One of the significant issues yet to be resolved for using
burnup credit ~BUC! for spent nuclear fuel ~SNF! is establishing
a set of depletion parameters that produce an adequately conservative
representation of the fuel’s isotopic inventory. Depletion
parameters ~such as local power, fuel temperature, moderator temperature,
burnable poison rod history, and soluble boron concentration!
affect the isotopic inventory of fuel that is depleted in a
pressurized water reactor ~PWR!. However, obtaining the detailed

The effect of fixed absorbers on the reactivity of pressurized water reactor (PWR) spent nuclear fuel (SNF) in support of burnup-credit criticality safety analyses is examined. A fuel assembly burned in conjunction with fixed absorbers may have a higher reactivity for a given burnup than an assembly that has not used fixed absorbers. As a result, guidance on burnup credit, issued by the U.S. Nuclear Regulatory Commission's Spent Fuel Project Office, recommends restricting the use of burnup credit to assemblies that have not used burnable absorbers.

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).

Utilization of burnup credit in criticality safety analysis for long-term disposal of spent
nuclear fuel allows improved design efficiency and reduced cost due to the large mass of fissile
material that will be present in the repository. Burnup-credit calculations are based on depletion
calculations that provide a conservative estimate of spent fuel contents (in terms of criticality
potential), followed by criticality calculations to assess the value of the effective neutron

The isotopic composition of mixed-oxide fuel (fabricated with both uranium and plutonium
isotopes) discharged from reactors is of interest to the Fissile Material Disposition Program. The
validation of depletion codes used to predict isotopic compositions of MOX fuel, similar to studies
concerning uranium-only fueled reactors, thus, is very important. The EEI-Westinghouse Plutonium
Recycle Demonstration Program was conducted to examine the use of MOX fuel in the San Onofre

The purpose of this report is to demonstrate that postclosure temperature limits can be met, and certain thermal characteristics of the postclosure thermal reference case can be preserved, with alternative thermal loading schemes. The analysis considers certain variations from the base case.waste stream, the predicted postclosure temperatures that develop within the rock mass due to these waste stream variations, and then compares these temperatures to postclosure temperature limits.

The purpose of these calculations is to characterize the criticality safety concerns for the storage of Fast Flux Test Facility (FFTF) nuclear fuel in a Department of Energy spent nuclear fuel (DOE SNF) canister in a co-disposal waste package. These results will be used to support the analysis that will be done to demonstrate concept viability related to use in the Monitored Geologic Repository (MGR) environment.

This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to provide an assessment of the present waste package design from a criticality risk standpoint. The specific objectives of this initial analysis are to:
1. Establish a process for determining the probability of waste package criticality as a function of time (in terms of a cumulative distribution function, probability distribution function, or expected number of criticalities in a specified time interval) for various waste package concepts;

The Disposal Criticality Analysis Methodology Topical Reporta prescribes an approach to the methodology for performing postclosure criticality analyses within the monitored geologic repository at Yucca Mountain, Nevada. An essential component of the methodology is the Configuration Generator Model for In-Package Criticality that provides a tool to evaluate the probabilities of degraded configurations achieving a critical state.

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.

Spent Fuel Project Office, Interim Staff Guidance - 8, Revision 2 - Burnup Credit in the Criticality Safety Analyses of PWR Spent Fuel in Transport
and Storage Casks

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.

The purpose of this report is to document and validate the external accumulation model that predicts accumulation of fissile materials in the invert, fractures and lithophysae in the rock beneath a degrading waste package containing spent nuclear fuel (SNF) in the monitored geologic repository at Yucca Mountain. (Lithophysae are hollow, bubblelike structures in the rock composed of concentric shells of finely crystalline alkali feldspar, quartz, and other materials (Bates and Jackson 1984 [DIRS 128109], p.

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.

This calculation file provides comparisons of one- and two-dimensional methods for calculating the isotopic content of spent nuclear fuel. The one-dimensional methods use the SAS2H sequence of SCALE 4.4a (Reference 7.1) and the SAS2 sequence of SCALE 5.0 (Reference 7.2). The two-dimensional method uses the TRITON control module along with the T-DEPL sequence of SCALE 5.0 (Reference 7.3). The SAS2H results for SCALE 4.4a are taken from Reference 7.4. Data from previous two-dimensional calculations (Reference 7.5) using CASM03 will also be used for comparisons with TRITON.

The purpose of this calculation is to document the Three Mile Island Unit 1 pressurized water reactor {PWR) reactivity calculations performed as part o f the commercial reactor critical (CRC) evaluation program. CRC evaluation reactivity calculations are performed at a number of statepoints, representing reactor start-up critical conditions at either beginning of life (BOL), beginning of cycle (BOC), or mid- cycle when the reactor resumed operation after a shutdown.

Disposal Criticality Analysis Methodology Topical Report describes a methodology for performing postclosure criticality analyses within the repository at Yucca Mountain, Nevada. An important component of the postclosure criticality analysis is the calculation of conservative isotopic concentrations for spent nuclear fuel. This report documents the isotopic calculation methodology. The isotopic calculation methodology is shown to be conservative based upon current data for pressurized water reactor and boiling water reactor spent nuclear fuel.

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.

This design calculation updates the previous criticality evaluation for the fuel handling, transfer, and staging operations to be performed in the Dry Transfer Facility (DTF) including the remediation area. The purpose of the calculation is to demonstrate that operations performed in the DTF and RF meet the nuclear criticality safety design criteria specified in the Project Design Criteria (PDC) Document (BSC 2004 [DIRS 171599], Section 4.9.2.2), the nuclear facility safety requirement in Project Requirements Document (Canori and Leitner 2003 [DIRS 166275], p.

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 Interim Staff Guidance on burnup credit for pressurized water reactor (PWR) spent nuclear fuel (SNF), issued by the United States Nuclear Regulatory Commission's (U.S. NRC) Spent Fuel Project Office, recommends the use of analyses that provide an "adequate representation of the physics" and notes particular concern with the "need to consider the more reactive actinide compositions of fuels burned with fixed absorbers or with control rods fully or partly inserted." In the absence of readily available information on the extent of control rod (CR) usage in U.S.

The purpose of this calculation is to document the Crystal River Unit 3 pressurized waste reactor (PWR) reactivity calculations performed as part of the commercial reactor critical (CRC) evaluation program. CRC evaluation reactivity calculations are performed at a number of statepoints, representing reactor start-up critical conditions at either beginning of life (BOL), beginning of cycle (BOC), or mid-cycle when the reactor resumed operation after a shutdown.

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