ML20321A101
| ML20321A101 | |
| Person / Time | |
|---|---|
| Issue date: | 04/30/2020 |
| From: | Office of Nuclear Material Safety and Safeguards |
| To: | |
| AWillis - 301.415.0479 ; SKumar | |
| Shared Package | |
| ML20321A086 | List:
|
| References | |
| NUREG-2215 | |
| Download: ML20321A101 (31) | |
Text
CRITICALITY EVALUATION Review Objective The objective of the U.S. Nuclear Regulatory Commissions (NRCs) review with regard to nuclear criticality safety is to ensure that spent nuclear fuel (SNF) to be placed into the dry storage under 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, remains subcritical under normal, off-normal, and accident conditions involving handling, packaging, transfer, and storage. This objective extends to the storage of high-level radioactive waste (HLW) at a specific license dry storage facility (DSF) that is a monitored retrievable storage installation (MRS). If reactor-related greater-than-Class-C (GTCC) waste is to be stored at a specific license DSF also storing SNF or HLW, then the review objective also includes ensuring that the storage of reactor-related GTCC waste does not adversely affect the safe storage of SNF and HLW and ensuring reactor-related GTCC waste remains subcritical if it includes fissile material. The objective also extends to other DSF structures, systems, and components (SSCs) in the specific license application for which criticality safety may be relevant (e.g., pools for storage or repackaging included as part of the DSF design).
Applicability This chapter of the Standard Review Plan (SRP) applies to the review of applications for specific licenses for an independent spent fuel storage installation (ISFSI) or an MRS, categorized as a DSF. It also applies to the review of applications for a certificate of compliance (CoC) of a dry storage system (DSS). Sections, paragraphs, or tables that apply only to DSF specific license applications have (SL) in the heading. Sections, paragraphs, or tables that apply only to DSS CoC applications have (CoC) in the heading. A subsection or paragraph without an identifier applies to both types of applications.
Areas of Review This chapter addresses the following areas of review:
x criticality design criteria and features x fuel specification
- fuel type
- nonfuel hardware (NFH)
- fuel condition x model specification
- configuration
- material properties x criticality analysis
- computer codes and cross section data
- neutron multiplication factor
- benchmark comparisons x burnup credit
- limits for the licensing basis
- licensing-basis model assumptions
- code validationisotopic depletion 7-1
- code validationkeff determination
- loading curve and burnup verification x reactor-related GTCC waste and HLW (SL) x supplemental information Regulatory Requirements and Acceptance Criteria This section summarizes those parts of 10 CFR Part 72 that are relevant to the review areas addressed by this chapter. The NRC staff reviewer should refer to the exact language in the regulations. Tables 7-1a and 7-1b match the relevant regulatory requirements to the areas of review this chapter covers. The reviewer should verify the association of regulatory requirements with the areas of review presented in the tables to ensure that no requirements are overlooked as a result of unique design features.
Table 7-1a Relationship of Regulations and Areas of Review for a DSF (SL) 10 CFR Part 72 Regulations Areas of Review 72.24 72.44(c) 72.124 72.40(a)(13)
Criticality Design Criteria and Features (b)(c)(g) (a)(b)(c)
Fuel Specifications (b)(c)(g) (a)(b)
Model Specification (d) (a)(b)
Criticality Analysis (d) (a)(b)
Burnup Credit (b)(c)(d)(g) (a)(b)
Reactor-Related GTCC Waste and (b)(c)(g) (a)
HLW Table 7-1b Relationship of Regulations and Areas of Review for a DSS (CoC) 10 CFR Part 72 Regulations Areas of Review 72.124 72.236(a) 72.236 Criticality Design Criteria and Features (a)(b)(c) (b)(c)(g)(h)(m)
Fuel Specification (a)(b) (b)(c)
Model Specification (a)(b) (b)(c)
Criticality Analysis (a)(b) (b)(c)
Burnup Credit (a)(b) (b)(c)(g)
The DSS or DSF SSCs must be designed to ensure the SNF remains subcritical under all credible conditions (10 CFR 72.124(a)). In general, the criticality evaluation seeks to ensure that a subcritical condition is maintained for the DSS or DSF design and operations by fulfilling the following acceptance criteria:
x The effective neutron multiplication factor, keff, including all biases and uncertainties at a 95-percent confidence level, should not exceed 0.95 under all credible normal, off-normal, and accident conditions for all storage operations (e.g., SNF handling, packaging, transfer, and storage).
7-2
x At least two unlikely, independent, and concurrent or sequential changes to the conditions essential to nuclear criticality safety, under normal, off-normal, and accident conditions, would need to occur before an accidental criticality is possible (i.e., double contingency principle; see 10 CFR 72.124(a)).
x When practicable, criticality safety of the design should be established on the basis of favorable geometry, permanently fixed neutron-absorbing materials (poisons), or both.
Where solid neutron-absorbing materials are used, the design should provide for a positive means to verify their continued efficacy during the storage period. The continued efficacy of the neutron-absorbing materials in the DSS or DSF storage containers may be confirmed by a demonstration or analysis before use, showing that significant degradation of these materials cannot occur over the life of the DSS or DSF (i.e., the certified or licensed period of storage). In other DSF SSCs, such as a pool, the neutron absorbers may be more likely to corrode; however, they will be more accessible.
Thus, appropriate periodic monitoring should be used to verify these absorbers continued efficacy.
x Criticality safety design may credit up to 90 percent of the neutron poison material in fixed neutron absorbers when subject to adequate acceptance testing (see Chapter 8, Materials Evaluation, Section 8.5.7, Criticality Control Materials, of this SRP).
x (SL) The DSF SSCs must be designed to ensure that reactor-related GTCC waste and HLW to be stored at the DSF and containing fissile material also remain subcritical under all credible conditions. The preceding criteria for SNF should be met, as applicable.
Review Procedures Figures 7-1a and 7-1b show the interrelationship between the criticality evaluation and the other areas of review described in this SRP for specific license and CoC applications, respectively.
To ensure that the DSS or DSF complies with 10 CFR Part 72, examine the criticality design features and criteria in the chapters of the applicants safety analysis report (SAR) on general information and principal design criteria, in addition to the chapter on criticality evaluation, for any additional details concerning criticality design features and criteria. Assess the bounding specifications for the SNF and assure consistency with the models the applicant used in the criticality analyses. Verify that the applicant has addressed criticality safety considerations under normal, off-normal, and accident conditions. In addition, verify that the criticality calculations determine the highest keff that might occur for all loading states under normal, off-normal, and accident conditions involving handling, packaging, transfer, and storage. To the extent practicable, use independent methods to perform any keff calculations to evaluate the applicants design. Review the operations descriptions to ensure the operations are consistent with and address the assumptions and parameters relied on in the criticality safety analyses, including any CoC or license conditions or technical specifications related to criticality safety.
(SL) The review guidance focuses mainly on the storage containers (i.e., the DSS for CoCs and the container design(s) to be used at the DSF for a specific license). However, for specific license applications, recognize that there may be other DSF SSCs for which criticality safety may be a concern and should be evaluated. Such SSCs would include any pool facilities used as part of operations for a specific license DSF (e.g., for loading, unloading, and repackaging of SNF) and included as part of the DSF design in the specific license application. Therefore, apply the 7-3
guidance in this chapter to the evaluation of these other DSF SSCs as applicable and appropriate.
The review guidance does address specific, unique considerations for these other DSF SSCs where necessary.
For evaluations that involve the use of industry standards, ensure the standards, including the revisions to the standards, are used in a manner consistent with the NRCs positions regarding those standards and their revisions. For example, in addition to items from specific industry standards addressed in this chapter, the NRC has documented its endorsements, including any exceptions, of various standards in Regulatory Guide 3.71, Nuclear Criticality Safety Standards for Fuels and Materials Facilities.
Chapter 1 - General Chapter 2 - Site Chapter 3 - Chapter 4 -
Information Characteristics Principal Design Structural Chapter 8 -
Evaluation Evaluation (SL) Criteria Evaluation Evaluation Materials Evaluation x DSF Description and x Meteorology x Classification of SSCs x SSCs Important to x Material Properties Operational Features x Surface and x Design Criteria for Safety safety x Environmental x Engineering Drawings Subsurface Hydrology Protection Systems x Other SSCs Subject Degradation; Chemical x Contents x Design Bases for SSCs to NRC Approval and Other Reactions Important to Safety x Fuel Cladding Integrity and x Design Criteria for Other Retrievability SSCs (SL) x Code Use and Quality Standards Chapter 7 - Criticality Evaluation (SL)
Criticality Design Criteria and Features Fuel Specification Burnup Credit x Fuel Type x Limits for the Licensing Basis Model Specification x Nonfuel Hardware x Licensing-Basis Model Assumptions x Configuration x Fuel Condition x Code Validation - Isotopic Depletion x Material Properties x Code Validation - keff Determination x Loading Curve and Burnup Verification Supplemental Information Criticality Analysis x Computer Codes and Cross Section Data Reactor-Related GTCC Waste and HLW x Neutron Multiplication Factor x Benchmark Comparisons Chapter 11 -
Chapter 12 - Chapter 17 -
Chapter 8 - Operation Chapter 16 -
Conduct of Technical Materials Procedures and Accident Analysis Operations Specifications Evaluation Systems Evaluation Evaluation Evaluation Evaluation x Material Properties x Operation Description x Acceptance Tests x Cause of Event x Functional and Operating x Environmental x Storage Container x Preoperational Testing x Detection of Event Limits, Monitoring Degradation; Chemical Loading and Startup Operations x Event Consequences Instruments, and Limiting and Other Reactions x Storage Container (SL) and Regulatory Control Settings x Fuel Cladding Integrity Handling and Storage x Maintenance Program Compliance x Limiting Conditions and Retrievability Operations x Surveillance x Code Use and Quality x Storage Container Requirements Standards Unloading x Design Features x Other Operating x Administrative Controls Systems (SL) x Operation Support Systems (SL)
Figure 7-1a Overview of Criticality Evaluation of Specific License Applications for a DSF (SL) 7-4
Chapter 1 -
Chapter 3 - Chapter 4 - Chapter 8 -
General Principal Criteria Structural Materials Information Evaluation Evaluation Evaluation Evaluation x DSS Description and x Classification of SSCs x SSCs Important to x Material Properties Operational Features x Design Bases for SSCs Safety x Environmental x Engineering Drawings Important to Safety x Other SSCs Subject Degradation; Chemical x Contents x Design Criteria for to NRC Approval and Other Reactions x Consideration of DSS Safety Protection x Fuel Cladding Integrity Transportability (CoC) Systems and Retrievability x Code Use and Quality Standards Chapter 7 - Criticality Evaluation (CoC)
Criticality Design Criteria and Features Fuel Specification Burnup Credit x Fuel Type x Limits for the Licensing Basis Model Specification x Nonfuel Hardware x Licensing-Basis Model Assumptions x Configuration x Fuel Condition x Code Validation - Isotopic Depletion x Material Properties x Code Validation - keff Determination x Loading Curve and Burnup Verification Supplemental Information Criticality Analysis x Computer Codes and Cross Section Data x Neutron Multiplication Factor x Benchmark Comparisons Chapter 11 -
Chapter 12 - Chapter 17 -
Chapter 8 - Operation Chapter 16 -
Conduct of Technical Materials Procedures and Accident Analysis Operations Specifications Evaluation Systems Evaluation Evaluation Evaluation Evaluation x Material Properties x Operation Description x Acceptance Tests x Cause of Event x Functional and x Environmental x Storage Container x Maintenance Program x Detection of Event Operating Limits, Degradation; Chemical Loading x Event Consequences Monitoring Instruments, and Other Reactions x Storage Container and Regulatory and Limiting Control x Fuel Cladding Integrity Handling and Storage Compliance Settings and Retrievability Operations x Limiting Conditions x Code Use and Quality x Storage Container x Surveillance Standards Unloading Requirements x Design Features x Administrative Controls Figure 7-1b Overview of Criticality Evaluation of Applications for a DSS (CoC) 7.5.1 Criticality Design Criteria and Features Examine the principal criticality design criteria presented in the chapter of the SAR on principal design criteria as well as any related details provided in the SAR chapter on criticality evaluation.
