Category of Content
Siting Experience Documents Only
Publication Date
Subject Matter
Keywords
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.
Spent Fuel Project Office, ISG-8 - Limited Burnup Credit in the Criticality Safety Analyses of PWR Spent Fuel in Transport and Storage Casks
Spent Fuel Project Office, ISG-8 - Limited Burnup Credit in the Criticality Safety Analyses of PWR Spent Fuel in Transport and Storage Casks
Spent Fuel Project Office Interim Staff Guidance - 8
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
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
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
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
Report on Intact and Degraded Criticality for Selected Plutonium Waste Forms in a Geologic Repository, Volume II: Immobilized In Ceramic
Report on Intact and Degraded Criticality for Selected Plutonium Waste Forms in a Geologic Repository, Volume II: Immobilized In Ceramic
As part of the plutonium waste form development and down-select process, repository analyses have been conducted to evaluate the long-term performance of these forms for repository acceptance. Intact and degraded mode criticality analysis of mixed oxide (MOX) spent fuel is presented in Volume I, while Volume II presents the evaluations of the waste form containing plutonium immobilized in a ceramic matrix.
Report On External Criticality of Plutonium Waste Forms In A Geologic Repository
Report On External Criticality of Plutonium Waste Forms In A Geologic Repository
This report presents the analyses and results for the potential occurrence of external criticality events which could result from plutonium waste forms emplaced in a geologic repository similar to the one being developed at Yucca Mountain. The analyses evaluate both the MOX spent fuel and the immobilized plutonium waste forms in a repository if the waste package has degraded and if the fissile material has migrated to the invert and out into the far-field.
Spent Fuel Project Office, Interim Staff Guidance - 8, Revision 1, Burnup Credit in the Criticality Safety Analyses of PWR Spent Fuel in Transport and Storage Casks
Spent Fuel Project Office, Interim Staff Guidance - 8, Revision 1, Burnup Credit in the Criticality Safety Analyses of PWR Spent Fuel in Transport and Storage Casks
Spent Fuel Project Office, Interim Staff Guidance - 8, Revision 1
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-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;
NRC SFST ISG-13: Real Individual
NRC SFST ISG-13: Real Individual
The purpose of this guidance is to, (1) clarify the meaning of a real individual as used in 10
CFR 72.104, (2) specify how the applicant may perform dose evaluations beyond the controlled
area for site-specific and general Independent Spent Fuel Storage Installation (ISFSI) licenses,
and (3) clarify standard review plan (SRP) text regarding dose calculations outside the
controlled area.
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-15: MATERIALS EVALUATION
NRC SFST ISG-15: MATERIALS EVALUATION
Due, in part, to a number of material-related issues identified during dry cask storage system
(DCSS) and transportation package application reviews and field implementation, the staff has
recognized the need for specific guidance for the review of materials selected by the applicant
for its DCSS or transportation package.
NRC SFST ISG-16: Emergency Planning
NRC SFST ISG-16: Emergency Planning
Issuance of specific guidance for review of Emergency Plans for facilities licensed pursuant to
10 CFR Part 72 and removal of the reference to Regulatory Guide 3.67, "Standard Format and
Content for Emergency Plans for Fuel Cycle and Materials Facilities," as included in NUREG-
1567, Standard Review Plan for Spent Fuel Dry Storage Facilities (March 2000).
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).