ML20321A103

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Chapter 9 - Confinement Evaluation
ML20321A103
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Issue date: 04/01/2020
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NUREG-2215
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CONFINEMENT EVALUATION Review Objective The objectives of the U.S. Nuclear Regulatory Commissions (NRCs) confinement review of the dry storage system (DSS) and dry storage facility (DSF) with regard to the confinement features and capabilities of the proposed storage container system is to ensure that radiological releases to the environment would be within the limits established by the regulations and that the stored spent fuel cladding and spent fuel assemblies will be sufficiently protected against degradation that might otherwise lead to gross ruptures. In addition, the review evaluates any proposed confinement-related monitoring systems.

Applicability This chapter applies to the review of applications for specific licenses for an independent spent fuel storage installation (ISFSI) or a monitored retrievable storage installation (MRS), categorized as a DSF. This chapter also applies to the review of applications for a certificate of compliance (CoC) of a DSS for use at a general license facility. Sections, tables, or paragraphs of this chapter that apply only to a DSF-specific license application (for an ISFSI and MRS) are identified with (SL). Sections, tables, or paragraphs that apply only to DSS CoC applications have (CoC). A subsection without an identifier applies to both types of applications.

Areas of Review This chapter provides guidance for use in evaluating the design and analysis of the proposed storage container confinement system for normal, off-normal, and accident conditions. This evaluation includes a more detailed assessment of the confinement-related design features and criteria initially presented in the chapters of the applicants safety analysis report (SAR) on general information and principal design criteria, as well as the proposed confinement monitoring capability, as applicable. In addition, the NRC staff reviews the applicants analyses that assess the potential releases of radionuclides associated with spent nuclear fuel (SNF) and that estimate their potential leakage to the environment and subsequent impact on a hypothetical individual located at or beyond the controlled area boundary.

This chapter addresses the following areas of review:

  • confinement design characteristics

- design criteria

- design features

  • confinement monitoring capability
  • nuclides with potential for release
  • confinement analyses

- normal conditions

- off-normal conditions (anticipated occurrences)

- design-basis accident conditions (including natural phenomenon events)

- identification of release events (SL)

- evaluation of release estimates for SNF and high-level radioactive waste (HLW)

(SL)

- evaluation of release estimates for greater-than-Class-C (GTCC) waste (SL)

  • supplemental information 9-1

Regulatory Requirements and Acceptance Criteria This section summarizes those parts of Title 10 of the Code of Federal Regulations (10 CFR) Part 72, Licensing Requirements for the Independent Storage of Spent Nuclear Fuel and High-Level Radioactive Waste, and Reactor-Related Greater than Class C Waste, that are relevant to the review areas this chapter addresses. The NRC staff reviewer should refer to the exact language in the regulations. Tables 9-1a and 9-1b match the relevant regulatory requirements to the areas of review covered in this chapter. The NRC staff reviewer should verify the association of regulatory requirements with the areas of review presented in the matrix to ensure that no requirements are overlooked as a result of unique design features.

Table 9-1a Relationship of Regulations and Areas of Review for a DSF (SL) 10 CFR Part 72 Regulations Areas of Review 72.24 72.44 72.104 72.106 72.120 72.122 72.126 72.128 Confinement Design (a)(d) (a)(b) (a)

Characteristics Confinement Monitoring (c) (a) (b) (h)(i) (d) (a)

Capability Nuclides with Potential for (a) (b)

Release Confinement Analyses (c) (a) (b) (d) (a)

Table 9-1b Relationship of Regulations and Areas of Review for a DSS (CoC) 10 CFR Part 72 Regulations Areas of Review 72.230 72.234 72.236 Confinement Design Characteristics (a)(b)(d)(e)(g) (j)(l)

Confinement Monitoring Capability (d)(e)(g)(l)

Nuclides with Potential for Release (a)(d)

Confinement Analyses (a)(d)

As prescribed in 10 CFR Part 72, the regulatory requirements for doses at and beyond the controlled area boundary include both the direct dose (i.e., from shielding review) and that from an estimated release of radionuclides to the atmosphere (based on the leak test of the confinement).

Thus, an overall assessment of the compliance of the proposed DSS with these regulatory limits is presented in Chapter 10, Radiation Protection Evaluation, of this standard review plan (SRP).

In addition, the performance of the storage container confinement system under accident conditions, as evaluated in this chapter, may also be addressed in the overall accident analyses presented in Chapter 16, Accident Analysis Evaluation, of this SRP.

In general, the DSS or DSF confinement evaluation seeks to ensure that the proposed design fulfills the following acceptance criteria, which the NRC staff considers to be minimally acceptable, to meet the confinement requirements in 10 CFR Part 72.

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9.4.1 Confinement Design Characteristics The design should provide redundant sealing of the confinement boundary (10 CFR 72.236(e)).

Typically, this means that field closures of the confinement boundary should either have two seal welds or two metallic O-ring seals.

The confinement design should be consistent with the regulatory requirements as well as the applicants general design criteria, reviewed in accordance with Chapter 3, Principal Design Criteria Evaluation, of this SRP. The NRC staff has previously accepted construction of the primary confinement barrier in conformance with the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section III, Rules for Construction of Nuclear Facility Components, Division 1, Subsections NB or NC. The B&PV Code defines the standards for all aspects of construction, including materials, design, fabrication, examination, testing, inspection, and certification, required in the manufacture and installation of components. In such instances, the staff has relied upon Section III to define the minimum acceptable margin of safety.

Therefore, the applicant must fully document and completely justify any deviations from the specifications of Section III. In some cases, after careful and deliberate consideration, the staff has made exceptions to this requirement. In addition, ASME published in 2005 Division 3 to Section III, which is written specifically for containments for the transportation and storage of SNF, but the NRC has not yet endorsed it.

The design must provide a nonreactive environment to protect fuel assemblies against fuel matrix degradation and fuel cladding degradation, which might otherwise lead to gross rupture (Knoll and Gilbert 1987). Measures for providing a nonreactive environment within the confinement storage container typically include drying and backfilling with a nonreactive cover gas (such as helium).

To reduce the potential for fuel oxidation and subsequent cladding failure, an inert atmosphere (e.g., helium cover gas) has been used for storing uranium dioxide (UO2) SNF in a dry environment. Chapter 11, Operation Procedures and Systems Evaluation, of this SRP provides more detailed information on the cover gas filling process. Note that other fuel types, such as graphite fuels for the high-temperature, gas-cooled reactors, may not exhibit the same oxidation reactions as UO2 fuels and, therefore, may not require an inert atmosphere; however, the application should discuss the prevention of fuel and cladding degradation.

