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;

This report presents the results of computer code benchmark simulations against spent fuel radiochemical assay
measurements from the Kansai Electric Ltd. Takahama-3 reactor published by the Japan Atomic Energy
Research Institute. Takahama-3 is a pressurized-water reactor that operates with a 17 × 17 fuel-assembly design.
Spent fuel samples were obtained from assemblies operated for 2 and 3 cycles and achieved a maximum burnup
of 47 GWd/MTU. Radiochemical analyses were performed on two rods having an initial enrichment of

Pressurized water reactor (PWR) burnup credit validation is
demonstrated using the benchmarks for quantifying fuel reactivity
decrements, published as Benchmarks for Quantifying Fuel Reactivity
Depletion Uncertainty, Electric Power Research Institute (EPRI)
report 1022909. This demonstration uses the depletion module
TRITON (Transport Rigor Implemented with Time-Dependent
Operation for Neutronic Depletion) available in the SCALE 6.1
(Standardized Computer Analyses for Licensing Evaluations) code

This report presents a methodology for validation of the isotopic
contents of spent light water reactor fuel for actinide-only burnup
credit with additional high-quality radiochemistry assay (RCA) data
obtained from the Yankee Rowe pressurized water reactor. The
additional Yankee Rowe RCA data were not included in previous
isotopic validation studies for burnup credit due to the difficulty of
accurately modeling the complex Yankee Rowe fuel assembly design
using the SAS2H one-dimensional sequence of the earlier SCALE

This paper provides insights into the neutronic similarities between a representative high-capacity rail-transport cask containing typical pressurized water reactor (PWR) spent nuclear fuel assemblies and critical reactor state-points, referred to as commercial reactor critical (CRC) state-points. Forty CRC state-points from five PWRs were analyzed, and the characteristics of CRC state-points that may be applicable for validation of burnup-credit criticality safety calculations for spent fuel transport/storage/disposal systems were identified.

This report is part of a report series designed to document benchmark-quality radiochemical isotopic
assay data against which computer code accuracy can be quantified to establish the uncertainty and bias
associated with the code predictions. The experimental data included in the report series were acquired
from domestic and international programs and include spent fuel samples that cover a large burnup range.
The measurements analyzed in the current report, for which experimental data is publicly available,

The purpose of this directory is to make available a convenient source of information on package designs approved by the U.S. Nuclear Regulatory Commission. To assist in identifying packages, an index by Model Number and corresponding Certificate of Compliance Number is included at the front of Volume 2. The report includes all package designs approved prior to the publication date of the directory as of September 2013.

Spent nuclear fuel comprises a fraction of the hazardous materials packages shipped annually in the United States. In fact, at the present time, fewer than 100 packages of spent nuclear fuel are shipped annually. At the onset of spent fuel shipments to the proposed Yucca Mountain, Nevada, repository, the U.S. Department of Energy (DOE) expects to ship 400 - 500 spent fuel transport casks per year over the life of the facility.

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.

The Monitored Geologic Repository (MGR) Waste Package Operations (WPO) of the Civilian Radioactive Waste Management System Management & Operating Contractor (CRWMS M&O) performed calculations to provide input for disposal of spent nuclear fuel (SNF) from the Shippingport Pressurized Water Reactor (PWR) (Ref. 1). The Shippingport PWR SNF has been considered for disposal at the proposed Yucca Mountain site.

The purpose of this calculation is to estimate volumes, masses, and surface areas associated with (a) an empty Department of Energy (DOE) 18-inch diameter, 15-ft long spent nuclear fuel (SNF) canister, (b) an empty DOE 24-inch diameter, 15-ft long SNF canister, (c) Shippingport Light Water Breeder Reactor (LWBR) SNF, and (d) the internal basket structure for the 18-in. canister that has been designed specifically to accommodate Seed fuel from the Shippingport LWBR.

The purpose of this calculation is to perform degraded mode criticality evaluations of plutonium disposed in a ceramic waste form and emplaced in a Monitored Geologic Repository (MGR). A 5 Defense High-Level Waste (DHLW) Canister Waste Package (WP) design, incorporating the can-in-canister concept for plutonium immobilization is considered for this calculation. Each HLW glass pour canister contains 7 tubes. Each tube contains 4 cans, with 20 ceramic disks (immobilized plutonium) in each.

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 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 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.

The purpose of this calculation file is to document criticality calculations performed on two different rod consolidation waste package designs. The results presented in this calculation file may be used to support further evaluation of the rod consolidation waste package design.

NRC initiated a research activity with the Center for Nuclear Waste Regulatory Analyses (CNWRA®) to develop a conceptual test plan for measuring the quantity of residual water remaining in a canister following vacuum drying to the criterion referenced in NUREG–1536. While residual water may be considered as unbound or bound (i.e., physi- or chemisorbed), the focus of this test plan is only the unbound water. This activity consists of the preparation of two technical letter reports. The first is the present report, which describes current industry drying practices and capabilities.

The purpose of this study is to provide insights into the neutronic similarities that may exist between a
generic cask containing typical spent nuclear fuel assemblies and commercial reactor critical (CRC) state-
points. Forty CRC state-points from five pressurized-water reactors were selected for the study and the
type of CRC state-points that may be applicable for validation of burnup credit criticality safety
calculations for spent fuel transport/storage/disposal systems are identified. The study employed cross-

This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) in response to a request received via a QAP-3-12 Design Input Data Request (Reference 5.1) from Waste Acceptance, Storage, & Transportation (WAST) Design (formerly MRSMPC Design). This design analysis is an answer to the Design Input Data Request to provide: Specific requirements for long-term criticality control.