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Technical Evaluation Report on the Content of the U.S. Department of Energy's Yucca Mountain Repository License Application
Technical Evaluation Report on the Content of the U.S. Department of Energy's Yucca Mountain Repository License Application
This “Technical Evaluation Report on the Content of the U.S. Department of Energy’s Yucca Mountain License Application; Postclosure Volume: Repository Safety After Permanent Closure” (TER Postclosure Volume) presents information on the NRC staff’s review of DOE’s Safety Analysis Report (SAR), provided on June 3, 2008, as updated by DOE on February 19, 2009. The NRC staff also reviewed information DOE provided in response to NRC staff’s requests for additional information and other information that DOE provided related to the SAR.
Overview of the Nuclear Regulatory Commission and Its Regulatory Process for the Nuclear Fuel Cycle for Light Water Reactors
Overview of the Nuclear Regulatory Commission and Its Regulatory Process for the Nuclear Fuel Cycle for Light Water Reactors
This paper provides a brief description of the United States Nuclear Regulatory Commission (NRC) and its regulatory process for the current nuclear fuel cycle for light water power reactors (LWRs). It focuses on the regulatory framework for the licensing of facilities in the fuel cycle. The first part of the paper provides an overview of the NRC and its regulatory program including a description of its organization, function, authority, and responsibilities.
Monitored Retrievable Storage Facility Design Criteria Policy Document - 2nd Draft
Monitored Retrievable Storage Facility Design Criteria Policy Document - 2nd Draft
Monitored Retrievable Storage Facility Conceptual Design Report
Monitored Retrievable Storage Facility Conceptual Design Report
Preliminary Site Requirements and Considerations for a Monitored Retrievable Storage Facility
Preliminary Site Requirements and Considerations for a Monitored Retrievable Storage Facility
In the November 1989 Report to Congress on Reassessment of the Civilian
Radioactive Waste Management Program (DOE/RW-0247), the Secretary of Energy
announced an initiative for developing a monitored retrievable storage (MRS) facility
that is to start spent-fuel acceptance in 1998. This facility, which will be licensed by
the U.S. Nuclear Regulatory Commission (NRC), will receive spent fuel from
commercial nuclear power plants and provide a limited amount of storage for this
Preliminary Feasibility Assessment for Several Specific MRS Design Alternatives with the Potential for Early Deployment Revision1
Preliminary Feasibility Assessment for Several Specific MRS Design Alternatives with the Potential for Early Deployment Revision1
This vintage 1990 document presents the results of WESTON'S preliminary assessment of the feasibility of several alternative fuel-transfer and storage concepts that have the potential for early spent-fuel acceptance at an MRS facility. The feasibility study was part of a series of studies conducted by the U.S. Department of Energy (DOE) during the late 1980's and early 1990's in an effort to establish an MRS design configuration.
MRS Siting Briefing
MRS Siting Briefing
This briefing paper is a component of the comprehensive briefing package developed for the Negotiator, and describes previous DOE experience in its attempt to site an MRS facility. The Background section highlights, in chronological order, significant events in DOE's MRS siting history from enactment of the Nuclear Waste Policy Act of 1982 to the issuance of the "Report to Congress on Reassessment of the Civilian Radioactive Waste Management Program" in November 1989.
Screening Analysis of Criticality Features, Events, and Processes for License Application
Screening Analysis of Criticality Features, Events, and Processes for License Application
Monitored Retrievable Storage Submission to Congress-Rev. 1
Monitored Retrievable Storage Submission to Congress-Rev. 1
In response to Section 141 of the Nuclear Waste Policy Act of 1982, the Department of Energy hereby submits a proposal for the construction of a facility for monitored retrievable storage (MRS). The approval of this proposal by the Congress would specifically--
• Approve the construction of an MRS facility at a site on the Clinch River in the Roane County portion of Oak Ridge, Tennessee.
• Limit the storage capacity at the MRS site to 15,000 metric tons of uranium.