(SL) The SAR must describe the confinement system for SNF, HLW, and waste management facilities. Chapter 13, Waste Management Evaluation, of this SRP discusses the review of waste management facilities.

(SL) If appropriate, the SAR must also describe the confinement features or system implemented for reactor-related GTCC waste. The applicant should provide assurance that the reactor-related GTCC waste will be adequately contained and shielded under normal, off-normal, and accident conditions in accordance with the 10 CFR Part 72 dose limits.

9.4.2 Confinement Monitoring Capability (SL) Confinement monitoring for an ISFSI and MRS has two aspects. The first is monitoring storage confinement closure seals or overall closure effectiveness. The second is providing a system to measure radionuclides released to the environment under normal, off-normal, and accident conditions. This second aspect includes all areas where there is the potential for significant releases to the environment and may include storage containers, pool facilities, and waste management facilities; Chapter 13 discusses the review of releases other than from storage containers, such as pool facilities and waste management facilities. The SAR should present a 9-3

discussion of the extent of monitoring required consistent with 10 CFR Part 72 requirements for both of these aspects of confinement monitoring.

The application should describe the proposed monitoring capability and surveillance plans for mechanical closure seals. In instances involving welded closures, the staff has accepted that no closure monitoring system is required. This practice is consistent with the fact that other welded joints in the confinement system are not monitored because the initial staff review considers the integrity of the confinement boundary for the licensing period. For welded closures, typical surveillances include checking for blockage of the air vents or temperature monitoring.

To show compliance with the requirement for continuous monitoring, 10 CFR 72.122(h)(4),

storage container vendors have proposed, and the staff has accepted, routine surveillance programs and active instrumentation to meet the continuous monitoring requirements.

(SL) For reactor-related GTCC waste, the SAR should describe the programs and procedures in place to maintain confinement of the GTCC waste and prevent degradation of the waste form and containers. In general, the SAR should describe programs that give full consideration to maximum anticipated storage time for any projected corrosion to ensure that the dose limits established in 10 CFR Part 72 are not exceeded.

9.4.3 Nuclides with Potential for Release Verify that the applicant estimated the maximum credible quantity of radionuclides with the potential for release to the environment. The radionuclides potentially available for release to the environment would be based on or derived from the same calculation as the radiological source term presented in Chapter 6, Shielding Evaluation, of this SRP.

9.4.4 Confinement Analyses The application should specify the maximum allowed leakage rates for the total primary confinement boundary and redundant seals. The maximum allowed leakage rate is based on the as-tested leak rate measured by the leak test performed on the entire confinement boundary.

Generally, as discussed below in the review procedures, the applicant evaluates the allowable leakage rate for its radiological consequences and its effect on maintaining an inert atmosphere within the storage container. However, the analyses discussed below are unnecessary 1 for a storage container, including its closure lid, that is designed and tested to be leaktight as defined in American National Standards Institute (ANSI) N14.5, American National Standard for Leakage Tests on Packages for Shipment of Radioactive Materials. Additional items to consider include the following:

  • The analysis of potential releases should be consistent with the methods described in ANSI N14.5.

1 The guidance provided in Sections 9.5.3 and 9.5.4 is not applicable for casks that are demonstrated to be leaktight, as defined in ANSI N14.5, recognizing that confinement boundary failure under design-basis normal, off-normal, and accident conditions is not acceptable and that the confinement boundary is to remain leaktight under all conditions.

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  • During normal operations and anticipated occurrences, verify that dose calculations based on the allowable leakage rate demonstrate that the annual dose equivalent to any real individual who is located beyond the controlled area does not exceed the limits given in 10 CFR 72.104(a).
  • For any design-basis accident, verify that dose calculations based on the allowable leakage rate demonstrate that an individual at the boundary or beyond the nearest boundary of the controlled area does not receive a dose that exceeds the limits given in 10 CFR 72.106(b) (discussed further in Chapter 16 of this SRP).
  • Verify that the analysis of potential releases demonstrates that an inert atmosphere will be maintained within the storage container during the licensed storage lifetime.
  • For storage containers that employ a pressurized inert gas to facilitate internal natural convection heat transfer, verify that the analysis of potential releases demonstrates that the pressurized atmosphere will be maintained within the storage container and keep temperatures below allowable limits during the licensed storage lifetime.

9.4.5 Supplemental Information The application should include all supportive information or documentation that justifies assumptions or analytical procedures.

Review Procedures Figures 9-1a and 9-1b show the interrelationship between the confinement evaluation and the other areas of review described in this SRP for specific licenses and CoC applications, respectively. The text within those chapters and sections that are related to confinement will help the confinement review.

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Chapter 1 - General Chapter 2 - Site Chapter 3 - Principal Chapter 4 - Structural Information Evaluation Characteristics Design Criteria Evaluation Evaluation (SL) Evaluation

  • DSF Description and
  • Geography and Demography
  • Classification of SSCs
  • Description of the SSCs Operational Features
  • Nearby Facilities
  • Design Bases for SSCs
  • Design Criteria
  • Engineering Drawings
  • Meteorology Important to Safety
  • Normal and Off-normal
  • Contents
  • Surface and Subsurface
  • Design Criteria for Safety Conditions Hydrology Protection Systems
  • Accident Conditions
  • Geology and Seismology Chapter 5 - Thermal Chapter 8 - Materials Evaluation Evaluation
  • Material and Design Limits
  • Material Properties
  • Analytical Methods, Models,
  • Environmental Degradation; and Calculations Chemical and Other Reactions
  • Code Use and Quality Standards Chapter 9 - Confinement Evaluation (SL)

Confinement Design Characteristics Confinement Analyses

  • Design Criteria
  • Normal Conditions
  • Design Features
  • Off-Normal Conditions (Anticipated Occurrences)
  • Design-Basis Accident Conditions Confinement Monitoring Capability (Including Natural Phenomenon Events)
  • Reactor-Related GTCC Waste
  • Identification of Release Events
  • Evaluation of Release Estimates for SNF and HLW (SL)
  • Evaluation of Release Estimates for GTCC Waste (SL)

Nuclides with Potential for Release Supplemental Information Chapter 8 - Materials Chapter 10A (SL) -