Examine the general storage container description presented in the SAR for any relevant information. Verify that the information in the chapter of the SAR on criticality evaluation is consistent with the information in the SARs chapters on general information and principal design criteria. Verify that all descriptions, drawings, figures, and tables are sufficiently detailed to support an indepth staff evaluation.
Criticality safety of the design must be based on favorable geometry, permanent fixed neutron absorbing materials, or both (10 CFR 72.124(b)). The criticality design of the storage container relies on the general dimensions of the containers components and the spacing of the fuel 7-5
assemblies. Tolerances for the material, fabrication, and assembly of SSCs can be important in identifying worst-case (lowest margin of safety) geometries, material compositions, and densities.
Ensure that the SAR uses the tolerances for the properties and construction of all SSCs involved in criticality analyses. Also ensure that the tolerances used in the analyses are identical to or conservatively bounding for the tolerances shown in the definition of the storage container design.
Verify that the analyses are based on the most conservative combination of tolerances.
The criticality design often relies on neutron poisons. These may be in the form of fixed poisons in the storage containers SNF basket structure, soluble poisons in the water of the SNF pool, or both. For fixed neutron-absorbing materials, the NRC has accepted a requirement for acceptance testing of the material during fabrication as a means for verifying the continued efficacy of solid neutron-absorbing materials incorporated in the SNF storage container (see also Section 8.5.7 of this SRP). During loading and unloading operations, the NRC staff accepts the use of borated water as a means of criticality control if the applicant specifies a minimum boron content and strict controls are established to ensure that the minimum required boron concentration is maintained.
This condition in turn becomes an operating control and limit in the SAR and in the CoC or license technical specifications. Include a discussion of these operation controls in the safety evaluation report (SER). Ensure that the technical specifications also include other design features significant to the criticality design, such as important basket dimensions that control the spacing of the fuel assemblies. These dimensions may be a minimum pitch for the basket cells or a minimum flux trap width.
If borated water is used for criticality control during loading and unloading operations, verify that the design and operations descriptions in the SAR include administrative controls or design features (with appropriate controls and design features included in the technical specifications), or both, to ensure that accidental flooding with unborated water is not credible. Otherwise, consider accidental flooding with unborated water. If the storage container is also intended for transport, the storage container design should not rely on borated water for criticality control. Borated water and any other liquids are not acceptable as a means of criticality control for a storage container in its dry storage configuration. This includes use of any credit in the criticality analysis for the presence of a liquid that may provide neutron shielding (and is external to the fuel basket);
however, its presence and most reactive density should be assumed if it increases keff. Also, if more than one certified or licensed basket design of the same supplier could fit in the storage container, then the type of basket to be used with the container should be stamped in a location on the container that allows for easy identification of the basket. Thus, the licensee will be able to easily verify the appropriateness of the fuel contents to be loaded in the basket.
The DSS or DSF SSCs must be designed so that at least two unlikely, independent, and concurrent or sequential changes to the conditions essential to criticality safety, under normal, off-normal, and accident conditions, must occur before an accidental criticality is possible (Double Contingency, as stated in American National Standards Institute (ANSI)/American Nuclear Society (ANS) 8.1, Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors; see 10 CFR 72.124(a)). For analysis, accidental criticality is defined as exceeding k eff of 0.95. Ensure that the criticality analysis demonstrates that keff is less than 0.95, with a 95 percent probability at the 95 percent confidence level, accounting for analysis uncertainty, bias, and bias uncertainty. Ensure that the applicant demonstrates that the double contingency criteria have been met for all configurations of the relevant DSS or DSF SSCs. For DSS or DSF storage container designs, these criteria are typically met by demonstrating a low likelihood of storage container failure and a low likelihood of flooding of the storage container to sufficient depth to cause criticality (i.e., to the height of the active fuel) in the containers dry storage configuration.
7-6
Other considerations and methods would be necessary to demonstrate that the double contingency criteria are met for other configurations of the storage container (e.g., during loading and unloading).
Under 10 CFR 72.124(c), a criticality monitoring system must be maintained in each area where special nuclear material is handled, used, or stored that will energize clearly audible alarm signals if accidental criticality occurs. This requirement does not apply while the special nuclear material is handled under water, including in a submerged storage container. The requirement also does not apply to dry storage areas where the storage container is in its dry storage configuration (i.e., drained, dried, and sealed closed). It is applicable when the storage container is removed from the pool during loading, until it is drained, dried, and sealed, and during unloading, beginning when the containers confinement barrier is no longer sealed (e.g., removal of vent or drain port covers).
Ensure that the criticality chapters of the SAR address how the criticality monitoring criteria will be met. The NRC has accepted the use of area radiation monitors, typically included in 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities, and 10 CFR Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants, reactor facilities spent fuel pool buildings, for meeting this requirement, provided the applicant can adequately justify that the area radiation monitors are sufficient to perform this function (in lieu of having a criticality monitoring system).
(SL) Ensure that the design of other relevant DSF SSCs, such as a pool that is part of the DSF design, are based on favorable geometry, permanently fixed absorber materials, or both. For fixed absorbers, work with the materials reviewer (SRP Chapter 8) to ensure the application provides appropriate means for verifying continued efficacy of the absorbers (e.g., periodic monitoring). If the pool also uses borated water, ensure that the facility design and operations include appropriate means to monitor and maintain the required boron concentrations for normal, off-normal, and accident conditions. Analyses of loss of soluble boron (boron dilution) may also be necessary. Ensure that the SAR demonstrates that the double contingency criteria are met for the pool and other relevant DSF SSCs under normal, off-normal, and accident conditions. Ensure appropriate controls and design features for the pool and other relevant SSCs are included in the license technical specifications. The preceding guidance should also be applied, consistent with the regulations, to all nonfuel materials to be stored at the DSF that include fissile material.
7.5.2 Fuel Specification Fuel Type Examine the specifications for the ranges or types of SNF that will be stored in the DSS or DSF storage containers as presented in the SAR chapters on general information and principal design criteria, as well as any related information in the SAR chapter on criticality. Verify that the SNF specifications given in the SAR chapter on criticality are consistent with, or bound, the specifications given in the SAR chapters on general information and principal design criteria and in the technical specifications. Keeping in mind that some specifications are more important than others, identify the specifications that are key to criticality safety, and verify that these are appropriately captured in the technical specifications. NUREG-1745, Standard Format and Content for Technical Specifications for 10 CFR Part 72 Cask Certificates of Compliance, lists some of the fuel specifications that may be key to maintaining the system subcritical, although others may be required. While NUREG-1745 discusses an option for controlling some parameters outside of the technical specifications (i.e., in the SAR only) and obtaining NRC 7-7
approval for contents alternatives regarding those parameters, the NRC has since determined that this option is not acceptable. Thus, any fuel parameters that are important to ensuring criticality safety should be captured in the license or CoC technical specifications.
Of primary interest is the type of fuel assemblies and maximum fuel enrichment that should be specified and used in the criticality calculations. Boiling-water reactors (BWRs) typically use multiple fuel pin enrichments, in which case the criticality calculations should use the maximum fuel pin enrichment present. Depending upon the fuel design, an applicant may propose use of assembly-averaged or lattice-averaged enrichments. This may be acceptable if the applicant can demonstrate that the applicants averaging technique is technically defensible and, for the criticality calculation, produces realistic or conservative results. Because of the natural uranium blankets present in many fuel designs, use of an assembly-averaged enrichment that includes the blankets is not normally considered appropriate or conservative.
Another parameter of interest is the fuel density assumed in the analysis. Ensure that the value of the fuel density used in the calculations is justified to be realistic or conservative.
Note that, while the majority of fuel assemblies burned in commercial reactors in the United States use uranium dioxide (UO2) as the fuel material, some fuel assemblies are made with mixed-oxide (MOX) fuel material. For MOX fuel, the material specification is typically given in terms of weight percent plutonium, and is further described by isotopic limits for the major plutonium isotopes important to criticality safety (i.e., plutonium-238, plutonium-239, plutonium-240, plutonium-241, and plutonium-242) along with the amount and maximum enrichment of the uranium in the fuel.
Plutonium-239 and plutonium-241 are fissile, and should have a maximum quantity or concentration limit, while the other plutonium isotopes are neutron absorbers, and should have minimum required mass ratios or concentrations. Alternatively, the applicant may choose to conservatively assume that all plutonium is fissile plutonium-239, which is acceptable.
