ML20321A105

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Chapter 10B - Radiation Protection Evaluation for Dry Storage Systems (CoC)
ML20321A105
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Issue date: 04/30/2020
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10B RADIATION PROTECTION EVALUATION FOR DRY STORAGE SYSTEMS (CoC) 10B.1 Review Objective The objective of the U.S. Nuclear Regulatory Commissions (NRCs) radiation protection evaluation is to (1) determine that the proposed spent nuclear fuel (SNF) dry storage system (DSS) complies with the applicable regulatory requirements for radiation protection and (2) ensure that the DSS design and operations include reasonable consideration of, and facilitate licensees compliance with, the requirements that licensees who use the DSS must meet.

For the purposes of this standard review plan (SRP) chapter, radiation protection refers to design and operational elements that are relied upon to limit radiation exposures from normal operations, anticipated occurrences (that is, off-normal conditions), and accidents and natural phenomenon events (collectively referred to as accident conditions or design-basis accidents (DBAs)). This includes those design features that may have a different primary function but are nonetheless credited or considered in the applicants radiation protection evaluation.

10B.2 Applicability This chapter applies to the review of applications for certificates of compliance (CoCs) for DSSs.

As such, the chapter title is denoted with (CoC) to signify that the scope of this chapter applies only to DSSs.

10B.3 Areas of Review The areas of review include means and methods used to protect workers and members of the public, DSS design features, DSS storage configurations, dose assessments and dose assessment methods, and operational elements and procedures.

This chapter addresses the following areas of review:

  • radiation protection design features
  • occupational exposures
  • exposures at or beyond the controlled area boundary

- normal operations and anticipated occurrences

- accidents and natural phenomenon events

  • as low as is reasonably achievable (ALARA) design

- design considerations

- procedures and engineering controls 10B.4 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, High-Level Radioactive Waste and Reactor-Related Greater Than Class C Waste, that are applicable to the review areas this chapter addresses. This section also includes specific 10 CFR Part 72 and 10 CFR Part 20, Standards for Protection Against Radiation, requirements that, while not applicable to CoC applicants (i.e., they only apply to licensees), the reviewer should consider in the review. This is because the radiation protection review may include elements 10B-1

needed to assist the general licensee in meeting these regulatory requirements and to ensure the DSS design and operations include reasonable consideration of and facilitate the licensees compliance with these requirements. The NRC reviewer should refer to the exact language in the applicable regulations. Table 10B-1 matches these regulatory requirements to the areas of review covered in this chapter. However, Table 10B-1 does not represent an exhaustive listing of regulations that may need consideration. Thus, the reviewer should confirm that all applicable regulations are identified and appropriately addressed in the safety analysis report (SAR).

In general, the radiation protection evaluation seeks to ensure that the proposed design fulfills the following acceptance criteria:

  • The SAR demonstrates that the DSS includes shielding and confinement that are sufficient to meet the requirements in 10 CFR 72.104, Criteria for radioactive materials in effluents and direct radiation from an ISFSI or MRS, and 10 CFR 72.106, Controlled area of an ISFSI or MRS, including the dose limits, in compliance with 10 CFR 72.236(d).
  • Dose rates, design features, and operations for the DSS are consistent with and demonstrate appropriate consideration for ALARA principles and objectives.
  • The DSS design includes features that facilitate decontamination to the extent practicable in meeting the requirements in 10 CFR 72.236(i) in minimizing radioactive contamination.

The SAR should address these acceptance criteria. The acceptance criteria are organized according to the areas of review specified in Section 10B.3 above. The reviewer should consider the applicability and implementation of NRC and industry guidance against that presented in the SAR. The radiation protection review also requires coordination with the shielding (SRP Chapter 6, Shielding Evaluation), confinement (SRP Chapter 9, Confinement Evaluation),

operating procedures (SRP Chapter 11, Operation Procedures and Systems Evaluation),

accident analysis (SRP Chapter 16, Accident Analysis Evaluation), and technical specifications (SRP Chapter 17, Technical Specifications Evaluation) reviews.

In general, the acceptance criteria listed in the SAR should adopt by reference appropriate NRC guidance or alternatively cite relevant and appropriate industry codes and standards. The SAR should identify and justify alternative approaches used to demonstrate compliance with applicable NRC guidance and industry codes and standards.

This guidance recognizes that applicants have options on how to demonstrate compliance with the NRC regulations and NRC guidance (e.g., rely only on NRC guidance or use alternative methods). With respect to the implementation of NRC guidance, the SAR should identify whether the NRC guidance has been adopted in whole or in part. The SAR should identify any differences between this SRP chapter and design features, analytical techniques, exposure and dose assessment codes, and procedural measures proposed for the DSS and discuss how the proposed alternatives provide acceptable methods of complying with regulations. In any case, the SAR should provide sufficient information and data for the staff to conduct an independent evaluation in confirming compliance with regulatory requirements and SRP acceptance criteria.

The reviewer will confirm that the applicant has adequately addressed these considerations in the SAR.