Evaluation Radiation Protection Evaluation

  • Material Properties
  • Radiation Exposures and Dose
  • Health Physics Program (SL)
  • Code Use and Quality Standards Chapter 11 - Operation Chapter 12 - Conduct Chapter 16 - Accident Chapter 17 - Technical Procedures and of Operations Analysis Evaluation Specifications Systems Evaluation Evaluation Evaluation
  • Storage Container Loading
  • Acceptance Tests
  • Cause of Event
  • Functional and Operating
  • Storage Container Unloading
  • Preoperational Testing and
  • Detection of Event Limits, Monitoring Instruments,
  • Analytical Sampling (SL) Startup Operations (SL)
  • Event Consequences and and Limiting Control Settings
  • Maintenance Program
  • Normal Operations (SL) Regulatory Compliance
  • Limiting Conditions
  • Maintenance Program
  • Corrective Course of Action
  • Personnel Selection, Training, and Certification (SL)

Figure 9-1a Overview of Confinement Evaluation of Specific License Applications for a DSF (SL) 9-6

Chapter 1 - General Chapter 3 - Principal Chapter 4 - Structural Chapter 5 - Thermal Information Evaluation Design Criteria Evaluation Evaluation Evaluation

  • DSS Description and
  • Classification of SSCs
  • Description of the SSCs
  • Material and Design Limits Operational Features
  • Design Bases for SSCs
  • Design Criteria
  • Analytical Methods, Models,
  • Engineering Drawings Important to Safety
  • Normal and Off-normal and Calculations
  • Design Criteria for Safety Conditions Protection Systems
  • Accident Conditions Chapter 6 - Shielding Chapter 8 - Materials Evaluation Evaluation
  • Shielding Design Description
  • Material Properties
  • Radiation Source Description
  • Environmental Degradation;
  • Shielding Model Specification Chemical and Other Reactions
  • Shielding Analyses
  • Code Use and Quality Standards Chapter 9 - Confinement Evaluation (CoC)

Confinement Design Characteristics Confinement Analyses

  • Design Criteria
  • Normal Conditions
  • Design Features
  • Off-Normal Conditions (Anticipated Occurrences)
  • Design-Basis Accident Conditions Confinement Monitoring Capability Nuclides with Potential for Release Supplemental Information Chapter 10B Chapter 11 - Chapter 12 -

Chapter 8 -

(CoC) - Operation Conduct of Materials Radiation Procedures and Operations Evaluation Protection Systems Evaluation Evaluation Evaluation

  • Material Properties
  • Occupational
  • Storage Container
  • Acceptance Tests
  • Maintenance Program and Retrievability
  • Exposures at or
  • Storage Container
  • Code Use and Quality Beyond the Unloading Standards Controlled Area Boundary Chapter 16 - Chapter 17 -

Accident Technical Analysis Specifications Evaluation Evaluation

  • Cause of Event
  • Functional and
  • Detection of Event Operating Limits,
  • Event Consequences Monitoring and Regulatory Instruments, and Compliance Limiting Control
  • Corrective Course of Settings Action
  • Limiting Conditions Figure 9-1b Overview of Confinement Evaluation of Applications for a DSS (CoC) 9-7

9.5.1 Confinement Design Characteristics Design Criteria Review the principal design criteria presented in the SAR, as well as any additional detail provided in the chapter of the SAR on confinement.

Design Features Review the general description of the storage container presented in the SAR, as well as any additional information provided in the chapter of the SAR on confinement. Verify that all drawings, figures, and tables describing confinement features provide sufficient detail to support in-depth staff evaluation.

Verify that the applicant has clearly identified the confinement boundaries. This identification should include the confinement vessel, its penetrations, valves, seals, welds, and closure devices and corresponding information concerning the redundant sealing. Details of the closure confinement boundary are found in Section 8.5.3, especially Figures 8-2 and 8-3, of this SRP.

Verify that the design and procedures provide for drying and evacuation of the storage container interior as part of the loading operations. Also, verify that the confinement design is acceptable for the pressures that may be experienced during normal, off-normal, and accident conditions.

Verify that, on completion of storage container loading and reaching thermal equilibrium, the gas fill of the storage container interior is at a positive pressure level that is expected to maintain a nonreactive environment and heat transfer capabilities of the storage container interior under both normal and off-normal conditions and events for the license period. Verification can include pressure testing, leak testing, seal monitoring, and maintenance for storage containers with seals that are not welded if these are included as described in Chapter 17, Technical Specifications Evaluation, of this SRP as conditions of use. Acceptance tests for pressure testing and leak testing are described in Sections 12.5.2.1, Structure and Pressure Tests, and 12.5.2.2, Leak Tests, of this SRP. Testing and writing the helium leak test procedures should be performed by qualified personnel. NRC Information Notice 2016-04, ANSI N14.5-2014 Revision and Leakage Rate Testing Considerations, contains additional relevant information on leak testing and should be reviewed. In addition, review details of leak testing are found in Section 8.5.3.3.2, Helium Leakage Testing, of this SRP.

Coordinate with the structural and materials disciplines conducting reviews under Chapter 4, Structural Evaluation, and Chapter 8, Materials Evaluation, of this SRP, respectively, to ensure that the applicant has provided proper specifications for all welds and, if applicable, that the bolt torque for closure devices is adequate and properly specified. If applicable, verify the capability of the seal to maintain long-term closure. Because of the performance requirements over the license period (e.g., 20 years, 40 years), evaluate the potential for seal deterioration associated with bolted closures (NRC Information Notice 2013-07). The NRC staff has accepted only metallic seals for the primary confinement. Details on seals are found in Section 8.5.10, Seals, of this SRP. Coordinate this review with the thermal discipline to ensure that the operational temperature range for the seals (specified by the manufacturer) will not be exceeded. For specific licenses in which originally offsite canisters are to be stored, ensure that the integrity of the confinement boundary and the content discussed in the SAR reflects their condition after arrival and being loaded at the site.

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Welded canisters can be used as a confinement system, provided the following design and qualification guidance, as appropriate, is met:

  • The canister is constructed from austenitic stainless steel.
  • The confinement welds meet the guidance of Chapter 8 of this SRP.
  • The canister maintains its confinement integrity during normal conditions, anticipated occurrences, and credible accidents and natural phenomena as required in 10 CFR Part 72.
  • The canister shell has been helium leak tested before its loading as required by 10 CFR 72.236(j). This test demonstrates that the canister is free of defects that could lead to a leakage rate greater than the design-basis leakage rate, which could result in doses at the control area boundary in excess of the regulatory limits.
  • Activities related to inspection, evaluation, documentation of fabrication, and closure welding of canisters are to be performed in accordance with an NRC-approved quality assurance program as required in 10 CFR Part 72, Subpart G, Quality Assurance.

