (TSPA) for the license application for the Yucca Mountain Project.

The Blue Ribbon Commission on America’s Nuclear Future (BRC) asked us to study whether
occupational safety and health conditions in today's U.S. nuclear industry are reasonably safe,
and if those conditions have improved since the Three Mile Island event in 1979. The BRC also
asked us to look to the future, to try to anticipate worker safety and health risks that should be
addressed by the industry, its government regulators and private watchdogs.
Over the eight weeks allotted, we performed a limited review of the literature and spoke with

This report was developed in accordance with the requirements in Technical Work Plan for Postclosure Waste Form Modeling (BSC 2005 [DIRS 173246]). The purpose of the in-package chemistry model is to predict the bulk chemistry inside of a breached waste package and to provide simplified expressions of that chemistry as a function of time after breach to Total Systems Performance Assessment for the License Application (TSPA-LA).

The objective of Calculation of Isotopic Bias and Uncertainty for BWR SNF is to quantify the computational bias and uncertainty in the multiplication factor (keff) to be used for Boiling Water Reactor (BWR) spent nuclear fuel (SNF) burn-up credit. The scope of this bias and uncertainty determination covers 38 different radiochemical assay (RCA) spent fuel samples from 14 different fuel assemblies that were irradiated in four different BWRs. The irradiated fuel samples evaluated span an enrichment range of 2.53 weight percent U-235 through 3.95 weight percent U-235.

The Reactor and Fuel Cycle Technology Subcommittee was formed to respond to the charge—set forth in the charter of the BRC—to evaluate existing fuel cycle technologies and R&D programs in terms of multiple criteria.

The objective of the Limerick Unit 1 Radiochemical Assay Comparisons to SAS2H Calculations is to determine the accuracy of the SAS2H control module of the baselined modular code system SCALE, Version 4.4A (STN: 10129-4.4A-00), in predicting the isotopic concentrations of spent fuel, and to quantify the overall effect that the differences between the calculated and measured isotopic concentrations have on the system reactivity. The scope of this calculation covers eight different spent fuel samples from a fuel assembly that was irradiated in the Limerick Unit 1 boiling water reactor (BWR).

The purpose of this analysis report is to develop the summary cladding degradation abstraction that will be used in the Total System Performance Assessment for the License Application (TSPA-LA). Most civilian commercial nuclear fuel is encased in Zircaloy cladding. The analysis addressed in this report is intended to describe the postulated condition of commercial Zircaloy-clad fuel as a function of postclosure time after it is placed in the repository.

This design calculation revises and updates the previous criticality evaluation for the canister handling, transfer and staging operations to be performed in the Canister Handling Facility (CHF) documented in BSC (Bechtel SAIC Company) 2004 (DIRS 167614).

The purpose of this report is to evaluate the potential for penetration of the Alloy 22 (UNS N06022) waste package outer barrier by localized corrosion due to the deliquescence of soluble constituents in dust present on waste package surfaces. The results support a recommendation to exclude deliquescence-induced localized corrosion (pitting or crevice corrosion) of the outer barrier from the total system performance assessment for the license application (TSPA-LA).