Some commercial fuel assemblies may include thorium fuel rods in addition to UO2 rods. Ensure that the material specification for the thorium fuel rods includes the weight percent of the rods that is thorium oxide and uranium oxide and the maximum uranium enrichment. Note that thorium-232 will absorb neutrons to become uranium-233, a fissile nuclide. Therefore, fuel assemblies with thorium fuel material may become more reactive with irradiation in the reactor. Ensure that the SAR includes a depletion analysis for such fuel in order to determine the most reactive fuel composition.
Although the burnup of the fuel affects its reactivity, many criticality analyses have assumed the storage container to be loaded with fresh fuel (the fresh fuel assumption). Alternatively, the NRC staff has provided guidance for burnup credit for intact pressurized-water reactor (PWR) fuel. This guidance is limited to burnup credit available from specific actinide and fission product compositions associated with UO2 fuel of 5.0 weight percent or less enrichment that has been irradiated in a PWR to an assembly-average burnup value not exceeding 60 gigawatt days per metric ton of uranium (GWd/MTU) and cooled out of the reactor for a time period between 1 and 40 years. Section 7.5.5 of this SRP chapter provides guidance for the review of a criticality analysis that involves burnup credit. Ensure that the SAR chapter on technical specifications and operational controls and limits evaluation includes specifications for the fuel that will be stored in the storage container, including those important for burnup credit (e.g., minimum burnup versus enrichment, moderator temperature, in-core soluble boron concentrations, operations under control rod banks or with control rod insertion), if applicable. Also ensure that the SER contains this same information. Ensure that the license or CoC technical specifications explicitly list those specifications determined to be key to criticality safety.
7-8
For analyses that use the fresh fuel assumption, inadvertent loading of the storage container with unirradiated or low burnup fuel is not a major concern. However, inadvertent loading of the storage container with unirradiated or low burnup fuel is a major concern for container designs that rely on criticality analyses that use burnup credit. Therefore, ensure that the detailed loading procedures for these storage containers include steps to prevent misloading of unirradiated or low burnup fuel. Regardless of which analysis is used, ensure that detailed loading procedures include steps to prevent misloading for cases when fuel exceeding the design basis for the DSS or DSF storage containers, for a DSF being licensed to store SNF from a co-located 10 CFR Part 50 or 10 CFR Part 52 reactor facility, is present in the 10 CFR Part 50 or 10 CFR Part 52 facilitys pool at the time of DSS or DSF storage container loading.
(CoC) Because DSSs typically are designed to store many types and configurations of fuel assemblies, verify that the applicant has demonstrated that criticality requirements are satisfied for the most reactive case. A determination of which fuel is bounding in a criticality analysis depends on many factors and usually requires examination of several types of fuel assemblies and compositions. Note that the most reactive assembly type may be different for fresh fuel analyses in fresh water versus borated water, and if burnup of the fuel is credited according to the recommendations of Section 7.5.5 of this chapter. Therefore, verify that the applicant has demonstrated that the design-basis fuel assembly is the most reactive for the specific DSS design, including requested level of burnup credit, if applicable. Ensure that the SAR chapter on general information clearly indicates the design-basis assemblies. Also ensure that the SER contains this same information.
(SL) For specific license applications that include storage of multiple types and configurations of fuel assemblies, the considerations described above for CoC applications would also apply.
However, for a specific license DSF that is co-located with a 10 CFR Part 50 or 10 CFR Part 52 reactor facility, the SNF assembly types and configurations are likely to be limited to those associated with the co-located reactor facility, which may have used only one or two fuel types with limited enrichment ranges.
Nonfuel Hardware Some fuel assemblies may also have nonfuel components that are positioned or operated within the envelope of the fuel assembly during reactor operation that an applicant may seek to store with the assemblies in the SNF storage container. These items include PWR control assemblies, such as rod cluster control assemblies, control element assemblies, burnable poison rod (BPR) assemblies, and axial power shaping rods. Applicants may also seek approval for storing fuel assemblies with other items that extend into an assemblys active fuel region, such as stainless steel rod inserts used to displace water in PWR assembly guide tube dashpots. For applications that propose to load assemblies containing NFH, ensure that the analysis considers the effects of both inclusion and neglect of NFH on system reactivity. If the application relies on the presence of the NFH to meet the subcritical criterion, verify that the NFH will remain in place under all normal, off-normal, and accident conditions.
7-9
Generally, the NRC staff does not allow reliance on, or credit for, fuel-related burnable neutron absorbers. This restriction includes residual neutron-absorbing material remaining in the NFH loaded with an assembly. However, credit for any negative reactivity for this latter absorbing material may be accepted if all of the following is true:
- 1. The remaining absorbing material content is established through physical measurement or by calculation where a sufficient margin of safety is included commensurate with the uncertainty in the method of measurement or calculation.
- 2. The axial distribution of the poison depletion is adequately determined with appropriate margin for uncertainties.
- 3. Adequate structural integrity and placement of the nonfuel hardware under accident conditions is demonstrated.
Ensure that the fuel specifications described in the SAR chapter on technical specifications and operation controls and limits include the important details about the NFH to be stored with the fuel assemblies and the associated residual neutron-absorbing material. Also ensure that the SER contains this same information. Ensure that those details key to criticality safety are included in the CoC or license technical specifications, as appropriate. Also, verify that operating procedures are established that ensure that NFH loaded with assemblies meets the approved specifications and will remain in position under normal, off-normal, and accident conditions.
Fuel Condition Determine whether the applicant has included any specifications regarding the fuel condition. To date, a number of applications have requested approval for storage of fuel that is damaged as well as intact or undamaged. Consult Section 8.5.15.1, Spent Fuel Classification, of this SRP for the most current staff guidance for detailed descriptions of what constitutes damaged, undamaged, and intact fuel. This guidance gives the applicant the latitude to define fuel with defects (such as missing rods but not loose rods or debris) as undamaged fuel as long as the fuel can meet all the fuel-specific or system-related functions. For purposes of criticality safety, undamaged fuel is fuel that (1) is in the form of an assembly; (2) has structural and material properties such that the assembly can withstand normal, off-normal, and accident conditions while maintaining its geometric configuration; and (3) has had any damaged or missing fuel rods replaced with solid dummy rods that displace an equal or greater amount of water as the original rods. Fuel that cannot meet these criteria is considered to be damaged. However, a fuel assembly with missing fuel rods may be considered undamaged fuel if analyses are performed that show the criterion for subcriticality will be met with the fuel rods missing.
A fuel assembly that is classified as damaged should be placed in a damaged fuel canister, or in an acceptable alternative, for loading into the DSS or DSF storage container. For a storage container that is also intended for transport, keep in mind that the more severe conditions of transport may require reanalysis of assemblies classified as undamaged under storage-only conditions before transport. Confirm that specifications concerning the condition of the fuel to be stored in the DSS or DSF storage container and the loading of damaged fuel, as applicable, are included in the chapter of the SAR on technical specifications and operation controls and limits and in the CoC or license (in the technical specifications). Also ensure that the SER contains this same information.
7-10
Verify that the criticality analysis addresses the conditions of the fuel to be stored in the storage container. Ensure that the analyses for storage containers designed to store damaged fuel bound the configuration of the damaged fuel assemblies under all credible normal, off-normal, and accident conditions. For example, some analyses have performed calculations that model the damaged fuel as arrays of bare fuel rods (i.e., the cladding is assumed to be completely removed) having an optimized rod pitch.
7.5.3 Model Specification Verify that the applicant has specified manufacturing and fabrication tolerances. Verify that the applicant used the most reactive combination of tolerances, within the ranges of their acceptable values, in the analysis models.
Configuration Verify that SAR adequately describes the criticality models used to evaluate normal, off-normal, and accident conditions. Coordinate with the structural, materials, and thermal reviewers to understand any damage that could result from accident conditions, which include natural phenomena events.
Examine the sketches or figures of the models used for criticality calculations. Verify that the dimensions and materials of the models are consistent with the engineering drawings. Ensure that the SAR identifies any differences between the actual DSS or DSF storage container configurations and the models and demonstrates that the models are conservative. Substitution of end sections and support structures of the fuel with ordinary water, or a combination of water and steel, is a common and usually conservative practice in criticality analysis. However, substitution with borated water is typically not conservative. Ensure that the applicant justified any such substitutions.
Confirm that the applicant defined tolerances for poison material dimensions and concentrations and used the most reactive conditions in the criticality analysis. In addition, ensure that the SAR identifies all important design conditions and then addresses these conditions for potential variations during normal, off-normal, and accident conditions.
Verify that the applicant has considered deviations from nominal design configurations. The evaluation of keff should not be limited to a model in which all of the fuel bundles are neatly centered in each basket compartment, with the center line of the basket coincident with the center line of the storage container. For example, a storage container with steel confinement and lead shielding may have a higher keff when the basket and fuel assemblies are positioned as close as possible to the lead. However, in some designs, the most reactive configuration may be when all fuel assemblies are shifted toward the center of the basket.
In addition to a fully flooded storage container, confirm that the SAR addresses configurations in which the container is filled with partial-density water or is partially filled with water (borated, if applicable) and the remainder of the container is filled with steam consisting of ordinary water at partial density. These configurations are considered to be possible during loading and unloading operations. Confirm that the SAR also considers the possibility of preferential or uneven flooding within the storage container, if such a scenario is credible for the container design (e.g., because of blockage in small flow or drain paths). In particular, watch for situations where there is water in the fuel regions but not in the flux traps, if applicable. Storage container designs for which this type of flooding is credible are generally unacceptable. Confirm that the SAR also considers 7-11
flooding in the fuel rod pellet-to-clad gap regions with unborated water. Additionally, for damaged fuel stored in a damaged fuel canister, the tops and bottoms of the damaged fuel canister will typically have screens to allow water drainage during loading. Note that the screened damaged fuel canisters will drain slower than the rest of the storage container, resulting in the potential for preferential flooding in the damaged fuel canisters. This moderation condition could potentially be more reactive than a fully flooded condition. Verify that the applicant has evaluated this condition, if damaged fuel canisters are to be used in the DSS or DSF design and operations. Above all, ensure that the analysis demonstrates that the storage container remains subcritical for all credible conditions of moderation (10 CFR 72.124(a)).