(SL) For reactor-related GTCC waste, review the general description of the reactor-related GTCC waste confinement systems presented in the SAR. Verify that the programs and procedures in place concerning the confinement system for reactor-related GTCC waste are clearly identified in relation to the form of the GTCC waste. Acceptable program descriptions specify the maximum leakage rate from each reactor-related GTCC container or the maximum leakage rate permitted from the total reactor-related GTCC inventory at the ISFSI or MRS.

9.5.2 Confinement Monitoring Capability The NRC staff has found that storage containers closed entirely by welding do not require seal monitoring. However, for storage containers with bolted closures, the staff has found that a seal monitoring system is required to adequately demonstrate that seals can function to limit releases and maintain an inert atmosphere in the storage container. A seal monitoring system, combined with periodic surveillance, enables a determination as to when to take corrective action to maintain safe storage conditions.

Although the details of the monitoring system may vary, the general design approach has been to pressurize the region between the redundant seals with a nonreactive gas to a pressure greater than that of the storage container cavity and the atmosphere. The storage container lid design should prevent exposure of the outer seal to the atmosphere and potential resulting deterioration.

The monitoring system is leak tested to the same leak rate as the confinement boundary.

Installed instrumentation is routinely checked per surveillance requirements. A decrease in pressure between these seals indicates that the nonreactive gas is leaking either into the storage container cavity or out to the atmosphere. For normal operations, radioactive material should not be able to leak to the atmosphere; hence, this design allows for detecting a faulty seal without radiological consequence. Note that the volume between the redundant seals should be pressurized using a nonreactive gas, thereby preventing contamination of the interior cover gas.

If the region between redundant, confinement boundary, mechanical seals is maintained at a pressure greater than that in the storage container cavity, the monitoring system boundaries are tested to a leakage equal to the confinement boundary, the pressure is routinely checked, and the 9-9

instrumentation is verified to be operable in accordance with a technical specification surveillance requirement, the NRC staff has accepted that no discernible leakage is credible for the pressure monitoring system and, therefore, the pressure monitoring system does not have to be included in the confinement dose calculations at the controlled area boundary from atmospheric releases during normal conditions.

The staff has accepted the classification of monitoring systems as not important to safety for those systems designed such that failure of the monitoring system alone would not result in a gross release of radioactive material. This is because, although its function is to monitor confinement seal integrity, the failure of the monitoring system alone would not result in a gross release of radioactive material. It is classified as not important to safety because most of the associated hardware has not met the program controls important to safety, such as design or procurement.

Consequently, the monitoring system for bolted closures need not be designed to the same requirements as the confinement boundary (i.e., ASME B&PV Code,Section III). Additional review details associated with monitoring are described in Section 3.5.3.2, Other Safety Protection Systems, of this SRP.

Depending on the monitoring system design, there could be a lag time before the monitoring system indicates a postulated degraded seal leakage condition. Degraded seal leakage is leakage greater than the tested rate that is not identified within a few monitoring system surveillance cycles. The occurrence of a degraded seal without detection is considered a latent condition and should be presumed to exist concurrently with other off-normal and design-basis events (see Section 3.5.2.4, External Conditions, of this SRP). Verify that once the degraded seal condition is detected, the storage container user will initiate corrective actions.

For the latent condition, the monitoring system boundary would remain intact, and this condition would be bounded by the off-normal analysis. If the monitoring system would not maintain integrity under design-basis accident conditions, additional safety analysis may be necessary.

The staff recognizes that the possibility of a degraded seal condition is small and that the possibility of a degraded seal condition concurrent with a design-basis event that breaches the monitoring system pressure boundary is very remote. However, these probabilities have not been quantified. To address this concern, the staff has accepted a demonstration that the dose consequences of this event are within the limits of 10 CFR 72.106(b).

Verify that the specified pressure of the gas in the monitored region is higher than both the storage container cavity and the atmosphere. Coordinate with the structural and thermal reviewers (Chapter 4 and Chapter 5, respectively) of this SRP to verify the pressure in the storage container cavity.

Verify that the SAR indicates the total volume of gas in the cavity is such that normal seal leakage will not cause all of the gas to escape over the lifetime of the storage container. Confirm that the proposed maximum leakage rate is based on the confinement evaluation described in Sections 9.5.3 and 9.5.4 below. Verify that the maximum allowable leakage rate is specified as a minimum acceptance test criterion in the chapters of the SAR on acceptance tests and the maintenance program and on technical specifications and operating controls and limits, even though the actual leakage rate of the seals is expected to be significantly lower.

For redundant welded closures, ensure that the applicant has provided adequate justification that the welds have been sufficiently designed, fabricated, tested, and examined to ensure that the weld will behave similarly to the adjacent parent material of the storage container.

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Verify that any leakage test, monitoring, or surveillance conditions are appropriately and consistently specified in the chapters of the SAR on acceptance tests and the maintenance program, accident analysis, and technical specifications and operational controls and limits and in the CoC, as applicable. Discussion of acceptance tests is in Section 12.5.2, Acceptance Tests, of this SRP.

9.5.3 Nuclides with Potential for Release For determination of the radionuclide inventory available for release, the NRC staff has accepted, as a minimum for the analysis, the activity from the cobalt-60 in the crud, the activity from iodine, fission products that contribute greater than 0.1 percent of design-basis fuel activity, and actinide activity that contributes greater than 0.01 percent of the design-basis activity. In some cases, the applicant may have to consider additional radioactive nuclides, depending on the specific analysis. The total activity of the design-basis fuel should be based on the storage container design loading that yields the bounding radionuclide inventory (considering initial enrichment, burnup, and cool time). If necessary, the output of the depletion codes used by the shielding reviewer can provide nuclide quantities and can be used as an independent confirmation of the values described in the SAR confinement chapter.