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If there are multiple versions of a guidance document, such as a regulatory guide or an industry standard, the applicant should note which version of the guidance document has been adopted in the SAR, whether it is the most current revision, and the basis for using the selected version. In the case of an industry standard, the applicant should consider what, if any, staff position exists with respect to acceptability of the standard and different revisions of the standard as part of that selection. As a result, the reviewer will identify the guidance documents the applicant used and assess whether the version of each document the applicant adopted is adequate for demonstrating compliance with NRC requirements.

Table 10B-1 Relationship of Regulations and Areas of Review 10 CFR Part 72 Regulations Areas of Review 72.104A 72.106(b)A 72.126B 72.236 Radiation Protection Design (a)(b)(c) (a)(1)(2)(4)(5)(6), (d) (b)(d)(g)(i)

Features Occupational Exposures (a)(1)(2)(4)(5)(6) (b)(g)(i)

Exposures at or Beyond the (a)(c) (d) (b)(d)(g)(i)

Controlled Area Boundary ALARA (b) (a)(1)(2)(4)(5)(6), (d) (b)(d)(i) 10 CFR Part 20 RegulationsB Areas of Review 20.1101 20.1201(a) 20.1301(a)(b)

Radiation Protection Design (b)(d)

Features Occupational Exposures (b)

Exposures at or Beyond the (b)(d)

Controlled Area Boundary ALARA (b)(d)

A This requirement applies to CoCs and CoC applications through the requirement in 10 CFR 72.236(d).

B While not directly applicable to CoCs, DSS design should facilitate general licensee compliance with these requirements.

10B.4.1 Radiation Protection Design Features The SAR should describe the DSS design features relied on for shielding and confinement to meet radiation protection criteria and requirements. The descriptions of these features should be in the respective SAR chapters and address the information described in Chapters 6 and 9 of this SRP.

The radiation protection chapter of the SAR should include any additional information that is needed beyond what is in the shielding and confinement chapters to demonstrate that together these features are sufficient to ensure, or enable, compliance with regulatory requirements for radiation protection and ALARA objectives. This information should include an evaluation of the use of DSS features during operations. The descriptions should include any additional or supplemental shielding that is needed for radiation protection whether for occupational workers or 10B-3

the public. This would include features such as shield berms relied on in the dose analyses and shielding used in the DSS preparation area and with the transfer equipment to enable personnel to perform operations safely around the DSS. The SAR should also include descriptions of how the DSS design features facilitate decontamination (see 10 CFR 72.236(i)) as well as inspection and servicing in consideration of regulations such as 10 CFR 72.126(a)(5).

The SAR should also describe any operational controls and limits that are necessary to use the DSS and ensure compliance with regulatory requirements and ALARA objectives. While establishing operational controls and limits for ensuring compliance with requirements such as 10 CFR 72.104(b), 10 CFR 72.104(c), 10 CFR 72.126(a), and 10 CFR 72.126(d) is mainly the responsibility of the general licensee using the DSS, it may be necessary or appropriate for some controls and limits to be included as part of the DSS design. The inclusion of these controls and limits may be needed to support DSS evaluations for 10 CFR 72.236(d) as well as to ensure that the design facilitates the licensees compliance with other requirements. These controls and limits may include surface dose rate limits and measurement requirements for prominent DSS design features that are important for doses to personnel or the public. These controls and limits may also include limits and measurement requirements for (removable) contamination. The SAR descriptions should also include the bases for the proposed controls and limits. Some of these may be included as conditions of the CoC, in the technical specifications (see Chapter 17 of this SRP). Note that controls and limits include appropriate specifications of the allowable contents for storage in the DSS, which specifications are adequately supported by and used in the analysis in the SAR and defined in the technical specifications. SAR descriptions of operating procedures should also include or reflect implementation of these controls and limits as well as efforts to minimize contamination and ensure doses are ALARA.

10B.4.2 Occupational Exposures The SAR should include estimates of doses to workers as a result of DSS operations. These estimates should include individual and collective doses. Separate estimates may be provided for different activities or operational sequences. For example, the applicant should provide dose estimates for the sequence beginning with DSS loading in the SNF pool and ending with placement of the DSS at the ISFSI pad, the reverse sequence of operations, and for the conduct of maintenance and surveillance activities. Since a general licensee may load multiple DSSs in a year and licensee personnel are likely to perform other functions at the general licensees 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities, or 10 CFR Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants, facility, in addition to dry storage operations, the applicants dose estimates should indicate that individual doses to workers will be well below the dose limits specified in 10 CFR 20.1201(a). Collective doses should be consistent with the objectives of a well-structured ALARA program. Additional justification of acceptability of the DSS design may be necessary for systems with high occupational dose estimates.

Typically, the applicant only needs to estimate doses for normal DSS operations. However, for DSSs for which operational conditions and dose rates to which personnel may be exposed in responding to anticipated occurrences may differ significantly from normal operations, the SAR should also include dose estimates for actions performed to recover from the anticipated occurrences. The SAR operating procedures chapter should also include a description of such recovery actions for such kinds of anticipated occurrences.