(SL) For other relevant DSF SSCs (e.g., a pool), in addition to the preceding considerations, also ensure that the applicant has identified and addressed any unique aspects of these SSCs that may also impact criticality safety. These aspects may include addressing boron dilution in the DSFs pool.
Material Properties Verify that the SAR provides compositions and densities for all materials used in the calculational models. Ensure that these compositions and densities are consistent with and account for the impacts of normal, off-normal, and accident conditions. Confirm that the SAR, in the chapter on materials evaluation, includes the source of all materials data, particularly the data for fuel and fixed poison materials. In coordination with the materials reviewer, determine the acceptability of the sources of data that are important to the criticality safety function of the storage container.
Also in coordination with the materials reviewer, ensure that the applicant addressed the validation of the fixed neutron absorbers poison concentration in the chapter of the SAR that describes the acceptance tests and maintenance programs. Criticality computer codes generally will allow the densities to be input directly in units of grams per cubic centimeter or units of atoms per barn-centimeter. In either case, pay attention to the final values used directly by the code.
Confirm that the values used for neutron poisons (solid and soluble) match the minimum required values credited in the criticality analysis. Also, for the solid, fixed absorbers, confirm that the analysis does not take credit for more than the minimum amount of neutron absorber verified by the acceptance testing, subject to the criteria in Section 7.4 of this chapter (see also Section 8.5.7 of this SRP).
Among other specifications, 10 CFR Part 72 requires that the applicant provide a positive means to verify the continued efficacy of solid neutron-absorbing materials when these materials are used. Verify that the SAR indicates that the neutron flux from the irradiated fuel results in a negligible depletion of poison material over the storage period. In coordination with the materials and structural reviewers, ensure that the applicant demonstrates that the required acceptance testing of the poisons during fabrication (stated in the chapter of the SAR on acceptance tests and maintenance program evaluation) has been satisfactorily specified and, by analysis or demonstration, that the applicant has shown the poison materials durability and resistance to degradation during the certified or licensed storage period.
The neutron flux used for this analysis should be the maximum that may be produced by feasible loadings of irradiated or unirradiated fuel. Coordinate review of the applicants acceptance testing and assessment of the poison materials durability with the materials reviewer to verify that the applicant provided a valid and accurate demonstration of the absorber materials continued efficacy. Consider the effects of physical and chemical actions as well as irradiation (gamma and neutron). There may be other ways to provide positive means of verifying the neutron absorbers continued efficacy. For applications that propose an alternative method, verify that the proposed 7-12
method is reasonable (considering any effects on meeting confinement, shielding, or other system design criteria), valid, and accurate in demonstrating the absorbers continued efficacy.
(SL) When applying this guidance to absorbers used in a pool that is part of the DSF design and operations in a specific license application, ensure that the SAR appropriately considers the operating environment to which these absorbers will be exposed. Given the configuration for the pool and the operating environment, periodic monitoring of the absorbers may be necessary.
Work with the materials reviewer (SRP Chapter 8) to evaluate the adequacy of the proposed monitoring.
7.5.4 Criticality Analysis Computer Codes and Cross-Section Data Both Monte Carlo and deterministic computer codes may be used for criticality calculations.
Monte Carlo computer codes are better suited to three-dimensional geometry and, therefore, are more widely used to evaluate DSS and DSF storage container designs. The most frequently used Monte Carlo codes are the KENO V.a and KENO VI sequences of the SCALE code system (ORNL 2011) and MCNP (LANL 2003). These codes permit the use of either multi-group or continuous-energy cross sections. Determine whether the applicant has used a computer code that is appropriate for the particular application and has used that code correctly. Ensure that the SAR describes the code the applicant used for its analyses and provides appropriate supplemental information for codes other than those described above to enable this determination. Verify that the information regarding the model configuration, material properties, and cross sections is properly input into the code.
Determine whether the applicant has chosen an acceptable set of cross sections. Cross sections may be distributed with the criticality computer codes or developed independently from another source. Ensure that the applicant provided or referenced the source of cross section data. For user-generated cross sections, verify that the applicant specified the method used to obtain the actual data employed in the criticality analysis. For multi-group calculations, the neutron flux spectrum used to construct the group cross sections should be similar to that of the contents in the storage container. In addition to selecting a cross section set collapsed with an appropriate flux spectrum, a more detailed processing of the multi-group cross sections is necessary to properly account for resonance absorption and self-shielding. The use of multi-group KENO as part of the critical safety analysis sequences in SCALE will directly enable appropriate cross section processing.
More recent versions of Monte Carlo criticality codes can use continuous-energy cross section libraries, which require little or no cross section processing. Use continuous-energy cross sections in the confirmatory analyses, if available, particularly when the applicant has used a multi-group cross section library. This can serve as a check on the cross section processing techniques the applicant employed.
Information has been published concerning problems with some cross section libraries once commonly distributed with SCALE and KENO. One library, the working-format library, was used for calculations of the code manuals sample problems but is not intended for criticality calculations of actual systems (see Information Notice 91-26, Potential Nonconservative Errors in the Working Format Hansen-Roach Cross-Section Set Provided with the KENO and Scale Codes, dated April 2, 1991). Another library, the SCALE 123-group library, is inadequate for non-thermalized, highly enriched systems (see NUREG/CR-6328, Adequacy of the 123-Group 7-13
Cross-Section Library for Criticality Analyses of Water-Moderated Uranium Systems), and may result in non-conservative estimates of keff.
Pay particular attention to the proper selection of scattering cross section data for important compounds that may be in the system. Use of a free atom cross section for nuclides in a compound may not adequately account for the scattering effects of atoms bound in molecules and lattices. This is particularly true for hydrogen bound in water, which is the most common moderator in SNF storage containers. This misrepresentation can cause the under-prediction of keff, particularly in the case of a well-moderated system where energetic up-scattering plays a significant role in the neutronics of the system.
For analyses of a storage container model with separate regions of water and steam, the use of a multi-group cross section set raises additional concerns. Verify that the applicant has addressed the differences in the flux spectra in the two regions. If the results of these calculations indicate that keff is close to 0.95, it may be necessary to conduct additional independent calculations using a different code, cross section library (a library derived from a different cross section database if possible and appropriate), or both, to confirm the applicants calculated keff. Closely examine the applicants benchmark analysis to verify that the critical experiments the applicant considered are applicable to water- and steam-moderated systems. Note that if dissolved boron is credited for criticality control, it will not be present in the steam region.
Neutron Multiplication Factor Examine the results and discussion of the keff calculations for the DSS or DSF. Verify that the calculations determine the highest keff that might occur during all operational states under normal, off-normal, and accident conditions. The applicant may have used sensitivity analyses to provide the required demonstration that the highest keff, with a 95 percent probability at a confidence level of 95 percent, has been determined. Verify that the SAR explains the variations in the results caused by differences in the models and sensitivity analyses and that such variations are reasonable.
For Monte Carlo calculations, assess whether the number of neutron histories and convergence criteria are appropriate. As the number of neutron histories increases, the mean value for keff should approach a fixed value, and the standard deviation associated with each mean value should decrease. Depending on the code the applicant used, a number of diagnostic calculations are generally available to demonstrate adequate convergence and statistical variation. For deterministic codes, a convergence limit is often prescribed in the input. Confirm that the SAR, or supporting criticality calculations, describes and demonstrates the selection of a proper convergence limit and the achievement of this limit. When burnup credit is included in the criticality analysis, confirm that proper neutron sampling and convergence have been achieved because the flux will be concentrated in the low-burnup ends of the fuel assemblies.
Because of the importance and complexity of the criticality evaluation, perform independent calculations to ensure that the applicant has addressed the most reactive conditions, the reported keff is conservative, and the applicant has appropriately modeled the storage container geometry and materials. In deciding the level of effort necessary to perform independent confirmatory calculations, consider the following factors:
x the calculation method (computer code) used by the applicant 7-14
x uniqueness and complexity of the design and analysis, compared with previously approved DSSs and DSF storage containers x the degree of conservatism in the applicants assumptions and analyses x the extent of the margin between the calculated result and the acceptance criterion of keff less than 0.95 As with any design and review, a small margin below the acceptance criterion or a small degree of conservatism (or both) may necessitate a more extensive staff analysis.
Develop a model that is independent of the applicants model. If the reported keff for the most reactive case is substantially lower than the acceptance criterion of 0.95, a simple model(s) known to produce very bounding results may be all that is necessary for the independent calculations.
If possible and appropriate, perform the independent calculations with a computer code different from the code the applicant used. Likewise, use of a different cross-section set, derived from a different cross section database, where possible and appropriate (e.g., ENDF/B, JEF, JENDL, UKNDL), can provide a more independent confirmation. The continuous-energy cross sections created for use with KENO in the SCALE code system are generated by the AMPX processing code rather than the more widely used NJOY code. Even though some cross section libraries may not have fully independent databases because they are all derived from ENDF/B data, the continuous-energy library in SCALE still can provide some level of independence and is useful for checking computations performed with libraries that were generated by using NJOY. Describe the staffs independent analysis, the analysiss general results, and the staffs conclusions in the SER.
Although a keff of 0.95 or lower meets the acceptance criterion, watch for design features or content specifications where small changes could result in large changes in the value of keff.
When the value of keff is highly sensitive to system parameters that could vary, the acceptable keff limit may need to be reduced to below 0.95. When establishing a keff limit below 0.95, consider the degree of sensitivity to system parameter changes and the likelihood and extent of potential parameter variations.
Benchmark Comparisons Computer codes for criticality calculations should be benchmarked against critical experiments. A thorough comparison provides justification for the validity of the computer code, its use for a specific hardware configuration, its use for the SNF to be stored, the neutron cross sections used in the analysis, and consistency in modeling by the analyst. Ultimately, the benchmarking process establishes a bias and bias uncertainty for the particular application of the code (using the benchmark results for calculations performed by another analyst does not address this last issue).
Calculated keff values should then be adjusted to include the appropriate biases and bias uncertainties from the benchmark calculations.
Examine the general description of the benchmark comparisons. This examination includes verifying that the analysis of the experiments used the same computer code, computer system, cross section data, modeling methods, and code options that were used to calculate the keff values for the storage containers.