The staff has determined that, as a minimum, the fractions of radioactive materials available for release from SNF, provided in Table 9-2 for pressurized-water reactor (PWR) fuel and boiling-water reactor (BWR) fuel for normal, anticipated occurrences (off-normal), and accident conditions, should be used in the confinement analysis to demonstrate compliance with 10 CFR Part 72. These fractions account for radionuclides trapped in the fuel matrix and radionuclides that exist in a chemical or physical form that is not releasable to the environment under credible normal, off-normal, and accident conditions. Other release fractions may be used in the analysis, provided the applicant properly justifies the basis for their usage. For example, the staff has accepted, with adequate justification, reduction of the mass fraction of fuel fines that can be released from the storage container. Also, when an applicant uses the release fractions in Table 9-2, ensure that the condition of the fuel described in the SAR is bounded by the experimental data presented in NUREG/CR-6487, Containment Analysis for Type B Packages Used to Transport Various Contents, issued November 1996. Specifically, these experimental data are based on low burnup fuel and the release from a single breach of one fuel rod; these data should not be used for SNF described as damaged. The reviewer may consider other release fractions for conditions other than those described in NUREG/CR-6487 if the applicant has provided adequate justification.

For fuel rods that are classified as damaged, verify that the applicant has established release fractions (particulates, gases, cred, volatiles) for normal/accident conditions based on applicable physical data and other analyses that account for the specific type of fuel, estimated number of damaged fuel rods, presence of a damaged fuel can, impacts of accidents, and damaged condition of the DSS following an impact.

Fuel rods that are damaged because of a preloading cladding breach may not have a driving force for the release of particulate from the rod under normal or off-normal conditions, providing the canister is not pressurized. However, under an impact accident, damaged fuel rods might release additional fuel fines from the fracture of the fuel, especially the rim region in high burnup fuel. In addition, some canisters may be pressurized to several atmospheres and storage container blowdown could also affect releases. Alternatively, a leak-tight confinement boundary may be specified to preclude the release analyses of damaged fuel.

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Table 9-2 Fractions of Radioactive Materials Available for Release from Spent Fuela Variable Fractions Available for Releaseb PWR and BWR Fuel Normal and Off- normal Design-Basis Accident Conditions Conditions Fraction of Fuel Rods Assumed to 0.01 (normal) 1 Fail 0.10 (off-normal)

Fraction of Gases Released from a 0.3 0.3 Cladding Breach, fGc Fraction of Volatiles Released from a 2x10-4 2x10-4 Cladding Breach, fVc Mass Fraction of Fuel Released as 3x10-5 3x10-5 Fines from a Cladding Breach, fF Fraction of Crud that Spalls Off 0.15d 1.0d Cladding, fC a Values in this table are taken from NUREG/CR-6487.

b Except for cobalt-60, only failed fuel rods contribute significantly to the release. Total fraction of radionuclides available for release should be multiplied by the fraction of fuel rods assumed to have failed.

c In accordance with NUREG/CR-6487, gases species include hydrogen-3, iodine-129, krypton-81, krypton-85, and xenon-127; volatile species include cesium-134, cesium-135, cesium-137, ruthenium-103, ruthenium-106, strontium-89, and strontium-90.

d The source of radioactivity in crud is cobalt-60 on fuel rods. At the time of discharge from the reactor, the specific activity, Sc, is estimated to be 140 microcuries per square centimeter (Ci/cm2) for PWRs and 1,254 Ci/cm2 for BWRs. Total cobalt-60 activity is this estimate times the total surface area of all rods in the storage container (Sandoval et al. 1991). Decay of cobalt-60 to determine activity at the minimum time before loading is acceptable.

The staff has accepted the rod breakage fractions in Section 5.5.4.6, Pressure Analysis, of this SRP for the confinement evaluations. It is important to recognize that confinement boundary failure under design-basis normal, off-normal, or accident conditions is not acceptable.

Confinement boundary structural integrity during design-basis conditions is confirmed by the structural analysis. The confinement analyses demonstrate that, at the measured leakage rates and assumed relevant nominal meteorological conditions, the requirements of 10 CFR 72.104(a) and 10 CFR 72.106(b) can be met. Each DSS or DSF is also required to have a site-specific confinement analysis and dose assessment to demonstrate environmental compliance with these regulations for SNF, HLW, and reactor-related GTCC waste containers.

9.5.4 Confinement Analyses In general, the NRC evaluates analyses for normal, off-normal, and accident conditions. The reviewer should note that the dose limits differ between 10 CFR 72.104(a) (annual limits for normal plus off-normal conditions) and 10 CFR 72.106(b) (limit per event for accident conditions).

For 10 CFR 72.104(a), the limits are for whole body dose and doses to the thyroid and any other critical organ. These limits are based on the methodology in the International Commission on Radiological Protection (ICRP) Publication 2, Report of Committee II on Permissible Dose for Internal Radiation, issued in 1959. For 10 CFR 72.106(b), the limits are for total effective dose equivalent (TEDE), the sum of the deep dose equivalent (DDE) and the committed dose equivalent (CDE) for any individual organ or tissue, lens dose equivalent (LDE), and shallow dose equivalent (SDE) to skin or any extremity. These limits are based on the methodology in ICRP Report 26, Recommendations of the International Commission on Radiological Protection, issued in 1977. As noted in SRP Chapter 10A, Radiation Protection Evaluation for Dry Storage 9-12

Facilities, and later in this chapter, the NRC has accepted the use of other dose quantities as surrogates for whole body dose, that is, TEDE and effective dose equivalent from external exposures (EDEX).

Review the applicants confinement analysis and the resulting doses for the normal, off-normal, and accident conditions at the controlled area boundary. The analysis typically includes the following common elements:

  • calculation of the specific activity (curies per cubic centimeter) for each radioactive isotope in the storage container cavity based on rod breakage fractions, release fractions, isotopic inventory, and cavity free volume
  • using the tested leak rate and conditions during testing as input parameters, calculation of the adjusted maximum confinement boundary leakage rates (cubic centimeters per second) under normal, off-normal, and accident conditions (e.g., temperatures and pressures)
  • calculation of isotope specific leak rates Qi (curies per second) by multiplying the isotope specific activity by the maximum confinement boundary leakage rates for normal, off-normal, and accident conditions
  • determination of doses for which limits are specified in 10 CFR 72.104(a) and 10 CFR 72.106(b) from inhalation and immersion exposures at the controlled area boundary (considering atmospheric dispersion factors -/Q; (second per cubic meters),

as described in Regulatory Guide (RG) 1.145, Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants)

Verify that the application specifies maximum allowable as tested confinement boundary leakage rates as a technical specification, as discussed in SRP Chapter 17. Guidance on the calculations of the specific activity for each isotope in the storage container and the maximum allowable helium confinement boundary leakage rates for normal, off-normal, and accident conditions can be found in NUREG/CR-6487 and ANSI N14.5. The minimum distance between the storage containers and the distance to the controlled area boundary is generally also a design criterion; however, 10 CFR 72.106(b) requires this distance to be at least 100 meters (328 feet) from the DSS or DSF.

