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Waste Package Probabilistic Criticality Analysis: Summary Report of Evaluations in 1997
Waste Package Probabilistic Criticality Analysis: Summary Report of Evaluations in 1997
The emplacement of nuclear waste in the proposed geologic repository must satisfy relevant regulatory requirements with respect to criticality, 10CFR60. I31 (h) (Ref. 25). The waste packages for the various waste forms will be designed to preclude criticality (typically by the inclusion of neutron absorbers) even if the waste package becomes filled with water. Criticality may, however, be possible if the contents of the waste package become degraded in such a way that the fissile material can be separated from the neutron absorbers, while sufficient moderator is retained.
Frequency of SNF Misload for Uncanistered Fuel Waste Package
Frequency of SNF Misload for Uncanistered Fuel Waste Package
The purpose ofthis engineering calculation is to estimate the frequency of misloading spent nuclear fuel (SNF) assemblies that would result in exceeding the criticality design basis of a waste package (WP). This type of misload - a reactivity misload - results from the incorrect placement of one or more fuel assemblies into a waste package such that the criticality controls do not match the required controls for the fuel assemblies.
External Criticality Risk of Immobilized Plutonium Waste Form in a Geologic Repository
External Criticality Risk of Immobilized Plutonium Waste Form in a Geologic Repository
This technical report provides an updated summary of the waste package (WP) external criticalityrelated
risk of the plutonium disposition ceramic waste form, which is being developed and
evaluated by the Office of Fissile Materials Disposition of the U.S. Department of Energy (DOE).
The ceramic waste form consists of Pu immobilized in ceramic disks, which would be embedded
in High-Level Waste (HLW) glass in the HLW glass disposal canisters, known as the "can-incanister"
NRC Waste Confidence Positions
NRC Waste Confidence Positions
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.
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
Plutonium Fuel: An Assessment Report by an Expert Group
Plutonium Fuel: An Assessment Report by an Expert Group
Ever since the 1950s, plutonium, used in fas reactors, has been seen as the key to unlocking the vast energy resource contained in the the world's uranium reserves. However, the reductions in expected nuclear reactor installation rates, combined with discovery of additional uranium, have led to a lengthening in the perceived time interval before fast reactors, the most effective users of plutonium, will make large demands on plutonium supplies. THere are several options concerning its use or storage in the meantime.
Evaluation of Cross-Section Sensitivities in Computing Burnup Credit Fission Product Concentrations
Evaluation of Cross-Section Sensitivities in Computing Burnup Credit Fission Product Concentrations
U.S. Nuclear Regulatory Commission Interim Staff Guidance 8 (ISG-8) for burnup credit covers actinides only, a position based primarily on the lack of definitive critical experiments and adequate radiochemical assay data that can be used to quantify the uncertainty associated with fission product credit.
Dry Transfer Facility Criticality Safety Calculations
Dry Transfer Facility Criticality Safety Calculations
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.
slides - Observations on Key Storage and Transport Technical Issues
slides - Observations on Key Storage and Transport Technical Issues
Presented at the NEI Used Fuel Management Conference, St. Petersburg, FL, May 7-9, 2013
Draft Report for Comment: Identification and Prioritization of the Technical Information Needs Affecting Potential Regulation of Extended Storage and Transportation of Spent Nuclear Fuel
Draft Report for Comment: Identification and Prioritization of the Technical Information Needs Affecting Potential Regulation of Extended Storage and Transportation of Spent Nuclear Fuel
he U.S. Nuclear Regulatory Commission (NRC) regulates storage of spent nuclear fuel (SNF) from commercial nuclear power plants. An increasing amount of the SNF in storage is in dry storage systems, mostly at current and decommissioned plants. As directed by the Commission (in SRM-COMSECY-10-0007; December 6, 2010), in expectation of continued use of dry storage for extended periods of time, the NRC staff is examining the technical needs and potential changes to the regulatory framework that may be needed to continue licensing of SNF storage over periods beyond 120 years.
Study of the Effect of Integral Burnable Absorbers for PWR Burnup Credit
Study of the Effect of Integral Burnable Absorbers for PWR Burnup Credit
The Interim Staff 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. This restriction eliminates a large portion of the currently discharged spent fuel assemblies from cask loading, and thus severely limits the practical usefulness of burnup credit.
Memo to Timothy Collins from Laurence Kopp - Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants
Memo to Timothy Collins from Laurence Kopp - Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants
Guidance concerning regulatory requirements for criticality analysis of new and spent fuel storage at light-water reactor power plants used by the Reactor Systems Branch.
CRC Depletion Calculations for Three Mile Island Unit 1
CRC Depletion Calculations for Three Mile Island Unit 1
The purpose of this calculation is to document the Three Mile Island Unit 1 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.