7-15
Closely examine the applicants benchmark analysis to determine whether the benchmark experiments are relevant to the actual storage container design. No critical benchmark experiment will precisely match the fissile material, moderation, neutron poisoning, and geometric configuration in the actual storage container. However, the applicant can perform a proper benchmark analysis by selecting experiments that adequately represent storage container and fuel features and parameters that are important to reactivity. Key features and parameters that should be considered in selecting appropriate critical experiments include the type of fuel, enrichment, hydrogen-to-uranium ratio (dependent largely on rod diameter and pitch), reflector material, neutron energy spectrum, and poisoning material and placement. Confirm that the applicant discusses and properly considers the differences between the benchmark experiments and the storage containers and their contents. Ensure that the SAR addresses the overall quality of the benchmark experiments and the uncertainties in the experimental data (e.g., mass, density, dimensions). Verify that the applicant treated these uncertainties in a conservative manner (i.e., used in a way that results in a lower calculated keff for the benchmark experiment).
Verify the applicants justification of the suitability of the critical experiments chosen to benchmark the criticality code and calculations. Techniques such as the sensitivity and uncertainty method developed by Oak Ridge National Laboratory (ORNL 2011) can be helpful when assessing the applicability of the critical experiments used to benchmark the design analysis. UCID-21830, Determination and Application of Bias Values in the Criticality Evaluation of Storage Cask Designs, issued January 1990; the Nuclear Energy Agencys International Handbook of Evaluated Criticality Safety Benchmark Experiments; and NUREG/CR-6361, Criticality Benchmark Guide for Light-Water-Reactor Fuel in Transportation and Storage Packages, issued March 1997, provide information on benchmark experiments that may apply to the storage containers being analyzed.
Assess whether the applicant analyzed a sufficient number of appropriate benchmark experiments (which is dependent, in part, on the statistical treatment used) and how the applicant converted the results of these benchmark calculations to a bias for the keff calculations. Simply averaging the biases from a number of benchmark calculations typically is not sufficient, such as when one benchmark yields results that are significantly different from the others, the number of experiments is limited, or when groups of experiments are heavily correlated. In addition, the bias may exhibit trends with respect to parameter variations (such as pitch-to-rod-diameter ratio, assembly separation, reflector material, neutron absorber material). Verify that the applicant has adequately assessed the benchmark analysis results to identify any bias trends and considered these trends in developing a bias for the keff calculations. UCID-21830 and NUREG/CR-6361 provide some guidance; however, other methods, when adequately explained, have also been considered appropriate.
For Monte Carlo codes, ensure that the applicant also addresses the statistical uncertainties of both the benchmark and the keff calculations. The uncertainties should be applied to at least the 95 percent confidence level. As a general rule, if the acceptability of the result depends on these rather small differences, question the overall degree of conservatism of the calculations.
Considering the current availability of computer resources, a sufficient number of neutron histories can readily be used so that the treatment of these uncertainties should not significantly affect the results.
Verify that the applicant has applied only biases that increase the calculated keff. If the benchmark analysis results in a positive bias (i.e., one that would decrease the calculated keff), the bias should 7-16
be conservatively set to zero. Only corrections that increase keff should be applied to preserve conservatism.
The reviewer may have already performed a number of benchmark calculations applicable to storage containers and may have a reasonable estimation of the bias to be applied to the independent calculation of the keff for the storage containers. If such is not the case, or if the acceptability depends on small bias differences, determine whether sufficient conservatism has been applied to the calculations.
7.5.5 Burnup Credit The regulations in 10 CFR Part 72 require that SNF remain subcritical in storage. While unirradiated reactor fuel (or fresh fuel) has a well-specified nuclide composition that provides a straightforward and bounding approach to the criticality safety analysis of transportation and storage systems, the nuclide composition changes as the fuel is irradiated in the reactor. Ignoring the presence of burnable poisons, this composition change will cause the reactivity of the fuel to decrease. In the criticality safety analysis, allowance for the decrease in fuel reactivity resulting from irradiation is termed burnup credit.
This section provides recommendations to the NRC reviewer for accepting, on a design-specific basis, a burnup credit approach in the criticality safety analysis of PWR SNF storage containers.
The recommendations are based on DSS-type storage container designs; however, they may also be applied to other SNF storage container designs, with appropriate consideration of the differences between container designs. For specific license applications, the recommendations may also be applied to criticality analyses for SNF in other relevant DSF SSCs (e.g., a pool that is part of the DSF design and operations), with appropriate consideration of impacts of these SSCs features on the bases for and the application of the recommendations. The guidance represents one methodology for demonstrating compliance with the criticality safety requirement in 10 CFR Part 72 using burnup credit. Follow this guidance to determine whether the applicant has adequately demonstrated that the storage system meets the applicable criticality safety regulations in 10 CFR Part 72. Consider proposed alternative methodologies on a case-by-case basis, using this guidance to the extent practicable.
The recommendations that follow were developed with intact fuel as the basis but may also be applicable to fuel that is not intact. If an applicant requests burnup credit for fuel that is not intact, apply the recommendations provided below, as appropriate, to account for uncertainties that can be associated with fuel that is not intact and establish an isotopic inventory and assumed fuel configuration for normal, off-normal, and accident conditions that bound the uncertainties.
The recommendations in this chapter do not include burnup credit for BWR fuel assemblies, as the technical basis for BWR burnup credit in SNF storage containers has not been fully developed. The NRC has initiated a research project to obtain that technical basis. BWR fuel assemblies typically have neutron-absorbing material, typically gadolinium oxide, mixed in with the uranium oxide of the fuel pellets in some rods. This neutron absorber depletes more rapidly than the fuel during the initial parts of its irradiation, which causes the fuel assembly reactivity to increase and reach a maximum value at an assembly average burnup typically less than 20 GWd/MTU. Then reactivity decreases for the remainder of fuel assembly irradiation. Criticality analyses of BWR spent fuel pools typically employ what are known as peak reactivity methods to account for this behavior. NUREG/CR-7194, Technical Basis for Peak Reactivity Burnup Credit for BWR Spent Nuclear Fuel in Storage and Transportation Systems, reviews several existing peak reactivity methods, and demonstrates that a conservative set of analysis conditions 7-17
can be identified and implemented to allow criticality safety analysis of BWR spent fuel assemblies at peak reactivity in SNF storage containers. Consult NUREG/CR-7194 if the applicant used peak reactivity BWR burnup credit methods in its criticality analysis.
This SRP does not address credit for BWR burnup beyond peak reactivity; the NRC is currently evaluating this as part of a research program to investigate methods for conservatively including such credit in a BWR criticality analysis for SNF storage containers. The NRC does not recommend burnup credit beyond peak reactivity at this time. Consider conservative analyses of BWR burnup credit beyond peak reactivity on a case-by-case basis, consulting the latest research results in this area (i.e., NRC letter reports, NUREG/CRs).
The recommendations in this section also do not include burnup credit analyses for MOX or thorium fuel assemblies. Evaluate MOX burnup credit analyses on a case-by-case basis, noting that there is little MOX data available for isotopic depletion or criticality code validation. Such evaluations should include a large amount of conservatism in the representation of MOX material in the criticality model, and large keff penalties for unvalidated fuel materials. Thorium fuel criticality analyses will require a depletion analysis to determine the most reactive fuel composition with irradiation. Similar to MOX fuel, there is little code validation data available for thorium fuel, and criticality analyses should include large conservatisms and keff penalties for unvalidated materials.
Appendix 7A to this SRP chapter provides more information on the technical bases for the recommendations provided below.
Limits for the Licensing Basis Available data support allowance for burnup credit where the safety analysis is based on major actinide compositions only (i.e., actinide-only burnup credit) or limited actinide and fission product compositions (see Table 7-2 below) associated with UO2 fuel irradiated in a PWR up to an assembly-average burnup value of 60 GWd/MTU and cooled out of reactor for a period between 1 and 40 years. The range of available measured assay data for irradiated UO2 fuel supports an extension of the licensing basis up to 5.0 weight percent enrichment in uranium-235.
Table 7-2 Recommended Set of Nuclides for Burnup Credit Type of Burnup Credit Recommended Set of Nuclides 234 235 238 238 239 240 241 242 241 Actinide-only burnup credit U, U, U, Pu, Pu, Pu, Pu, Pu, Am Additional nuclides for 95Mo, 99Tc, 101Ru, 103Rh, 109Ag, 133Cs, 143Nd, 145Nd, 147Sm, 149Sm, actinide-plus-fission product 150Sm, 151Sm, 152Sm, 151Eu, 153Eu, 155Gd, 236U, 237Np, 243Am burnup credit Within this range of parameters, exercise care in assessing whether the analytic methods and assumptions used are appropriate, especially near the limits of the parameter ranges recommended here for the licensing basis. Verify that the use of actinide and fission product compositions associated with burnup values or cooling times outside these specifications is accompanied by the measurement data or justifies extrapolation techniques, or both, necessary to extend the isotopic validation and quantify or bound the bias and bias uncertainty. If the applicant credits neutron-absorbing isotopes other than those identified in Table 7-2, ensure that the applicant gives assurance that such isotopes are nonvolatile, nongaseous, and relatively stable, and provides analyses to determine the additional depletion and criticality code bias and bias uncertainty associated with these isotopes.
7-18
A certificate or license condition indicating the time limit on the validity of the burnup credit analysis may be necessary in light of the potential need for extended dry storage. Such a condition would depend on the type of burnup credit and the credited post-irradiation decay time.
Licensing-Basis Model Assumptions Confirm that the applicant calculated the actinide and fission product compositions used to determine a value of keff for the licensing basis using fuel design and reactor operating parameter values that appropriately encompass the range of design and operating conditions for the proposed contents. Verify that the applicant performed the calculation of the keff value using models and analysis assumptions that allow accurate representation of the physics in the storage container, as discussed in Section 7A.4 of Appendix 7A to this chapter. Pay attention to the need to do the following:
x Account for and effectively model the axial and horizontal variation of the burnup within a spent fuel assembly (e.g., the selection of the axial burnup profiles, number of axial material zones).
x Consider the potential for increased reactivity because of the presence of burnable absorbers or control rods (fully or partially inserted) during irradiation.
x Account for the irradiation environment factors to which the proposed assembly contents were exposed, including fuel temperature, moderator temperature and density, soluble boron concentration, specific power, and operating history.
