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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
Analysis of Experimental Data for High Burnup PWR Spent Fuel Isotopic Validation--Calvert Cliffs, Takahama, and Three Mile Island Reactors
Analysis of Experimental Data for High Burnup PWR Spent Fuel Isotopic Validation--Calvert Cliffs, Takahama, and Three Mile Island Reactors
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,
Spent Nuclear Fuel Discharges from U.S. Reactors 1994
Spent Nuclear Fuel Discharges from U.S. Reactors 1994
A Critical Review of the Practice of Equating the Reactivity of Spent Fuel to Fresh Fuel in Burnup Credit Criticality Safety Analyses for PWR Spent Fuel Pool Storage
A Critical Review of the Practice of Equating the Reactivity of Spent Fuel to Fresh Fuel in Burnup Credit Criticality Safety Analyses for PWR Spent Fuel Pool Storage
This research examines the practice of equating the reactivity of spent fuel to that of fresh fuel for the purpose of performing burnup credit criticality safety analyses for PWR spent fuel pool (SFP) storage conditions. The investigation consists of comparing kf estimates based on reactivity "equivalent" fresh fuel enrichment (REFFE) to kl estimates using the actual spent fuel isotopics.
Dry Transfer System for Spent Fuel: Project Report: A System Designed to Achieve the Dry Transfer of Bare Spent Fuel Between Two Casks
Dry Transfer System for Spent Fuel: Project Report: A System Designed to Achieve the Dry Transfer of Bare Spent Fuel Between Two Casks
Use of an on-site dry transfer system (DTS) allows utilities with limited crane capacities or other plant restrictions to take advantage of large efficient storage systems. By using this system, utilities can also transfer fuel from loaded storage casks to transport casks without returning to their fuel storage pool.
Failure Modes and Effects Analysis (FMEA) of Welded Stainless Steel Canisters for Dry Cask Storage Systems
Failure Modes and Effects Analysis (FMEA) of Welded Stainless Steel Canisters for Dry Cask Storage Systems
Due to the delayed opening of a final geological repository for spent nuclear fuel, the lifespan of dry cask storage systems may be increased to 120 years or longer. To ensure safety over this extended period of interim storage, degradation mechanisms that have the potential to cause penetration of the canister confinement boundary must be evaluated and understood.
NRC SFST ISG-2: Fuel Retrievability
NRC SFST ISG-2: Fuel Retrievability
This Interim Staff Guidance (ISG) provides guidance to the staff for determining if
storage systems to be licensed under 10 CFR Part 72 allow ready retrieval of spent fuel.
This guidance is not a regulation or a requirement.
NRC SFST ISG-8: Burnup Credit in the Criticality Safety Analyses of PWR Spent Fuel in Transportation and Storage Casks
NRC SFST ISG-8: Burnup Credit in the Criticality Safety Analyses of PWR Spent Fuel in Transportation and Storage Casks
Title 10 of the Code of Federal Regulations (10 CFR) Part 71, Packaging and Transportation of
Radioactive Material, and 10 CFR Part 72, Licensing Requirements for the Independent
Storage of Spent Nuclear Fuel, High-Level Radioactive Waste, and Reactor-Related Greater
Than Class C Waste, require that spent nuclear fuel (SNF) remain subcritical in transportation
and storage, respectively. Unirradiated reactor fuel has a well-specified nuclide composition
that provides a straightforward and bounding approach to the criticality safety analysis of
NRC SFST ISG-9: Storage of Components Associated with Fuel Assemblies
NRC SFST ISG-9: Storage of Components Associated with Fuel Assemblies
The purpose of this ISG is to clarify the technical criteria for types of materials that will be |
considered associated with the storage of spent fuel assemblies. While control rods are |
mentioned in the Standard Review Plan as possible contents, specific information and guidance
is lacking.