Waste Package Filler Material Testing Report
Waste Package Filler Material Testing Report
As part of the Mined Geologic Disposal System Waste Package Development design activities, it has been determined that it may be beneficial to add material to fill the otherwise free spaces remaining in waste package after loading high-level nuclear waste. The use of filler material will benefit criticality control in spent nuclear fuel waste packages, by the moderator displacement method.
CRC Reactivity Calculations for Quad Cities Unit 2
CRC Reactivity Calculations for Quad Cities Unit 2
In the development of a methodology to account for exposure effects on the reactivity of spent Boiling Water Reactor (BWR) fuel in the proposed Monitored Geologic Repository (MGR) at Yucca Mountain, the accuracy of the methods used to predict the inventories of fissile and fissionable nuclides as well as neutron poisons present in the spent fuel must be established. One aspect ofthis confirmatory effort is accomplished by performing benchmark problems for known in-reactor critical configurations - Commercial Reactor Criticals (CRCs).
Nuclear Waste Policy Act of 1982
Nuclear Waste Policy Act of 1982
An Act to provide for the development of repositories for the disposal of high-level radioactive waste and spent nuclear fuel, to establish a program of research, de- velopment, and demonstration regarding the disposal of high-level radioactive waste and spent nuclear fuel, and for other purposes.
Analysis of Mechanisms for Early Waste Package / Drip Shield Failure
Analysis of Mechanisms for Early Waste Package / Drip Shield Failure
The purpose of this analysis is to evaluate the types of defects or imperfections that could occur in a waste package or a drip shield and potentially lead to its early failure, and to estimate a probability of undetected occurrence for each type. An early failure is defined as the through-wall penetration of a waste package or drip shield due to manufacturing or handling-induced defects, at a time earlier than would be predicted by mechanistic degradation models for a defect-free waste package or drip shield.
EQ6 Calculations for Chemical Degradation of TRIGA Codisposal Waste PacKages
EQ6 Calculations for Chemical Degradation of TRIGA Codisposal Waste PacKages
The Monitored Geologic Repository Waste Package Operations 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 Training, Research, Isotopes, General Atomics (TRIGA) reactor (Ref. 1). The TRIGA SNF has been considered for disposal at the potential Yucca Mountain site.
EQ6 Calculation for Chemical Degradation of Shippingport LWBR (Th/U Oxide) Spent Nuclear Fuel Waste Packages
EQ6 Calculation for Chemical Degradation of Shippingport LWBR (Th/U Oxide) Spent Nuclear Fuel Waste Packages
The Monitored Geologic Repository (MGR) Waste Package Department 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 Light Water Breeder Reactor (LWBR) (Ref. 1). The Shippingport LWBR SNF has been considered for disposal at the potential Yucca Mountain site.
Monitored Retrievable Storage Facility Design Criteria Policy Document - 2nd Draft
Monitored Retrievable Storage Facility Design Criteria Policy Document - 2nd Draft
Evaluation of Internal Criticality of the Plutonium Disposition Ceramic Waste Form
Evaluation of Internal Criticality of the Plutonium Disposition Ceramic Waste Form
The purpose of this calculation is to perform partially and fully degraded mode criticality evaluations of plutonium disposed of in a ceramic waste form and emplaced in a Monitored Geologic Repository. The partially degraded mode is represented by the immobilized plutonium ceramic discs piled in the bottom of the waste package (WP) while neutron absorbers begin to leach out of the discs.
PWR Axial Burnup Profile Analysis
PWR Axial Burnup Profile Analysis
The purpose of this activity is to develop a representative “limiting” axial burnup profile for pressurized water reactors (PWRs), which would encompass the isotopic axial variations caused by different assembly irradiation histories, and produce conservative isotopics with respect to criticality. The effect that the low burnup regions near the ends of spent fuel have on system reactivity is termed the “end-effect”. This calculation will quantify the end-effects associated with Pressurized Water Reactor (PWR) fuel assemblies emplaced in a hypothetical 21 PWR waste package.
Criticality Consequence Calculation Involving Intact PWR MOX SNF in a Degraded 21 PWR Assembly Waste Package
Criticality Consequence Calculation Involving Intact PWR MOX SNF in a Degraded 21 PWR Assembly Waste Package
Waste Package, LCE, CRC, and Radiochemical Assay Comparison Evaluation
Waste Package, LCE, CRC, and Radiochemical Assay Comparison Evaluation
The purpose of this calculation is to document the validity of the commercial reactor criticals (CRC) as a source for a spent nuclear fuel benchmark, and to characterize the neutronic similarities between a CRC and a waste package (WP). This report illustrates comparisons of neutron spectrum and the effects on criticality arising from physical differences between a WP and a CRC. This report is an engineering calculation supporting the development of the disposal criticality analysis methodology, performed under Quality Administrative Procedure (QAP)-3-15 Revision 0.