YAEC-1937, Axial Burnup Profile Database for Pressurized Water Reactors, issued May 1997, provides a source of representative data that can be used for establishing profiles to use in the licensing-basis safety analysis. However, exercise care when reviewing profiles intended to bound the range of potential keff values for the proposed contents for each burnup range, particularly near the upper end of the licensing basis parameter ranges stated in this guidance.
NUREG/CR-6801, Recommendations for Addressing Axial Burnup in PWR Burnup Credit Analyses, issued March 2003, provides additional guidance on selecting axial profiles.
A licensing-basis modeling assumption, where the assemblies are exposed during irradiation to the maximum (neutron absorber) loading of BPRs for the maximum burnup, encompasses all assemblies that may or may not have been exposed to BPRs. Such an assumption in the licensing-basis safety analysis should also encompass the impact of exposure to fully inserted or partially inserted control rods in typical domestic PWR operations. Assemblies exposed to atypical insertions of control rods (e.g., full insertion for one full cycle of reactor operation) should not be loaded unless the safety analysis explicitly considers such operational conditions. If the assumption on BPR exposure is less than the maximum for which burnup credit is requested, confirm that the applicant has provided a justification commensurate with the selected value. For example, the lower the exposure, the greater the need to (1) support the assumption with available data, (2) indicate how administrative controls would prevent a misload of an assembly exposed beyond the assumed value, and (3) address such misloads in a misload analysis.
For assemblies exposed to integral burnable absorbers, the appropriate analysis assumption for absorber exposure varies depending upon burnup and absorber material. The appropriate assumption may be to neglect the absorber while maintaining the other assembly parameters (e.g., enrichment) the same for some absorber materials or for exposures up to moderate burnup levels (typically 20-30 GWd/MTU). Thus, a safety analysis including assemblies with integral 7-19
burnable absorbers should include justification of the absorber exposure assumptions used in the analysis. For assemblies exposed to flux suppressors (e.g., hafnium suppressor inserts) or combinations of integral absorbers and BPRs or control rods, the safety analysis should use assumptions that provide a bounding safety basis, in terms of the effect on storage container keff, for those assemblies.
Confirm that the applicants licensing-basis evaluation includes analyses that use irradiation conditions that produce bounding values for keff, as discussed in Section 7A.4 of Appendix 7A to this chapter. The bounding conditions may differ for actinide-only burnup credit versus actinide-plus-fission product burnup credit and may depend on the population of fuel intended to be loaded in the storage container (e.g., all PWR assemblies versus a site-specific population).
Loading limitations tied to the actual operating conditions may be needed unless the operating condition values used in the licensing-basis evaluation can be justified as those that produce the maximum keff values for the anticipated SNF inventory.
Code ValidationIsotopic Depletion Confirm that the applicant validated the computer codes used to calculate isotopic depletion.
A depletion computer code is used to determine the concentrations of the isotopes important to burnup credit. To ensure accurate criticality calculation results, the selected code should be validated and the bias and bias uncertainty of the code should be determined at a 95-percent probability, 95-percent confidence level. Ensure that the application reflects the following considerations in the selection of the code and code validation approach for the fuel depletion analysis.
The selected depletion code and cross section library should be capable of accurately modeling the fuel geometry and the neutronic characteristics of the environment in which the fuel was irradiated. Two-dimensional depletion codes have been effectively used in burnup credit analyses. Although one-dimensional codes have been used in some applications and suffice for making assembly average isotopic predictions for fuel burnup, they are limited in their ability to model increasingly complex fuel assembly designs and generally produce larger bias and bias uncertainty values because of the approximations necessary in the models. Section 7A.4 of Appendix 7A to this chapter provides detailed discussions of the modeling considerations for the code validation analyses.
The destructive radiochemical assay (RCA) data selected for code validation should include detailed information about the SNF samples. This information should include the pin location in the assembly, axial location of the sample in the pin, any exposure to strong absorbers (control rods, BPRs), the boron letdown, moderator temperature, specific power, and any other cycle-specific data for the cycles in which the sample was irradiated. Note that some RCA data are not suitable for depletion code validation because the depletion histories or environments of these samples are either difficult to accurately define in the code benchmark models or are unknown. NUREG/CR-7108, An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety AnalysesIsotopic Composition Predictions, issued April 2012, provides a recommended set of RCA data suitable for depletion code validation.
The selected code validation approach should be adequate for determining the bias and bias uncertainty of the code for the specific application. The burnup credit analysis results should be adjusted using the bias and bias uncertainty determined for the fuel depletion code, accounting for any trends of significance with respect to different control parameters such as burnup-to-enrichment ratio or ratio of uranium-235 to plutonium-239. NUREG/CR-6811, Strategies for 7-20
Application of Isotopic Uncertainties in Burnup Credit, issued June 2003, provides several methodologies the NRC finds acceptable for isotopic depletion validation, including the isotopic correction factor, direct-difference, and Monte Carlo uncertainty sampling methods. Section 7A.4 of Appendix 7A to this chapter provides detailed discussions of the advantages and disadvantages of these methods. In general, the isotopic correction factor method is considered to be the most conservative because individual nuclide composition uncertainties are represented as worst case. The direct-difference method provides a realistic best estimate of the depletion code bias and bias uncertainty, in terms of difference in keff (Neff). The Monte Carlo uncertainty sampling method is more complex and computationally intensive than the other methods, but it provides a way to make use of limited measurement data sets for some nuclides.
NUREG/CR-7108 provides detailed descriptions of the direct-difference and Monte Carlo uncertainty sampling methods.
In lieu of an explicit benchmarking analysis, the applicant may use the bias (i) and bias uncertainty (Ni) values estimated in NUREG/CR-7108 using the Monte Carlo uncertainty sampling method, as shown in Tables 7-3 and 7-4. These values may be used directly, provided that all of the following is true:
x The applicant uses the same depletion code and cross section library as were used in NUREG/CR-7108 (SCALE/TRITON and the ENDF/B-V or ENDF/B-VII cross section library).
x The applicant can justify that its storage container design is similar to the hypothetical 32-PWR-assembly-capacity, generic burnup credit cask (GBC-32) system design (NUREG/CR-6747, Computational Benchmark for Estimation of Reactivity Margin from Fission Products and Minor Actinides in PWR Burnup Credit, issued October 2001) and used as the basis for the NUREG/CR-7108 isotopic depletion validation.
x Credit is limited to the specific nuclides listed in Table 7-2.
Section 7A.5 of Appendix 7A to this chapter provides detailed discussions of the technical basis for the restrictions on directly applying the bias and bias uncertainty values. Bias values should be added to the calculated storage container keff, while bias uncertainty values may be statistically combined with other independent uncertainties. Table 7-5 summarizes the recommendations related to isotopic depletion code validation.
Table 7-3 Isotopic keff %LDV8QFHUWDLQW\ Ni) for the Representative PWR SNF System Model 8VLQJ(1')%9,,'DWD i = 0) as a Function of Assembly Average Burnup Actinides and Burnup (BU) Range Actinides Only Fission Products (GWd/MTU) ki ki 0%8 5 0.0145 0.0150 5%8 10 0.0143 0.0148 10%8 18 0.0150 0.0157 18%8 25 0.0150 0.0154 25%8 30 0.0154 0.0161 30%8 40 0.0170 0.0163 40%8 45 0.0192 0.0205 45%8 50 0.0192 0.0219 50%860 0.0260 0.0300 7-21
Table 7-4 Isotopic keff %LDV i DQG%LDV8QFHUWDLQW\ Ni) for the Representative PWR SNF System Model Using ENDF/B-V Data as a Function of Assembly Average Burnup Burnup (BU) Range i for Actinides and ki for Actinides and Fission (GWd/MTU)a Fission Products Products 0%8 10 0.0001 0.0135 10%8 25 0.0029 0.0139 25%840 0.0040 0.0165
- a. Bias and bias uncertainties associated with ENDF/B-V data were calculated for a maximum of 40 GWd/MTU. For burnups higher than this, confirm that the applicant provided an explicit depletion code validation analysis, using one of the methods described in Appendix 7A to this chapter, along with appropriate RCA data.
Table 7-5 Summary of Code Validation Recommendations for Isotopic Depletion Applicants Approach Recommendation Applicant uses SCALE/TRITON and the ENDF/B-V Use code bias and bias uncertainty values from or ENDF/B-VII cross section library, and Tables 7-3 and 7-4 of this SRP.
demonstrates that the design application is similar to GBC-32.
- or -
Applicant uses other code or cross section library, Use either isotopic correction factor or or both, or design application is not similar to direct-difference method to determine code bias GBC-32. and bias uncertainty.
Code Validationkeff Determination Actinide-Only Credit Actinide credit should be limited to the specific nuclides listed in Table 7-2. Criticality validation for these actinides should be based on the critical experiments available in NUREG/CR-6979, Evaluation of the French Haut Taux de Combustion (HTC) Critical Experiment Data, issued September 2008, also known as the HTC data, supplemented by MOX critical experiments as appropriate. NUREG/CR-7109, An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety AnalysesCriticality (keff) Predictions, issued April 2012, contains a detailed discussion of available sets of criticality validation data for actinide isotopes, and the relative acceptability of these sets. Note that NUREG/CR-7109 demonstrates that fresh UO2 experiments are not applicable to burned fuel compositions.
Verify that the applicants determination of the bias and bias uncertainty associated with actinide-only burnup credit was performed according to the guidance in NUREG/CR-6361. This guidance includes criteria for the selection of appropriate benchmark data sets, as well as statistics and trending analysis for the determination of criticality code bias and bias uncertainty.
Section 6 of NUREG/CR-7109 provides an example of bias and bias uncertainty determination for actinide-only burnup credit.
Fission Product and Minor Actinide Credit Confirm that the applicant has determined an adequate and conservative bias and bias uncertainty associated with fission product and minor actinide credit. The applicant may credit the 7-22
minor actinide and fission product nuclides listed in Table 7-2, provided the bias and bias uncertainty associated with the major actinides is determined as described above. The bias from these minor actinides and fission products is conservatively covered by 1.5 percent of their worth.