Revision 1
NRC SFST ISG-10: Alternatives to the ASME Code
NRC SFST ISG-10: Alternatives to the ASME Code
There is no existing American Society of Mechanical Engineers (ASME) Code for the design
and fabrication of spent fuel dry storage casks. Therefore, ASME Code Section III, is
referenced by NUREG-1536, “Standard Review Plan for Dry Cask Storage Systems,” as an
acceptable standard for the design and fabrication of dry storage casks. However, since dry
storage casks are not pressure vessels, ASME Code Section III, cannot be implemented
without allowing some alternatives to its requirements.
Revision 1
NRC SFST ISG-11: Cladding Considerations for the Transportation and Storage of Spent Fuel
NRC SFST ISG-11: Cladding Considerations for the Transportation and Storage of Spent Fuel
The staff has broadened the technical basis for the storage of spent fuel including assemblies
with average burnups exceeding 45 GWd/MTU. This revision to Interim Staff Guidance No. 11
(ISG-11) addresses the technical review aspects of and specifies the acceptance criteria for
limiting spent fuel reconfiguration in storage casks. It modifies the previous revision of the ISG
in three ways: (1) by clarifying the meaning of some of the acceptance criteria contained in
NRC SFST ISG-14: Supplemental Shielding
NRC SFST ISG-14: Supplemental Shielding
Guidance regarding supplemental shielding that may be installed at an independent
spent fuel storage installation (ISFSI) to meet the requirements of 10 CFR 72.104(a).
NRC SFST ISG-17: Interim Storage of Greater Than Class C Waste
NRC SFST ISG-17: Interim Storage of Greater Than Class C Waste
Guidance is necessary on the interim storage of greater than Class C (GTCC) waste due to the
revision of Title 10 of the Code of Federal Regulations (10 CFR) Part 72. The revision to 10
CFR Part 72 is documented in final rule, ìInterim Storage for GTCC Waste,î and permits the
storage of GTCC wastes at independent spent fuel storage installations (ISFSI) or monitored
retrievable storage (MRS) facilities. The GTCC wastes, if stored at an ISFSI, must be in solid
form, and stored in a separate container (i.e., GTCC waste may not be stored in a cask that
NRC SFST ISG-18: The Design and Testing of Lid Welds on Austenitic Stainless Steel Canisters as the Confinement Boundary for Spent Fuel Storage
NRC SFST ISG-18: The Design and Testing of Lid Welds on Austenitic Stainless Steel Canisters as the Confinement Boundary for Spent Fuel Storage
The purpose of this ISG is to address the design and testing of the various closure welds (“lid
welds”) associated with the redundant closure of all-welded austenitic stainless steel canisters:
As an acceptable confinement boundary under 10 CFR Part 72.236(e) (Ref. 1) for
purposes of demonstrating no credible leakage of radioactive material during storage
and satisfying the dose limits under normal and off-normal conditions in 10 CFR Parts
72.104(a) and 72.106(b).
Industry Spent Fuel Storage Handbook
Industry Spent Fuel Storage Handbook
The Industry Spent Fuel Storage Handbook (“the Handbook”) addresses the relevant aspects of at-reactor spent (or used) nuclear fuel (SNF) storage in the United States. With the prospect of SNF being stored at reactor sites for the foreseeable future, it is expected that all U.S. nuclear power plants will have to implement at-reactor dry storage by 2025 or shortly thereafter. The Handbook provides a broad overview of recent developments for storing SNF at U.S. reactor sites, focusing primarily on at-reactor dry storage of SNF.
Technical Bases for Extended Dry Storage of Spent Nuclear Fuel
Technical Bases for Extended Dry Storage of Spent Nuclear Fuel
Independent spent fuel storage installations (ISFSIs) are currently licensed for 20 years. However, delays in developing permanent spent fuel disposal capability require continued ISFSI storage beyond the 20-year term. This report provides a technical basis for demonstrating the feasibility of extended spent fuel storage in ISFSIs.
