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slides - Operating Experience, Session I, Cask Cranes
slides - Operating Experience, Session I, Cask Cranes
Presented at the NEI Used Fuel Management Conference, St. Petersburg, FL, May 7-9, 2013
slides - Cumulative Impact of Industry and NRC Actions
slides - Cumulative Impact of Industry and NRC Actions
Presented at the NEI Used Fuel Management Conference, St. Petersburg, FL, May 7-9, 2013
Validation of important fission product evaluations through CERES integral benchmarks
Validation of important fission product evaluations through CERES integral benchmarks
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
Yankee Companies Urge Federal Government to Remove Used Nuclear Fuel in Wake of Favorable Federal Court Decision
Yankee Companies Urge Federal Government to Remove Used Nuclear Fuel in Wake of Favorable Federal Court Decision
U.S Court of Federal Claims Senior Judge James F. Merow issued a favorable ruling on September 30, 2006 for the three former Yankee nuclear power plants in their litigation with the federal government over its failure to remove used nuclear fuel from the three New England sites. Yankee Vice- President and Chief Financial Officer Michael Thomas said, “While the Court’s decision will need to be reviewed and evaluated, the Yankee companies’ initial reaction to the monetary award is very positive.
slides - NRC Management Perspectives
slides - NRC Management Perspectives
Presented at the NEI Used Fuel Management Conference, St. Petersburg, FL, May 7-9, 2013
Slides - Retrievability, Cladding Integrity, and Safety Handling during Storage and Transportation
Slides - Retrievability, Cladding Integrity, and Safety Handling during Storage and Transportation
Presented at the NEI Used Fuel Management Conference, St. Petersburg, FL, May 7-9, 2013
January 2013 Presentation to the Institute fo Nuclear Materials Management on Near Term Planning for Storage and Transportation of Used Nuclear Fuel
January 2013 Presentation to the Institute fo Nuclear Materials Management on Near Term Planning for Storage and Transportation of Used Nuclear Fuel
This is the Nuclear Fuels Storage and Transportation Project Director's presentation on Near Term Planning for Stroage and Transportation of Used Nuclear Fuel presented to the Institute of Nuclear Materials Management on January 14, 2013 in Arlington Va.
slides - International Perspectives on the Management of Spent Fuel from Power Reactors
slides - International Perspectives on the Management of Spent Fuel from Power Reactors
Presented at the NEI Used Fuel Management Conference, St. Petersburg, FL, May 7-9, 2013
Maine Yankee: A brief history of operation, decommissioning, and the interim storage of spent nuclear fuel
Maine Yankee: A brief history of operation, decommissioning, and the interim storage of spent nuclear fuel
A brief history of operation, decommissioning, and the interim storage of spent nuclear fuel
slides - Deep Borehole Disposal of Spent Fuel
slides - Deep Borehole Disposal of Spent Fuel
Presented at the NEI Used Fuel Management Conference, St. Petersburg, FL, May 7-9, 2013
Fission Product Experiment Program: Validation and Calculational Analysis
Fission Product Experiment Program: Validation and Calculational Analysis
From 1998 to 2004, a series of critical experiments referred to as the fission product (FP) experimental program was performed at the Commissariat à l'Energie Atomique Valduc research facility. The experiments were designed by Institut de Radioprotection et de Sûreté Nucléaire (IRSN) and funded by AREVA NC and IRSN within the French program supporting development of a technical basis for burnup credit validation.
Sensitivity and Uncertainty Analysis of Commercial Reactor Criticals for Burnup Credit
Sensitivity and Uncertainty Analysis of Commercial Reactor Criticals for Burnup Credit
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.
Review of Yucca mountain Disposal Criticality Studies
Review of Yucca mountain Disposal Criticality Studies
Review of Results for the OECD/NEA Phase VII Benchmark: Study of Spent Fuel Compositions for Long-Term Disposal
Review of Results for the OECD/NEA Phase VII Benchmark: Study of Spent Fuel Compositions for Long-Term Disposal
Actinide-Only Burnup Credit for Pressurized Water Reactor Spent Nuclear Fuel - I: Methodology Overview
Actinide-Only Burnup Credit for Pressurized Water Reactor Spent Nuclear Fuel - I: Methodology Overview
A conservative methodology is presented that would allow taking credit for burnup in the criticality safety analysis of spent nuclear fuel (SNF) packages. The method is based on the assumption that the isotopic concentration in the SNF and cross sections of each isotope for which credit is taken must be supported by validation experiments. The method allows credit for the changes in the 234U, 235U, 236U, 238U, 238Pu, 239Pu, 240Pu, 241Pu, 242Pu, and 241Am concentration with burnup. No credit for fission product neutron absorbers is taken. The methodology consists of five major steps:
Actinide-Only Burnup Credit for Pressurized Water Reactor Spent Nuclear Fuel - III: Bounding Treatment of Spatial Burnup Distributions
Actinide-Only Burnup Credit for Pressurized Water Reactor Spent Nuclear Fuel - III: Bounding Treatment of Spatial Burnup Distributions
A flat, uniform axial burnup assumption, preferred for its computational simplicity, does not always conservatively estimate the pressurized water reactor spent-fuel-cask multiplication factors. Rather, the reactivity effect of the significantly underburned fuel ends, usually referred to as the "end effect," can be properly treated by explicit modeling of the axial burnup distribution based on limiting axial burnup profiles.
Actinide-Only Burnup Credit for Pressurized Water Reactor Spent Nuclear Fuel - II: Validation
Actinide-Only Burnup Credit for Pressurized Water Reactor Spent Nuclear Fuel - II: Validation
The calculation of isotopic concentrations in spent nuclear fuel (SNF) assemblies and the subcritical multiplication factor of SNF packages are two of the essential requirements of the actinide-only burnup credit methodology. To justify the accuracy of the computed values, the code systems used to perform the calculations must be validated. Here, the techniques used for actinide-only burnup credit isotopic and criticality validation are presented and demonstrated.
Use of Reactor-Follow Data to Determine Biases and Uncertainties for PWR spent Nuclear Fuel
Use of Reactor-Follow Data to Determine Biases and Uncertainties for PWR spent Nuclear Fuel
Modeling BWR Spent-Fuel Isotopics with SAS2H and CASMO-3
Modeling BWR Spent-Fuel Isotopics with SAS2H and CASMO-3
Effects of Integral Burnable Absorbers on PWR Spent Nuclear Fuel
Effects of Integral Burnable Absorbers on PWR Spent Nuclear Fuel
Spent Fuel Burnup Credit in Casks: An NRC Perspective
Spent Fuel Burnup Credit in Casks: An NRC Perspective
Until now, the Nuclear Regulatory Commission's (NRC) approval of criticality safety evaluations for spent fuel in transport and storage casks has been based on analyzing the fuel as though it were fresh and without burnable poisons. The well-known nuclide composition of fresh fuel has provided a straightforward and bounding approach for showing that spent fuel systems will remain subcritical under normal and accident conditions. Burnup credit refers to the approval of criticality safety evaluations that consider the decrease in fuel reactivity caused by. irradiation in the reactor.