Because of the conservatism in this value, no additional uncertainty in the bias needs to be applied. This estimate is appropriate if the applicant does the following:
x uses the SCALE code system with the ENDF/B-V, ENDF/B-VI, or ENDF/B-VII cross section libraries, or MCNP5 or MCNP6 with the ENDF/B-V, ENDF/B-VI, ENDF/B-VII, or ENDF/B-VII.1 cross section libraries.
x can justify that its storage container design is similar to the hypothetical GBC-32 system design (NUREG/CR-6747) used as the basis for the NUREG/CR-7109 criticality validation x demonstrates that the credited minor actinide and fission product worth is no greater than 0.1 in keff For well-qualified industry standard code systems other than SCALE or MCNP, the applicant may use a conservative estimate for the bias associated with minor actinide and fission product nuclides of 3.0 percent of their worth. If the applicant uses a minor actinide and fission product bias less than 3.0 percent, ensure that the application includes additional justification that the lower value is an appropriate estimate of the bias associated with that code system (e.g., a minor actinide and fission product worth comparison to SCALE results or an analysis similar to that described in NUREG/CR-7109 or NUREG/CR-7205, Bias Estimates Used in Lieu of Validation of Fission Products and Minor Actinides in MCNP Keff Calculations for PWR Burnup Credit Casks).
Table 7-6 summarizes the recommendations related to minor actinide and fission product code validation for keff determination. For actinide criticality validation in all cases, the applicant should perform criticality code validation analyses to determine bias and bias uncertainty associated with actinides using HTC critical experiments, supplemented by applicable MOX critical experiments.
Ensure that the applicant performed the validation analyses correctly and adequately.
Table 7-6 Summary of Minor Actinide and Fission Product Code Validation Recommendations for keff Determination Applicants Approach Recommendation Applicant uses SCALE code system with ENDF/B-V, Use bias equal to 1.5 percent of minor ENDF/B-VI, or ENDF/B-VII cross section libraries, or MCNP5 actinide and fission product worth.
or MCNP6 with the ENDF/B-V, ENDF/B-VI, ENDF/B-VII, or ENDF/B-VII.1 cross section libraries; design application is similar to GBC-32; and credited minor actinide and fission SURGXFWLVZRUWK LQNeff.
- or -
Applicant uses other code with ENDF/B-V, ENDF/B-VI, or Use bias equal to 3.0 percent of minor ENDF/B-VII cross section libraries; design application is similar actinide and fission product worth, or to GBC-32; and credited minor actinide and fission product is provide justification for lower number.
ZRUWK LQNeff.
- or -
Applicant uses cross section library other than ENDF/B-V, Perform explicit criticality code ENDF/B-VI, or ENDF/B-VII; design application is not similar to validation for minor actinide and fission GBC-32; or credited minor actinide and fission product is worth product nuclides.
>0.1 in keff.
7-23
Loading Curve and Burnup Verification Confirm that the applicant provided burnup credit loading curves to determine which fuel assemblies may be loaded in a storage container. Confirm that the burnup credit evaluations include loading curves that specify the minimum required assembly average burnup as a function of initial enrichment for the purpose of loading SNF storage containers. Confirm that separate loading curves are established for each set of applicable licensing conditions. For example, a separate loading curve should be provided for each minimum cooling time to be considered in the container loading. In addition, confirm that the SAR includes a justification of the applicability of the loading curve to bound various fuel types or burnable absorber loadings.
Ensure that the criticality analysis and operations description chapters in the SAR include performance of burnup verification to ensure that a storage container evaluated using burnup credit is not loaded with an assembly more reactive than those included in the loading criteria.
Verification should include a measurement that confirms the reactor record for each assembly.
Confirmation of reactor records using measurement of a sample of fuel assemblies will be considered if the sampling method can be justified in comparison to measuring every assembly.
The assembly burnup value to be used for loading acceptance (termed the assigned burnup loading value) should be the confirmed reactor record value as adjusted by reducing the record value by a combination of the uncertainties in the record value and the measurement.
NUREG/CR-6998, Review of Information for Spent Nuclear Fuel Burnup Confirmation, issued December 2009, contains bounding estimates of reactor record burnup uncertainty.
Measurements should be correlated to reactor record burnup, enrichment, and cooling time values. Measurement techniques should account for any measurement uncertainty (typically within a 95-percent confidence interval) in confirming reactor burnup records. They should also include a database of measured data (if measuring a sampling of fuel assemblies) to justify the adequacy of the procedure in comparison to procedures that measure each assembly.
Misload Analyses Misload analyses may be performed in lieu of a burnup measurement. A misload analysis should address potential events involving the placement of assemblies into a SNF storage container that do not meet the proposed loading criteria. Confirm that the applicant has demonstrated that the container remains subcritical for misload conditions, including calculation biases, uncertainties, and an appropriate administrative margin that is not less than 0.02 N. If any administrative margin less than the normal 0.05 k is used, verify that the SAR provides an adequate justification that includes the level of conservatism in the depletion and criticality calculations, sensitivity of the container to further upset conditions, and the level of rigor in the code validation methods.
If used, ensure that the misload analysis considers (1) misloading of a single, severely underburned assembly and (2) misloading of multiple, moderately underburned assemblies.
The severely underburned assembly for the single misload analysis should be chosen such that the misloaded assemblys reactivity bounds 95 percent of the discharged PWR fuel population considered unacceptable for loading in a particular storage container with 95-percent confidence.
The moderately underburned assemblies for the multiple-misload analysis should be assumed to make up at least 50 percent of the container payload and should be chosen such that the misloaded assemblies reactivity bounds 90 percent of the total discharged PWR fuel population.
7-24
The NRC finds the results of the most recent Energy Information Administration nuclear fuel data survey, RW-859, Nuclear Fuel Data Survey, or later similar fuel data sources, acceptable to estimate the discharged fuel population characteristics.
Also ensure that the misload analysis considers the effects of placing the underburned assemblies in the most reactive positions within the loaded container (e.g., middle of the fuel basket). If removable nonfuel absorbers were credited as part of a criticality safety analysis (e.g., poison rods added to guide tubes), ensure that the misload analysis considers misloading of these absorbers. Additionally, ensure that the misload analysis considers assemblies with greater burnable absorber or control rod exposure than assumed in the criticality analysis if the assumed exposure is not bounding. NUREG/CR-6955, Criticality Analysis of Assembly Misload in a PWR Burnup Credit Cask, issued January 2008, illustrates the magnitude of keff changes that can be expected as a result of various misloads in a theoretical GBC-32 SNF storage system.
Administrative Procedures Confirm that the applicant has included administrative loading procedures that will protect against misloads. Ensure that the misload analysis is coupled with additional administrative procedures to ensure that the SNF storage container will be loaded with fuel that is within the specifications of the approved contents. Procedures the applicant may consider to protect against misloads in storage containers that rely on burnup credit for criticality safety include the following:
x verification of the location of high-reactivity fuel (i.e., fresh or severely underburned fuel) in the spent fuel pool, both before and after loading x qualitative verification that the assembly to be loaded is burned (visual or gross measurement) x quantitative measurement of any fuel assemblies without visible identification numbers x independent, third-party verification of the loading process, including the fuel selection process and generation of the fuel move instructions x minimum soluble boron concentration in pool water, to offset the misloads described above, during loading and unloading Table 7-7 summarizes the recommendations for burnup verification.
Table 7-7 Summary of Burnup Verification Recommendations Applicants Approach Recommendation Applicant takes burnup verification Perform measurement for each assembly to be loaded or for measurement. a statistically significant sample of assemblies.
- or -
Applicant conducts misload analysis and Analyze misload of fuel assembly that bounds reactivity of provides additional administrative 95 percent of underburned fuel population with 95-percent procedures. confidence.
Analyze misload of 50 percent of system capacity with fuel assemblies with reactivity that bounds 90 percent of total fuel population.
Include additional administrative procedures as part of storage container loading.
7-25
7.5.6 Reactor-Related Greater-Than-Class-C Waste and HLW (SL)
(SL) The specifications for materials stored at a DSF should include the ranges of properties of concern for criticality analysis, which may include HLW and reactor-related GTCC waste characteristics if these wastes are to be stored at the DSF and they contain fissile material.
Chapter 3, Principal Design Criteria Evaluation, of this SRP provides guidance on the data that should be provided.
(SL) For these wastes, characteristics of concern for the various criticality analyses include those listed below. Verify that the SAR states these characteristics for the radioactive materials in these wastes for which criticality analysis is appropriate. Ensure that the SAR identifies radioactive materials that, because of their atomic or physical properties, are not of criticality concern, and includes this as the rationale for not including criticality analyses. Verify that the applicant has provided the data identified below, regardless of whether or not they are included in the applicants analytical approach, as they may be needed for confirmatory and independent analyses by the NRC staff:
x the isotopes present and their densities x means by which the fissile and fissionable isotope densities are limited x geometric data on the configuration (e.g., racks, basket) holding the materials, including tolerances and uncertainties, and neutron-absorption material integral to the configuration x characteristics (materials, densities, geometries, tolerances, uncertainties) of any encapsulation used to provide confinement and structural support during handling and when within the storage container (SL) Verify that the applicant has demonstrated that HLW and reactor-related GTCC wastes containing fissile material will remain subcritical. In general, reactor-related GTCC waste containers are not expected to contain significant amounts of fissile material. The most likely types of reactor-related GTCC waste that may contain fissile material are fission chambers, some neutron sources, filters, and ion-exchange resins. Verify that the applicant has addressed these potential sources of fissile material (if present) and has demonstrated that their quantity is insignificant. For those HLW and reactor-related GTCC waste forms for which criticality is a concern, verify that the applicant has demonstrated that the most reactive configurations of the wastes have been analyzed and that their keff values remain below 0.95. Also verify that the analysis includes adequate benchmarking, consistent with the guidance in Section 7.5.4.3 of this SRP but appropriately applied for these wastes. In general, for reactor-related GTCC wastes, it is not necessary to perform independent confirmatory analyses.
(SL) Also verify that the applicant has demonstrated that storage of GTCC waste will not adversely affect the safe storage of SNF and HLW at the DSF. In general, containers of GTCC waste located with SNF and HLW storage containers at an ISFSI or MRS are not expected to increase the reactivity of the SNF and HLW storage containers.
7.5.7 Supplemental Information Ensure that the SAR includes all supportive information or documentation. This may include, but not be limited to, justification of assumptions or analytical procedures, test results, photographs, 7-26
computer program descriptions, input/output, and applicable pages from referenced documents.
In addition, confirm that the SAR includes a list of fuel designs with the acceptable parametric limits and the maximum enrichments for which the criticality analysis is valid. Request any additional information needed to complete the review.
