Presented at the NEI Used Fuel Management Conference, St. Petersburg, FL, May 7-9, 2013

The first objective of this calculation is the identification of the degraded configurations of the Enhanced Design Alternatives (EDA) II design that have some possibility of criticality and that can occur within 10,000 years of placement in the repository. The next objective is to evaluate the criticality of these configurations and to estimate the probability of occurrence for those configurations that could support criticality.

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 the transportation, aging and disposal (TAD) canister-based systems.

Transport of the damaged core materials from the Unit 2 reactor of the Three
Mile Island Nuclear Power Station (TMI-2) to the Idaho National Engineering Laboratory
(INEL) for examination and storage presented many technical and institutional
challenges, including assessing the ability to transport the damaged core;
removing and packaging core debris in ways suitable for transport; developing a
transport package that could both meet Federal regulations and interface with the

Presented at the NEI Used Fuel Management Conference, St. Petersburg, FL, May 7-9, 2013

Slides - 2014 WM Symposia, March 2-6, 2014, Phoenix, AZ

This report attempts to summarize and consolidate the existing knowledge on axial
burnup distribution issues that are important to burnup credit criticality safety calculations.
Recently released Nuclear Regulatory Commission (NRC) staff guidance permits limited burnup
credit, and thus, has prompted resolution of the axial burnup distribution issue. The reactivity
difference between the neutron multiplication factor (keff) calculated with explicit representation

The purpose of this methods report is to document: (1) the origin, and the methods used in the development of a comprehensive list of features, events, and/or processes (FEPs) that could potentially affect the postclosure performance of the Yucca Mountain disposal system; (2) the methodology and guidance used to screen FEPs for inclusion or exclusion from Total System Performance Assessment for the License Application (TSPA-LA) analysis; (3) the methodology and guidance used to create scenario classes; and (4) compliance with NUREG-1804 (NRC 2003.

US policy for management of used nuclear fuel (UNF) and high level radioactive wastes (HLRW) is at a crossroads, and the success of new policy directions will depend in part on broad public acceptance and support. In this paper I provide an overview of the evidence concerning the beliefs and concerns of members of the American public regarding UNF and HLNW. I also characterize the evidence on American’s policy preferences for management of these materials.

Presented at the NEI Used Fuel Management Conference, St. Petersburg, FL, May 7-9, 2013

This report summarizes the results of an initial investigation into the uncertainties associated with the burnup records maintained by nuclear power plants. The results indicate that there is an overall uncertainty of about 2 percent in the burnup records, which must be accounted for in spent fuel applications.

Direct disposal of the large canisters currently being used by the commercial nuclear power industry is beyond the current experience base domestically and internationally and potentially represents many other significant engineering and scientific challenges. Pragmatically, it is reasonable to assume that the packages that will be disposed of in the future may be significantly different from what is being used for storage today.

Uncertainties in the predicted isotopic concentrations in spent nuclear fuel represent one of the largest
sources of overall uncertainty in criticality calculations that use burnup credit. The methods used to
propagate the uncertainties in the calculated nuclide concentrations to the uncertainty in the predicted
neutron multiplication factor (keff) of the system can have a significant effect on the uncertainty in the
safety margin in criticality calculations and ultimately affect the potential capacity of spent fuel transport

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 report presents a two-phased approach to develop and analyze a “thermal envelope” to represent the postclosure response of the repository to the anticipated range of repository design thermal loadings. In Phase 1 an estimated limiting waste stream (ELWS) is identified and analyzed to determine the extremes of average and local thermal loading conditions. The coldest thermal loading condition is represented by an emplacement drift loaded exclusively with high-level radioactive waste (HLW) and/or defense spent nuclear fuel (DSNF).

Burnup credit is an ongoing technical concern for many countries that operate commercial
nuclear power reactors. In a multinational cooperative effort to resolve burnup credit issues, a
Burnup Credit Working Group has been formed under the auspices of the Nuclear Energy Agency
of the Organization for Economic Cooperation and Development. This working group has
established a set of well-defined calculational benchmarks designed to study significant aspects of
burnup credit computational methods. These benchmarks are intended to provide a means for the

The Nuclear Waste Administration Act of 2013 discussion draft is intended to implement the recommendations of the Blue Ribbon Commission on America’s Nuclear Future to establish a nuclear waste administration and create a consent-based process for siting nuclear waste facilities. The bill enables the federal government to fulfill its commitment to managing nuclear waste, ending the costly liability the government bears for its failure to dispose of commercial spent fuel.

The purpose of this engineering calculation is to provide the chemical composition for the Department of Energy (DOE) Savannah River Site (SRS) High-Level Waste (HLW) glass. Since the glass is to be co-disposed with other DOE spent nuclear fuels (SNFs) in the Monitored Geologic Repository (MGR), its chemical composition is needed for the design of the co-disposal canisters and waste packages in term of criticality and degradation.

The objective of this calculation is to evaluate the peak temperatures due to thermal loading and boundary conditions of the TAD Waste Package design under nominal Monitored Geologic Repository conditions.

Unirradiated reactor fuel has a well-specified nuclide composition that provides a
straightforward and bounding approach to the criticality safety analysis of transport and storage
casks. As the fuel is irradiated in the reactor, the nuclide composition changes and, ignoring
the presence of burnable poisons, this composition change will cause the reactivity of the fuel to
decrease. Allowance in the criticality safety analysis for the decrease in fuel reactivity resulting

The purpose of this calculation is to establish the relative change in the effective neutron multiplication factor (keff) due to the use of MCNP unique identifiers (ZAIDs) in Nuclear Criticality Calculations for Canister-Based Facilities - DOE SNF (Reference 2.2.1, Attachment 3, MCNP inputs.zip) that are different to the ZAIDs used in the Analysis of Critical Benchmark Experiments and Critical Limit Calculation for DOE SNF (Reference 2.2.5, Table 5-3).

The purpose of this calculation is to characterize the criticality safety aspects of N-Reactor fuel stored in a Department of Energy spent nuclear fuel (DOE-SNF) canister that contains four Multi-Canister Overpacks (MCO's). These calculations will be done to support the analysis that will be done to demonstrate concept viability related to pre-emplacement storage and use in the Monitored Geologic Repository (MGR) environment for the pre-closure time frame.

The Range of Neutronic Parameters for Repository Criticality Analyses technical report contains a summary of the benchmark criticality analyses (including the laboratory critical experiment [LCEs] and the commercial reactor criticals [CRCs]) used to support the validation of the criticality evaluation methods. This report also documents the development of the Critical Limits (CLs) for the repository criticality analyses.

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