For dose calculations, the NRC staff has accepted the use of either an adult breathing rate (BR) of 2.5x10-4 cubic meters per second (m3/s) (8.8x10-3 cubic feet per second (ft3/s)), as specified in RG 1.109, Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I, or a worker breathing rate of 3.3x10-4 m3/s (1.2x10-2 ft3/s), as specified in the Environmental Protection Agency (EPA)

Federal Guidance Report No. 11, Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion, issued September 1988. Ensure that the calculation uses the dose conversion factors (DCFs) in EPA Federal Guidance Report No. 11 for committed effective dose equivalent (CEDEthe total dose to the body from internal exposures) and the CDE (the dose to an organ from internal exposures) for the thyroid and other organs from inhalation. Confirm that the SAR reflects a bounding DCF from EPA Federal Guidance Report No. 11 for each isotope unless the applicant justifies an 9-13

alternate value. The staff does not accept weighting or normalization of the DCFs. For each isotope (i), CEDEi or CDEi is calculated as follows:

CEDEi or CDEi = Qi

  • DCFi */Q
  • BR
  • Duration
  • conversion factor The conversion factor, if required, converts the input units into the desired form (e.g., Sv, rem, mrem, mSv). The duration term is 1 year for normal conditions and an appropriate duration for each individual off-normal and accident condition. Thus, the results should be in terms of mrem.

However, the dose for an off-normal condition is summed with the annual normal condition dose to give a total annual dose (mrem in a year) to evaluate compliance with 10 CFR 72.104(a) limits.

This also applies to the calculations of doses described in the equations below.

For the contributions to the EDEX (total dose to the body from external exposures) and the dose equivalent (DEext) to the thyroid, other organs, and the skin from air submersion (external) exposure, ensure that the SAR reflects the DCFs in EPA Federal Guidance Report No. 12, External Exposure to Radionuclides in Air, Water, and Soil, issued September 1993. Again, the NRC staff does not accept weighting or normalization of the DCFs.

The EDEXi, the DEext,i for each organ, and the SDEi are calculated as follows:

EDEXi, DEext,I, or SDEi = Qi

  • DCFi */Q
  • BR
  • Duration
  • conversion factor The description above for the duration and conversion-factor terms apply in this equation as well.

Summing the calculated doses over all isotopes (i) results in the total effluent contributions for the CEDE, EDEX, CDE, DEext, and SDE. For compliance with the limits in 10 CFR 72.104(a) and 10 CFR 72.106(b) that include internal and external dose contributions, ensure that the SAR uses the following equations:

TEDE = CEDE + EDEX For a given organ or tissue, the total dose to the organ or tissue = CDE + DEext 10 CFR 72.106(b) organ doses = EDEX + CDE As already described, the actual dose limits in 10 CFR 72.104(a) include a limit for whole body dose. EPA Federal Guidance Report Nos. 11 and 12 do not give DCFs for whole body dose because of the differences in dose methodology compared to the regulatory limit. However, as noted in Chapter 10A of this SRP, the NRC has accepted the use of TEDE as a surrogate for whole body dose. Based on information in NRCs regulatory guidance, the EDEX may also be an appropriate surrogate for whole body dose when doses are calculated for uniform body exposures associated with semi-infinite cloud dose modeling (see RG 1.195, Methods and Assumptions for Evaluating Radiological Consequences of Design Basis Accidents at Light-Water Nuclear Power Reactors, Section 4.1.4). This calculation approach is consistent with the analysis assumptions that are the basis of the confinement evaluation.

The limits in 10 CFR 72.104(a) include limits for critical organs. EPA Federal Guidance Report Nos. 11 and 12 give DCFs for some of the critical organs for the radionuclides (critical organs vary from one radionuclide to another) considered in the confinement analysis. Because the doses from effluents have been very small compared to the 10 CFR 72.104(a) dose limits and compared to the direct radiation doses for the analyzed organs, the NRC expects that the doses to the other critical organs for the analyzed radionuclides, for which DCFs are not provided, would also be 9-14

similarly very small. However, in cases where analyzed organ doses from effluents are relatively significant and analyzed doses are close to the limits, calculations for the other critical organs using appropriate methods may be necessary.

Note that the actual organ dose limits in 10 CFR 72.106(b) are stated to be the summation of the CDE for the organ or tissue and the DDE. However, EPA Federal Guidance Report No. 12 does not include DCFs for the DDE. A true calculation of DDE may likely require the use of computer codes that are capable of analyses for external doses from effluent plumes and that include DDE as an analytical result. The DCFs in Report No. 12 calculate the EDEX. Based on information in NRCs regulatory guidance, the EDEX is nominally equivalent to the DDE if the whole body is irradiated uniformly for submergence (in a semi-infinite cloud) exposure situations (see RG 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, Section 4.1.4). The assumptions that are the basis of the confinement evaluation are consistent with these conditions. Hence the equation above for the 10 CFR 72.106(b) organ doses is written using the EDEX instead of the DDE.

The limits in 10 CFR 72.106(b) also include limits for LDE. EPA Federal Guidance Report No. 12 does not include DCFs for LDE. While not the same as LDE, the DDE and the SDE may be acceptable surrogates for estimating the LDE based on the following. Various National Council on Radiation Protection (NCRP) and ICRP reports (e.g., NCRP Report Number 122, Use of Personal Monitors to Estimate Effective Dose Equivalent and Effective Dose to Workers For External Exposure to Low-LET Radiation, issued in 1995, and ICRP Publication 103, The 2007 Recommendations of the International Commission on Radiological Protection) indicate that SDE and DDE may be used for LDE under certain conditions, including the following, which are consistent with the analysis approach for the confinement evaluation. First, the analyses assume uniform external exposure of the body from an effluent plume. Second, the effluent contribution to dose is minor compared to the contribution from direct radiation, or the total dose is significantly less than the regulatory limits. Additionally, Bordy (2015) indicates that the SDE and DDE can be bounding for LDE over most gamma energies of interest. That SDE and DDE do not bound LDE for all gamma energies of interest would be acceptable given the second condition described above for using SDE and DDE to estimate LDE. For instances where the second condition is not met, an appropriately justified factor should be applied to the DDE or SDE to account for gamma energies where they would under predict LDE.