Evaluation Findings The NRC reviewer should prepare evaluation findings upon satisfaction of the applicable regulatory requirements in Section 7.4 of this SRP. If the documentation submitted with the application fully supports positive findings for each of the regulatory requirements, the statements of findings should be similar to the following:
F7.1 The applicant has described the SSCs important to criticality safety in sufficient detail in Chapters ______ of the SAR to enable an evaluation of their effectiveness in accordance with [for SL use: 10 CFR 72.24(b) and 10 CFR 72.24(c); for CoC use: 10 CFR 72.236(b)].
F7.2 (CoC) The applicant has designed the ____ DSS, including its transfer cask for canister-based systems, to be subcritical under all credible conditions in accordance with 10 CFR 72.124(a) and 10 CFR 72.236(c).
(SL) The applicant has designed the _______ DSFs SSCs involved in the loading, unloading, packaging, handling, transfer, and storage of the SNF at the DSF to be subcritical under all credible conditions in accordance with 10 CFR 72.124(a).
F7.3 The applicant based the criticality design of the [DSS or DSF SSCs] on favorable geometry, fixed neutron poisons, and soluble poisons 1 [as applicable]. The applicants evaluation of the fixed neutron poisons in the storage container has shown that the fixed neutron poisons will remain effective for the storage term requested in the [CoC or specific license]
application and there is no credible way for the fixed neutron poisons to significantly degrade during the requested storage term in the [CoC or specific license] application. Therefore, there is no need to provide a positive means to verify their continued efficacy as required in 10 CFR 72.124(b).
[For specific license applications for a DSF, the design and operations of which include a pool or other SSCs that use fixed neutron poisons, use the following finding for the applicants evaluation of these fixed poisons:
The applicant has provided an adequate means to verify, during the licensed storage term, the continued efficacy of the fixed neutron poisons in the [list applicable DSF SSCs] as required in 10 CFR 72.124(b).]
F7.4 The applicants analysis and evaluation of the criticality design and performance of the [DSS or DSF SSCs] have demonstrated that the [DSS 1 Soluble poisons may be relied upon for wet loading or unloading. For DSSs and for DSFs that are co-located at 10 CFR Part 50- or 10 CFR Part 52-licensed reactor facilities and share the pool, this would be soluble poisons in the 10 CFR Part 50- or 10 CFR Part 52-licensed facilitys SNF pool. For DSF designs and operations that include a pool as part of the specific license application, this would be soluble poisons in the DSFs pool.
7-27
or DSF] will enable the storage of SNF for the term requested in the [CoC or specific license] application (for SL: 10 CFR 72.24(c); for CoC: 10 CFR 72.236(g)).
F7.5 (SL) The design and operations of the proposed DSF and the characteristics of the materials to be stored at the proposed DSF provide reasonable assurance that the activities authorized by the specific license can be conducted without endangering the health and safety of the public, in compliance with 10 CFR 72.40(a)(13). This includes the use of necessary criticality monitoring systems as required in 10 CFR 72.124(c),
and the necessary design and operations parameters to ensure HLW or reactor-related GTCC waste to be stored at the DSF and that contains fissile materials remains subcritical under all credible conditions.
F7.6 The design and proposed [use of the DSS/operations of the DSF],
including SSCs involved in the handling, packaging, transfer, and storage of the radioactive materials to be stored, acceptably ensure that the materials will remain subcritical and that, before a nuclear criticality accident is possible, at least two unlikely, independent, and concurrent or sequential changes must occur in the conditions essential to nuclear criticality safety. The applicants analyses in the SAR and confirmatory analysis by the NRC adequately show that acceptable margins of safety will be maintained in the nuclear criticality parameters commensurate with uncertainties in the data and methods used in calculations, and demonstrate safety for the handling, packaging, transfer, and storage conditions and in the nature of the immediate environment under accident conditions in compliance with 10 CFR 72.124(a) [and (for a CoC) 10 CFR 72.236(c)].
F7.7 The proposed [CoC or license] conditions, including the technical specifications, include those items necessary to ensure nuclear criticality safety in the design, fabrication, construction, and operation of the [DSS or DSF] SSCs [(for CoC) consistent with what is considered necessary to ensure compliance with 10 CFR 72.236(a), 72.236(b), and 72.236(c); (for SL) in accordance with the requirements in 10 CFR 72.24(g) and 10 CFR 72.44(c)].
F7.8 The SAR provides specifications of the [(for CoC) spent fuel contents to be stored in the [DSS designation]; (for SL) the materials to be stored at the [DSF designation)) in sufficient detail that adequately defines the allowed [contents/materials] and allows evaluation of the [DSS or DSF designation] nuclear criticality safety design for the proposed
[contents/materials]. The SAR includes analyses that are adequately bounding for the proposed [contents/materials] specifications. (CoC: 10 CFR 72.236(a); SL: 10 CFR 72.24(c))
7-28
F7.9 (CoC) The applicant has designed the ____ DSS, including its transfer cask for canister-based systems, for criticality safety purposes, to be compatible with wet and dry loading and unloading facilities and, to the extent practicable, removal of the stored spent fuel from the site and transportation in accordance with 10 CFR 72.236(h) and 10 CFR 72.236(m).
The reviewer should provide a summary statement similar to the following:
The staff concludes that the criticality design features for the [DSS or DSF designation]
are in compliance with 10 CFR Part 72, as exempted [if applicable], and that the applicable design and acceptance criteria have been satisfied. The evaluation of the criticality design provides reasonable assurance that the [DSS or DSF designation] will allow safe storage of SNF [and HLW and reactor-related GTCC waste, as applicable for the DSF]. This finding is reached on the basis of a review that considered the regulation itself, appropriate regulatory guides, applicable codes and standards, and accepted engineering practices.
References 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities.
10 CFR Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants.
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.
American National Standards Institute (ANSI)/American Nuclear Society (ANS) 8.1-1998 (Reaffirmed 2007), Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors, American Nuclear Society, La Grange Park, Illinois.
Regulatory Guide 3.71, Nuclear Criticality Safety Standards for Fuels and Materials Facilities.
Information Notice No. 91-26, Potential Nonconservative Errors in the Working Format Hansen-Roach Cross-Section Set provided with the KENO and Scale Codes, U.S. Nuclear Regulatory Commission, April 2, 1991.
International Handbook of Evaluated Criticality Safety Benchmark Experiments, Nuclear Science Committee, Nuclear Energy Agency, updated and published annually, https://www.oecd-nea.org/science/wpncs/icsbep/handbook.html.
MCNP5, MCNP - A General Monte Carlo N-Particle Transport Code, Version 5; Volume II:
Users Guide, LA-CP-03-0245, Los Alamos National Laboratory, April 2003.
NUREG-1745, Standard Format and Content for Technical Specifications for 10 CFR Part 72 Cask Certificates of Compliance, June 2001 (Agencywide Documents Access and Management System Accession No. ML011940387).
NUREG/CR-6328, Adequacy of the 123-Group Cross-Section Library for Criticality Analyses of Water-Moderated Uranium Systems, ORNL/TM-12970, Oak Ridge National Laboratory, June 1995.
7-29
NUREG/CR-6361, Criticality Benchmark Guide for Light-Water-Reactor Fuel in Transportation and Storage Packages, ORNL/TM-13211, Oak Ridge National Laboratory, March 1997.
NUREG/CR-6747, Computational Benchmark for Estimation of Reactivity Margin from Fission Products and Minor Actinides in PWR Burnup Credit, ORNL/TM-2000/306, Oak Ridge National Laboratory, October 2001.
NUREG/CR-6801, Recommendations for Addressing Axial Burnup in PWR Burnup Credit Analyses, ORNL/TM-2001/273, Oak Ridge National Laboratory, March 2003.
NUREG/CR-6811, Strategies for Application of Isotopic Uncertainties in Burnup Credit, ORNL/TM-2001/257, Oak Ridge National Laboratory, June 2003.
NUREG/CR-6979, Evaluation of the French Haut Taux de Combustion (HTC) Critical Experiment Data, ORNL/TM-2007/083, Oak Ridge National Laboratory, September 2008.
NUREG/CR-6955, Criticality Analysis of Assembly Misload in a PWR Burnup Credit Cask, ORNL/TM-2004/52, Oak Ridge National Laboratory, January 2008.
NUREG/CR-6998, Review of Information for Spent Nuclear Fuel Burnup Confirmation, ORNL/TM-2007/229, Oak Ridge National Laboratory, December 2009.
NUREG/CR-7108, An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety AnalysesIsotopic Composition Predictions, ORNL/TM-2011/509, Oak Ridge National Laboratory, April 2012.
NUREG/CR-7109, An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety AnalysesCriticality (keff) Predictions, ORNL/TM-2011/514, Oak Ridge National Laboratory, April 2012.
NUREG/CR-7194, Technical Basis for Peak Reactivity Burnup Credit for BWR Spent Nuclear Fuel in Storage and Transportation Systems, ORNL/TM-2014/240, Oak Ridge National Laboratory, April 2015.
NUREG/CR-7205, Bias Estimates Used in Lieu of Validation of Fission Products and Minor Actinides in MCNP Keff Calculations for PWR Burnup Credit Casks, ORNL/TM-2012/544, Oak Ridge National Laboratory, September 2015.
Oak Ridge National Laboratory, SCALE: A Comprehensive Modeling and Simulation Suite for Nuclear Safety Analysis and Design, ORNL/TM-2005/39, Version 6.1, June 2011. Available as CCC-785 from the Radiation Safety Information Computational Center at Oak Ridge National Laboratory, https://rsicc.ornl.gov/Catalog.aspx?c=CCC.
RW-859, Nuclear Fuel Data Survey, Energy Information Administration, https://www.eia.gov/nuclear/spent_fuel/.
UCID-21830, Determination and Application of Bias Values in the Criticality Evaluation of Storage Cask Designs, W.R. Lloyd, Lawrence Livermore National Laboratory, January 1990.
7-30
YAEC-1937, Axial Burnup Profile Database for Pressurized-Water Reactors, Yankee Atomic Electric Company, May 1997. Available as Data Package DLC-201, PWR-AXBUPRO-SNL, from the Radiation Safety Information Computational Center at Oak Ridge National Laboratory, https://rsicc.ornl.gov/Catalog.aspx?c=DLC.
7-31