Guidance concerning regulatory requirements for criticality analysis of new and spent fuel storage at light-water reactor power plants used by the Reactor Systems Branch.

The purpose of this calculation is to document the Three Mile Island Unit 1 pressurized water reactor (PWR) fuel depletion calculations performed as part of the commercial reactor critical (CRC) evaluation program. The CRC evaluations support the development and validation of the neutronics models used for criticality analyses involving commercial spent nuclear fuel in a geologic repository.

In the development of a methodology to account for exposure effects on the reactivity of spent Boiling Water Reactor (BWR) fuel in the proposed Monitored Geologic Repository (MGR) at Yucca Mountain, the accuracy of the methods used to predict the inventories of fissile and fissionable nuclides as well as neutron poisons present in the spent fuel must be established. One aspect ofthis confirmatory effort is accomplished by performing benchmark problems for known in-reactor critical configurations - Commercial Reactor Criticals (CRCs).

The purpose of this calculation is to estimate the probability of misloading a commercial spent
nuclear fuel waste package with a fuel assembly(s) that has a reactivity (i.e., enrichment and/or
burnup) outside the waste package design. The waste package designs are based on the expected
commercial spent nuclear fuel assemblies and previous analyses (Macheret, P. 2001, Section 4.1
and Table 1). For this calculation, a misloaded waste package is defined as a waste package that

The "Summary Report of Commercial Reactor Criticality Data for Davis-Besse Unit 1" contains the detailed information necessary to perform commercial reactor criticality (CRC) analyses for the Davis-Besse Unit 1 reactor.

The preclosure criticality analysis process described in this technical report provides a systematic approach for determining the need for criticality controls and for evaluating their effectiveness during the preclosure period of the Monitored Geologic Repository at Yucca Mountain, Nevada.

This report presents studies to assess reactivity margins and loading curves for pressurized water reactor
(PWR) burnup-credit criticality safety evaluations. The studies are based on a generic high-density 32-
assembly cask and systematically vary individual calculational (depletion and criticality) assumptions to
demonstrate the impact on the predicted effective neutron multiplication factor, keff, and burnup-credit
loading curves. The purpose of this report is to provide a greater understanding of the importance of

Past efforts by the Department of Energy (DOE), the Electric Power Research Institute (EPRI), the Nuclear Regulatory Commission (NRC), and others have provided sufficient technical information to enable the NRC to issue regulatory guidance for implementation of pressurized-water reactor (PWR) burnup credit; however, consideration of only the reactivity change due to the major actinides is recommended in the guidance.

The purpose of this analysis is to evaluate the transient behavior and consequences of a worst case criticality event involving intact pressurized water reactor (PWR) spent nuclear fuel (SNF) in a degraded basket configuration inside a 21 PWR assembly waste package (WP). The objective of this analysis is to demonstrate that the consequences of a worst case criticality event involving intact PWR SNF are insignificant in their effect on the overall radioisotopic inventory in a WP. An internal WP criticality is modeled in a manner analogous to transient phenomena in a nuclear reactor core.

The purpose of this calculation report, Range of Applicability and Bias Determination for Postclosure
Criticality of Commercial Spent Nuclear Fuel, is to validate the computational method used to perform
postclosure criticality calculations. The validation process applies the criticality analysis methodology
approach documented in Section 3.5 of the Disposal Criticality Analysis Methodology Topical Report.1
The application systems for this validation consist of waste packages containing transport, aging, and

The purpose of this calculation is to document the LaSalle Unit 1 boiling water reactor (BWR) fuel depletion calculations performed as part of the commercial reactor critical (CRC) evaluation program. The CRC evaluations constitute benchmark calculations that support the development and validation of the neutronics models used for criticality analyses involving commercial spent nuclear fuel in a geologic repository. This calculation incorporates control blade effects and minor variations in the SAS2H assembly modeling.

The purpose of this calculation is to document the Sequoyah Unit 2 pressurized water reactor (PWR) fuel depletion calculations performed as part of the commercial reactor critical (CRC) evaluation program. The CRC evaluations support the development and validation of the neutronics models used for criticality analyses involving commercial spent nuclear fuel in a geologic repository.