Normal Conditions For normal conditions, a bounding exposure duration assumes that an individual is present at the controlled area boundary for 1 full year (8,760 hours0.0088 days <br />0.211 hours <br />0.00126 weeks <br />2.8918e-4 months <br />). The NRC staff may consider an alternative exposure duration if the applicant provides justification.

Because any potential release resulting from confinement boundary leakage would typically occur over a substantial period of time, the staff has accepted calculation of the atmospheric dispersion factors (/Q) according to RG 1.145, assuming D-stability diffusion and a wind speed of 5 meters per second (m/s) (16 feet per second (ft/s)).

(SL) For a DSF, the number of storage containers will be known based on the (proposed) license condition that limits the amount of SNF, HLW, and reactor-related GTCC waste that can be stored at the facility. Thus, the analyses for normal conditions should be for the planned facility storage container array(s) and the number of storage containers that will be used at the DSF.

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(CoC) As noted above, a DSF will have multiple storage containers. When reviewing a DSS, therefore, confirm that the resulting doses from a single storage container will be a small fraction of the limits prescribed in 10 CFR 72.104(a) to accommodate an array of storage containers and the external direct dose.

Off-Normal Conditions (anticipated occurrences)

Off-normal conditions can affect confinement in a variety of ways (e.g., temperature and pressure within the storage container, larger release pathway); Section 9.5.2 above and Chapter 3 of this SRP provide further discussion on off-normal considerations. For off-normal conditions, the bounding exposure duration and atmospheric dispersion factors (/Q) are the same as those discussed above for normal conditions.

To demonstrate compliance with 10 CFR 72.104(a), the staff has accepted dose calculations for releases from a single storage container undergoing off-normal conditions. However, the dose contribution from storage container leakage should also be a fraction of the limits specified in 10 CFR 72.104(a) because the doses from normal conditions and doses from other radiation sources are added to this contribution. Coordinate this review with the SRP Chapters 6 and 10A/10B reviewers.

Design-Basis Accident Conditions (including natural phenomenon events)

For accident-level conditions, the duration of the release is assumed to be 30 days (720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br />).

A bounding exposure duration assumes that an individual is also present at the controlled area boundary for 30 days. This time period is the same as that used to demonstrate compliance for reactor facilities licensed in accordance with 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities, and provides good defense in depth because recovery actions to limit releases are not expected to exceed 30 days.

For accident conditions, the staff has accepted calculation of the atmospheric dispersion factors

(/Q) of RG 1.145 on the basis of F-stability diffusion and a wind speed of 1 m/s (3.3 ft/s). (Note:

RG 1.23, Meteorological Monitoring Programs for Nuclear Power Plants (Safety Guide 23), and RG 1.111, Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors, provide background information that describes atmospheric dispersion and deposition parameters.)

To demonstrate compliance with 10 CFR 72.106(b), the staff has accepted dose calculations for releases of radionuclides from a single storage container.

Identification of Release Events (SL)

Discuss the proposed site operations with other reviewers (e.g., structural, operations, site characteristics) to determine the spectrum of events that should be considered for the specific site. Focus on the physical condition of the confinement system for normal operations and off-normal operations, and for design-basis accidents. Use these discussions to understand (1) the physical condition of the equipment that might serve to contain radionuclides, and (2) the forces (e.g., physical displacement, pressure differences, temperatures) that could move radionuclides into the accessible environment if the confinement system fails. Categorize the selected events as either (1) normal operations and off-normal operations or (2) design-basis accidents.

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Evaluation of Release Estimates for Spent Nuclear Fuel and High-Level Radioactive Waste (SL)

Refer to Sections 9.5.3 and 9.5.4 (through 9.5.4.4) of this chapter.

Evaluation of Release Estimates for Reactor-Related Greater than Class C Waste (SL)

The issues considered for an evaluation of release estimates for reactor-related GTCC waste are similar to those for SNF; however, the activity and release associated with reactor-related GTCC may be less than that for SNF. For reactor-related GTCC waste, verify that the SAR, at a minimum, presents a clear description of the operating limits regarding the confinement features of the reactor-related GTCC storage design or system. Verify that the application identifies the quantity of radionuclides that would be released to the environment from the ISFSI or MRS during normal operations, off-normal operations, and design-basis accidents. The estimates should be based on an evaluation of the reactor-related GTCC waste form and the physical process that will move radionuclides into the environment or retain them in the confinement system.

Verify that the confinement system, analyses, and procedures demonstrate, with reasonable assurance, that for the package contents and assumed nominal meteorological conditions, the requirements of 10 CFR 72.104(a) and 10 CFR 72.106(b) can be met.

Analysis methods that determine the dose limits are not exceeded may include those used for SNF evaluations. Verify that the reactor-related GTCC dose calculations use the assumptions used for SNF (e.g., meteorological conditions, DCFs, breathing rates, distance of the real individual) unless the applicant can justify alternative assumptions. The applicant must adequately justify the value of the release fractions based on the form of reactor-related GTCC waste and the design of the container.

Verify that each ISFSI or MRS has a site-specific confinement analysis and dose assessment to demonstrate regulatory compliance. Meteorological conditions similar to those used to perform the confinement analyses for SNF or HLW should be used in the analysis. For DCFs, the NRC has accepted the use of EPA Federal Guidance Report Nos. 11 and 12.

9.5.5 Supplemental Information Ensure that all supportive information or documentation has been provided or is readily available.

This includes, but is not limited to, justification of assumptions or analytical procedures, test results, photographs, computer program descriptions, input and output, and applicable pages from referenced documents. Request any additional information needed to complete the review.

Consider relevant generic communications (e.g., NRC information notices) as part of the review.

Evaluation Findings The NRC reviewer should prepare evaluation findings upon satisfaction of the regulatory requirements in Section 9.4. If the documentation submitted with the application fully supports 9-17

positive findings for each of the regulatory requirements, the statements of findings should be similar to the following:

Certificate of Compliance F9.1 Chapter(s) _____ of the SAR describe(s) SSCs important to safety that are relied on for confinement in sufficient detail to permit evaluation of their effectiveness, in accordance with 10 CFR 72.230(a),

10 CFR 72.230(b), and 10 CFR 72.236.

F9.2 The design of the [DSS designation] adequately protects the SNF cladding against degradation that might otherwise lead to gross ruptures, in accordance with 10 CFR 72.236(g). The chapter of the safety evaluation report (SER) on thermal evaluation discusses the relevant temperature considerations.