NRC Waste Confidence Rulemaking, Federal Register, 1984, 1990, 1999, and 2008
NRC Waste Confidence Rulemaking, Federal Register, 1984, 1990, 1999, and 2008
NRC Waste Confidence Rulemaking, Federal Register, 1984, 1990, 1999, and 2008
Managing the Nation's Commercial High-Level Radioactive Waste
Managing the Nation's Commercial High-Level Radioactive Waste
With the passage of the Nuclear Waste Policy Act of 1982 (NWPA), Congress for the first time established in law a comprehensive Federal policy for commercial high-level radioactive waste management, including interim storage and permanent disposal. NWPA provides sufficient authority for
developing and operating a high-level radioactive waste management system based on disposal in mined geologic repositories. Authorization
Nuclear Waste: Is There a Need for Federal Interim Storage--Executive Summary--Report of the Monitored Retrievable Storage Commission
Nuclear Waste: Is There a Need for Federal Interim Storage--Executive Summary--Report of the Monitored Retrievable Storage Commission
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 ISG-1: Classifying the Condition of Spent Nuclear Fuel for Interim Storage and Transportation Based on Function
NRC ISG-1: Classifying the Condition of Spent Nuclear Fuel for Interim Storage and Transportation Based on Function
This Interim Staff Guidance (ISG) provides guidance to the staff on classifying spent nuclear
fuel as either (1) damaged, (2) undamaged, or (3) intact, before interim storage or
transportation. This is not a regulation or requirement and can be modified or superseded by
an applicant with supportable technical arguments.
Revision 2
NRC SFST ISG-3: Post Accident Recovery and Compliance with 10 CFR 72.122(l)
NRC SFST ISG-3: Post Accident Recovery and Compliance with 10 CFR 72.122(l)
Compliance with 10 CFR 72.122(l) has been interpreted to mean that a licensee, during any
point in the storage cycle, must have a means of retrieving and repackaging individual fuel
assemblies even after an accident. The staff has reevaluated this interpretation.
NRC SFST ISG-4: Cask Closure Weld Inspections
NRC SFST ISG-4: Cask Closure Weld Inspections
The closure weld for the outer cover plate for austenitic stainless steel designs may be
inspected using either volumetric or multiple pass dye penetrant techniques subject to the
following conditions:
• Dye penetrant (PT) examination may only be used in lieu of volumetric
examination only on austenitic stainless steels. PT examination should be done
in accordance with ASME Section V, Article 6, “Liquid Penetrant Examination.”
• For either ultrasonic examination (UT) or PT examination, the minimum
NRC SFST ISG-5: Confinement Evaluation
NRC SFST ISG-5: Confinement Evaluation
Several changes have occurred since the issuance of NUREG-1536, “Standard Review Plan
(SRP) for Dry Cask Storage Systems,” that affect the staff’s approach to confinement
evaluation. The attachment to this ISG integrates the current staff approach into a revision of
ISG-5. The highlights of the changes include:
• Reflects October 1998 revisions to 10 CFR 72.104 and 10 CFR 72.106.
• Expands and clarifies acceptance criteria associated with confinement analysis and
acceptance of “leak tight” testing instead of detailed confinement analysis.
NRC SFST ISG-6: Establishing minimum initial enrichment for the bounding design basis fuel assembly(s)
NRC SFST ISG-6: Establishing minimum initial enrichment for the bounding design basis fuel assembly(s)
The Standard Review Plan, NUREG-1536, Chapter 5, Section V, 2 recommends that “the
applicant calculate the source term on the basis of the fuel that will actually provide the
bounding source term,” and states that the applicant should, “either specify the minimum initial
enrichment or establish the specific source terms as operating controls and limits for cask use.”
A specified source term is difficult for most cask users to determine and for inspectors to verify.
NRC SFST ISG-7: Potential Generic Issue Concerning Cask Heat Transfer in a Transportation Accident
NRC SFST ISG-7: Potential Generic Issue Concerning Cask Heat Transfer in a Transportation Accident
Staff raised two major issues concerning the adverse effects of fission gases to the gas-mixture
thermal conductivity in a spent fuel canister in a post accident environment. The two major
concerns were: (1) the reduction of the thermal conductivity of the canister gas by the mixing of
fission gases expelled from failed fuel pins and (2) the resultant temperature and pressure rise
within the canister. Since the fission gas is typically of a lower conductivity than the cover gas,
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-12: Buckling of Irradiated Fuel Under Bottom End Drop Conditions
NRC SFST ISG-12: Buckling of Irradiated Fuel Under Bottom End Drop Conditions
Fuel rod buckling analyses under bottom end drop conditions have traditionally been performed
to demonstrate integrity of the fuel following a cask drop accident. The methodology described
by Lawrence Livermore National Laboratory (LLNL) to analyze the buckling of irradiated spent
fuel assembly under a bottom end drop in their report UCID-21246 is a simplified approach. It
assumed that buckling occurred when the fuel rod segment between the bottom two spacer
grids reached the Euler buckling limit. The weight of fuel pellets was neglected in the analysis;