The purpose of this calculation is to perform criticality evaluations for mixed oxide spent nuclear fuel (MOX SNF) in 12 and 21 Pressurized Water Reactor (PWR) waste packages (WPs) for both intact and degraded configurations.
The MOX assembly design considered in previous studies on Pu disposition in commercial reactors is based on the Westinghouse (W) 17x17 Vantage 5 assembly (Ref. 7.2). Depletion analyses of four Pu enrichment and burnup (expressed as gigawatt days/metric ton heavy metal; GWd/MTHM) combinations were performed in Reference 7.4. These are:

This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to determine the viability of the Defense High-Level Waste (DHLW) Glass waste package concept with respect to criticality regulatory requirements in compliance with the goals of the Waste Package Implementation Plan (Ref. 5.1) for conceptual design. These design calculations are performed in sufficient detail to provide a comprehensive comparison base with other design alternatives.

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

The purpose of this analysis is to document the Commercial Reactor Critical (CRC) benchmark evaluation performed for the Quad Cities Unit 1 boiling water reactor (BWR). The CRC benchmark is performed at a beginning of life (BOL) statepoint representing reactor start-up critical conditions. The objective of this CRC benchmark analysis is to provide a validation benchmark for the MCNP 4A analytic tool for use in the disposal criticality analysis methodology.

The purpose of this analysis is to document the MCNP evaluations of benchmark lattice Laboratory Critical Experiments (LCE's). The objective of this analysis is to quantify the MCNP 4A (Reference 5.4) code system's ability to accurately calculate the effective neutron multiplication factor (keff) for various measured critical (i.e., keff= 1.0) configurations. This analysis quantifies the effectiveness of the MCNP criticality calculation for lattice configurations containing U02 and Pu02 fissile oxide fuel using two different cross section data libraries.

This technical report provides an updated summary of the waste package (WP) external criticalityrelated
risk of the plutonium disposition ceramic waste form, which is being developed and
evaluated by the Office of Fissile Materials Disposition of the U.S. Department of Energy (DOE).
The ceramic waste form consists of Pu immobilized in ceramic disks, which would be embedded
in High-Level Waste (HLW) glass in the HLW glass disposal canisters, known as the "can-incanister"

Optimization of energy resources suggests increased fuel residence in reactor cores and hence improved
fission product evaluations are required. For thermal reactors the fission product cross sections in the JEF2.2 and
JEFF3.1 libraries plus new evaluations from WPEC23 are assessed through modelling the CERES experiment in
the DIMPLE reactor. The analysis uses the lattice code WIMS10. Cross sections for 12 nuclides are assessed. The
thermal cross section and low energy resonance data for 147,152Sm and 155Gd are accurate to within 4%. Similar data

U.S. Nuclear Regulatory Commission Interim Staff Guidance 8 (ISG-8) for burnup credit covers actinides only, a position based primarily on the lack of definitive critical experiments and adequate radiochemical assay data that can be used to quantify the uncertainty associated with fission product credit.

The purpose of this calculation is to perform partially and fully degraded mode criticality evaluations of plutonium disposed of in a ceramic waste form and emplaced in a Monitored Geologic Repository. The partially degraded mode is represented by the immobilized plutonium ceramic discs piled in the bottom of the waste package (WP) while neutron absorbers begin to leach out of the discs.

The purpose of this calculation is to document the validity of the commercial reactor criticals (CRC) as a source for a spent nuclear fuel benchmark, and to characterize the neutronic similarities between a CRC and a waste package (WP). This report illustrates comparisons of neutron spectrum and the effects on criticality arising from physical differences between a WP and a CRC. This report is an engineering calculation supporting the development of the disposal criticality analysis methodology, performed under Quality Administrative Procedure (QAP)-3-15 Revision 0.

This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to compare the criticality potential of Westinghouse 17x17 mixed oxide (MOX) PWR fuel with the Design Basis spent nuclear fuel (SNF) analyzed previously (Ref. 5.1, 5.2). The basis of comparison will be the conceptual design Multi- Purpose Canister (MPC) PWR waste package concepts.