F9.3 The design of the [DSS designation] provides redundant sealing of the confinement system closure joints, in accordance with 10 CFR 72.236(e),

by _______.

F9.4 The confinement system will be monitored with a _________ monitoring system as discussed above [if applicable] to demonstrate compliance with 10 CFR 72.236(d)(e)(g) and (l). No instrumentation is required to remain operational under accident conditions.

F9.5 The quantity of radioactive nuclides postulated to be released to the environment has been assessed to evaluate compliance with 10 CFR 72.236(d). The SER chapter on radiation protection shows that the dose from these releases will be added to the direct dose to show that the (DSS designation) satisfies the regulatory requirements of 10 CFR 72.104(a) and 10 CFR 72.106(b).

F9.6 The storage container confinement system will be inspected to ascertain that there are no cracks, pinholes, uncontrolled voids, or other defects that could significantly reduce its confinement effectiveness, in accordance with 10 CFR 72.236(j).

F9.7 The storage container confinement system has been evaluated (by appropriate tests or by other means acceptable to the NRC) to demonstrate that it will reasonably maintain confinement of radioactive material under normal, off-normal, and credible accident conditions, in accordance with 10 CFR 72.236(l).

Specific License F9.8 Chapter(s) _____ of the SAR describe(s) structures, systems, and components (SSCs) important to safety that are relied on for confinement in sufficient detail to permit evaluation of their effectiveness, in accordance with 10 CFR 72.128(a).

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F9.9 The quantity of radionuclides postulated to be released to the environment has been assessed as discussed above, in accordance with 10 CFR 72.104(a) and 10 CFR 72.106(b). The SER chapter on radiation protection shows that the dose from these releases will be added to the direct dose to show that the [DSF designation] satisfies the regulatory requirements of 10 CFR 72.104(a) and 10 CFR 72.106(b).

F9.10 If the confinement system is provided by an unsealed system, the following would be applicable: The [DSF designation] includes the following confinement systems that are important to safety and that require monitoring over anticipated ranges for normal and off-normal operations: ________ [identify]. The following monitoring systems must remain operational under accident conditions: ________ [identify]. The SAR acceptably describes instrumentation and control systems that should provide these capabilities, in compliance with 10 CFR 72.122(i) and 10 CFR 72.128(a).

F9.11 The proposed operations of the [DSF designation] provides adequate measures for protecting the SNF cladding against degradation that might otherwise lead to gross ruptures of the material to be stored, in compliance with 10 CFR 72.122(h)(1).

(SL) In the case of the evaluation of releases from confinement for specific licenses, the acceptability of releases can be determined only after reviewing the results of the dose assessment, which is addressed in Chapters 10 and 16 of this SRP.

The reviewer should provide a summary statement similar to the following:

The staff concludes that the design of the confinement system of the [storage container designation] is in compliance with 10 CFR Part 72 and that the applicable design and acceptance criteria have been satisfied. The evaluation of the confinement system design provides reasonable assurance that the [storage container designation] will allow for the safe storage of SNF. This finding is reached on the basis of a review that considered the regulation itself, appropriate regulatory guides, applicable codes and standards, the applicants analysis, and accepted engineering practices.

References 10 CFR Part 20, Standards for Protection Against Radiation.

10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities.

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) N14.5, Radioactive MaterialsLeakage Tests on Packages for Shipment, 2014.

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American Society of Mechanical Engineers (ASME) Boiler and Pressure (B&PV) Code, 2007 Addenda 2008.

Section III, Rules for Construction of Nuclear Facility Components.

Division 1, Metallic Components; Subsections NB and NC Division 3, Containments for Transportation & Storage of Spent Nuclear Fuel and High Level Radioactive Material & Waste (no NRC position on this has been established)

ASME NQA-1-2008, Quality Assurance Requirements for Nuclear Facility Applications, American Society of Mechanical Engineers, New York, NY.

ASME NQA-1A-2009 Addenda to ASME NQA-1-2008, Quality Assurance Requirements for Nuclear Facility Applications, American Society of Mechanical Engineers, New York, NY.

Bordy, J.M. 2015, Monitoring of eye lens doses in radiation protection, Radioprotection 50(3),

177-185.

International Commission on Radiological Protection (ICRP) Publication 2, Report of Committee II on Permissible Dose for Internal Radiation, Pergamon Press, 1959.

ICRP Publication 26, Recommendations of the International Commission on Radiological Protection, Annals of the ICRP, Pergamon Press, 1977.

ICRP Publication 103, The 2007 Recommendations of the International Commission on Radiological Protection, Annals of the ICRP, Elsevier, 2007.

Knoll, R.W. and E.R. Gilbert, Evaluation of Cover Gas Impurities and Their Effects on the Dry Storage of LWR Spent Fuel, PNL-6365, DE88 003983, Pacific Northwest National Laboratory, November, 1987.

National Council on Radiation Protection and Measurements Report No. 122, Use of Personal Monitors to Estimate Effective Dose Equivalent and Effective Dose to Workers for External Exposure to Low-LET Radiation, 1995.

NRC Information Notice 2013-07, Premature Degradation of Spent Fuel Storage Cask Structures and Components from Environmental Moisture, dated April 16, 2013.

NRC Information Notice 2016-04, ANSI N14.5-2014 Revision and Leakage Rate Testing Considerations, dated March 28, 2016 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML16063A287).

NUREG/CR-6487, Containment Analysis for Type B Packages Used to Transport Various Contents, UCRL-ID-124822, Lawrence Livermore National Laboratory, November 1996.

Regulatory Guide 1.23, Meteorological Monitoring Programs for Nuclear Power Plants (Safety Guide 23).

Regulatory Guide 1.109, Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I.

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Regulatory Guide 1.111, Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors.

Regulatory Guide 1.145, Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants.

Regulatory Guide 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors.

Regulatory Guide 1.195, Methods and Assumptions for Evaluating Radiological Consequences of Design Basis Accidents at Light-Water Nuclear Power Reactors.

Sandoval, R.P., R.E. Einziger, H. Jordan, A.P. Malinauskas, and W.J. Mings, Estimate of CRUD Contribution to Shipping Cask Containment Requirements, SAND88-1358, TTC-0811, UC-71, Sandia National Laboratories, January 1991.

U.S. Environmental Protection Agency (EPA) Federal Guidance Report No. 11, Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion, September 1988.

EPA Federal Guidance Report No. 12, External Exposure to Radionuclides in Air, Water, and Soil, September 1993.

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