Understanding the changing nuclear and mechanical characteristics of used nuclear fuel (UNF) over time and how these changing characteristics affect storage, transportation, and disposal options can require many tools and types of data. To streamline analysis capabilities for the waste management system, a comprehensive, integrated data and analysis tool has been assembled—UNF-Storage, Transportation & Disposal Analysis Resource and Data System (UNF-ST&DARDS).

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.

The purpose of this calculation is to apply the process described in the TDR-DS0-NU-000001 Rev. 02, Preclosure Criticality Analysis Process Report (Ref. 2.2.25) to aid in establishing design and operational criteria important to criticality safety and to identify potential control parameters and their limits important to the criticality safety of commercial spent nuclear fuel (CSNF) handling operations in the Wet Handling Facility (WHF)

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;

GAO Storage Study Request

The Disposal Criticality Analysis Methodology Topical Report (YMP 2003) presents the methodology for evaluating potential criticality situations in the monitored geologic repository. As stated in the referenced Topical Report, the detailed methodology for performing the disposal criticality analyses will be documented in model reports. Many of the models developed in support of the Topical Report differ from the definition of models as given in the Office of Civilian Radioactive Waste Management procedure AP-SIII.10Q, Models, in that they are procedural, rather than mathematical.

In response to the remand of the U.S. Court of Appeals for the District of Columbia Circuit (Minnesota v. NRC, 602 F.2d 412 (1979)), and as a continuation of previous proceedings conducted in this area by NRC (44 Fed. Reg. 61,372), the Commission initiated a generic rulemaking proceeding on October 25, 1979.

The main question before the Transportation and Storage Subcommittee was whether the United States
should change its approach to storing and transporting spent nuclear fuel (SNF) and high-level
radioactive waste (HLW) while one or more permanent disposal facilities are established.
To answer this question and to develop specific recommendations and options for consideration by the
full Commission, the Subcommittee held multiple meetings and deliberative sessions, visited several

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

The purpose of this calculation is to perform waste-form specific nuclear criticality safety calculations to aid in establishing criticality safety design criteria, and to identify design and process parameters that are potentially important to the criticality safety of Department of Energy (DOE) standardized Spent Nuclear Fuel (SNF) canisters.

The purpose of this document is to present the methodology to be used for development of the Subcritical Limit (SL) for post closure conditions for the Yucca Mountain repository. The SL is a value based on a set of benchmark criticality multiplier, keff> results that are outputs of the MCNP calculation method. This SL accounts for calculational biases and associated uncertainties resulting from the use of MCNP as the method of assessing kerr·

The purpose of the material degradation and release (MDR) model is to predict the fate of the waste package materials, specifically the retention or mobilization of the radionuclides and the neutron-absorbing material as a function of time after the breach of a waste package during the 10,000 years after repository closure. The output of this model is used directly to assess the potential for a criticality event inside the waste package due to the retention of the radionuclides combined with a loss of the neutron-absorbing material.

The attached documents are an extensive list of references relevant to burnup credit criticality analysis. Some of the references may be available within the CURIE document collection.