Partnering for Long-term Management of Radioactive Waste-Evolution and Current Practice in Thirteen Countries
Partnering for Long-term Management of Radioactive Waste-Evolution and Current Practice in Thirteen Countries
Fission Product Benchmarking for Burnup Credit Applications
Fission Product Benchmarking for Burnup Credit Applications
Progress toward developing a technical basis for a cost-effective burnup credit methodology for
spent nuclear fuel with initial U-235 enrichment up to 5% is presented. Present regulatory
practices provide as much burnup credit flexibility as can be currently expected. Further progress
is achievable by incorporating the negative reactivity effects of a subset of neutron-absorbing
fission product isotopes. Progress also depends on optimizing the procedure for establishing the
Radioactive Waste Repositories and Host Regions: Envisaging the Future Together
Radioactive Waste Repositories and Host Regions: Envisaging the Future Together
Environmental Impact Statement Comments, Management of Commercial High-Level and Transuranium-Contaminated Radioactive Waste
Environmental Impact Statement Comments, Management of Commercial High-Level and Transuranium-Contaminated Radioactive Waste
This report summarizes the results of EPA's review of the AEC
draft environmental statement, "Management of Commercial High-Level
and Transuranium-Contaminated Radioactive Waste" (WASH-1539). The
means by which high-level and long-lived radioactive wastes are
managed constitutes one of the most important questions upon which
the public acceptability of nuclear power, with its social and economic
benefits, will be determined. While the generation of power by
nuclear means offers certain benefits from the environmental viewpoint,
HTC Experimental Program: Validation and Calculational Analysis
HTC Experimental Program: Validation and Calculational Analysis
In the 1980s a series of the Haut Taux de Combustion (HTC) critical experiments with fuel pins in a water-moderated lattice was conducted at the Apparatus B experimental facility in Valduc (Commissariat à l'Energie Atomique, France) with the support of the Institut de Radioprotection et de Sûreté Nucléaire and AREVA NC. Four series of experiments were designed to assess profit associated with actinide-only burnup credit in the criticality safety evaluation for fuel handling, pool storage, and spent-fuel cask conditions.
Validation Issues for Depletion and Criticality Analysis in Burnup Credit
Validation Issues for Depletion and Criticality Analysis in Burnup Credit
This paper reviews validation issues associated with implementation of burnup credit in transport, dry storage,
and disposal. The issues discussed are ones that have been identified by one or more constituents of the
United States technical community (national laboratories, licensees, and regulators) that have been exploring the
use of burnup credit. There is not necessarily agreement on the importance of the various issues, which
sometimes is what creates the issue. The broad issues relate to the paucity of available experimental data
Safety Evaluation Report for Disposal Criticality Analysis Methodology Topical Report, Revision 0
Safety Evaluation Report for Disposal Criticality Analysis Methodology Topical Report, Revision 0
In January 1999, the U.S. Department of Energy (DOE)/Office of Civilian Radioactive
Waste Management (OCRWM) submitted the Disposal Criticality Analysis Methodology
Topical Report, Revision 0 (TR) to the U.S. Nuclear Regulatory Commission (NRC) for
review and approval. The TR presents an overall approach for consideration of postclosure
disposal criticality of commercial and defense high-level waste to be placed at
the proposed Yucca Mountain site. During the course of the review and interactions
ANSI/ANS-8.27-2008: Burnup Credit for LWR Fuel
ANSI/ANS-8.27-2008: Burnup Credit for LWR Fuel
This standard provides criteria for accounting for reactivity effects of fuel irradiation and radioactive decay in criticality safety control of storage, transportation, and disposal of commercial LWR UO2 fuel assemblies.
This standard assumes the fuel and any fixed burnable absorbers are contained in an intact assembly. Additional considerations could be necessary for fuel assemblies that have been disassembled, consolidated, damaged, or reconfigured in any manner.