The SAR should include information sufficient to support evaluation of compliance with these criteria. This information should include dose rates for representative points on and near the 10B-4

surfaces of the DSS structures, systems, and components (SSCs) for the different DSS configurations encountered during DSS operations. The information should also include dose rates at appropriate distances from the DSS and from a DSS array, as appropriate, for personnel performing surveillance and maintenance activities. The SAR should also describe the number of personnel involved in operations and the duration of the operations. Dose rate locations should be consistent with the locations of all personnel involved in the DSS operations, with dose rates for these locations being derived in the SARs shielding evaluation chapter. Dose estimates should be broken down by work tasks, or appropriate groupings of tasks (e.g., a group of tasks involve personnel at the same locations at and around the DSS surfaces for the same DSS configuration). The SAR should describe the bases for all assumptions used in the analysis and the reasonableness of these assumptions.

10B.4.3 Exposures At or Beyond the Controlled Area Boundary The SAR should include analyses of radiation exposures and doses to individuals at or beyond the controlled area boundary for normal operations, anticipated occurrences, and DBAs.

10B.4.3.1 Normal Operations and Anticipated Occurrences The doses during normal operations and anticipated occurrences to any real individual located beyond the ISFSI controlled area may not exceed the values specified in 10 CFR 72.104(a). A real individual is defined as a person who lives, works, or engages in recreation or other activities close to the dry storage facility for a significant portion of the year. For the purposes of the 10 CFR 72.104(a) limits, the analysis excludes the occupational doses radiation workers receive while working.

For DSS CoC applications, responsibility for determining compliance with these limits ultimately rests with the general licensee because demonstration of compliance considers factors that are specific to the licensees site (e.g., geometric arrangement of DSS arrays, distances to the controlled area boundary, information about public areas around the site, and maximum SNF quantity to be stored at the site). However, the CoC applicant is responsible for and must demonstrate that the DSS design complies with the requirements in 10 CFR 72.236 in accordance to 10 CFR 72.234(a). These requirements include that the DSSs shielding and confinement features are sufficient to meet 10 CFR 72.104, including dose limits (see 10 CFR 72.236(d)). 1 Section 10B.5.3.1 of this SRP describes acceptable ways to address this requirement. The analysis should address the contributions from direct radiation and any effluent releases from the DSS when the DSS has a specified, analyzed leak rate. Though, based on design requirements, radioactive materials are not expected to be released from DSSs, a DSS will have a specified, analyzed leak rate and effluent doses when it is not designed and tested to be 1 Note that the requirements in 10 CFR 72.236, Specific requirements for spent fuel storage cask approval and fabrication, are the responsibility of the CoC applicant (Volume 64 of the Federal Register, page 56114 (64 FR 56114), October 15, 1999). Thus, the regulations require the DSS to be designed to meet 10 CFR 72.104 and 10 CFR 72.106 (according to 10 CFR 72.236(d)) and place that responsibility with the DSS designer (CoC applicant). This responsibility cannot be passed to the general licensee through 10 CFR 72.212, Conditions of general license issued under § 72.210, or a 10 CFR Part 50 or 10 CFR Part 52 program. For canister-based systems, this applies to the transfer cask as well. It also applies regardless of the DSSs location; the requirements do not distinguish between a loaded DSS in a 10 CFR Part 50 or 10 CFR Part 52 SNF building or on the co-located ISFSI storage pad. This is consistent with the November 16, 2006, rulemakings definition of the regulatory boundary between 10 CFR Part 72 and 10 CFR Part 50 for criticality safety (71 FR 66648).

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either leak-tight or to meet the noncredible leakage criterion. The materials and confinement evaluation chapters of this SRP should include the details regarding the design criteria and testing related to the DSS leak rate. Also, the analysis should address doses from anticipated occurrences since the dose limits apply to the annual doses from both normal operations and anticipated occurrences.

10B.4.3.2 Accidents and Natural Phenomenon Events The doses to any individual located on or beyond the nearest boundary of the controlled area from any DBA may not exceed the limits specified in 10 CFR 72.106(b). As is described in Section 10B.4.3.1 above, responsibility for determining compliance with the 10 CFR 72.106(b) limits ultimately rests with the general licensee. However, the applicant is responsible for and must demonstrate that the DSS design complies with the requirements in 10 CFR 72.236 in accordance to 10 CFR 72.234(a). These requirements include that the DSSs shielding and confinement features are sufficient to meet 10 CFR 72.106, including the dose limits (see 10 CFR 72.236(d)). Section 10B.5.3.2 of this SRP describes acceptable ways to address this requirement. The analysis should address the contributions from direct radiation and, when the DSS has a specified, analyzed leak rate for DBA conditions, any effluents from the DSS.

10B.4.4 As Low As Is Reasonably Achievable Design The SAR should describe how the applicant has incorporated ALARA principles into the DSS design and operations to enable a general licensee using the DSS to ensure doses to workers and the public will be ALARA.

10B.4.4.1 Design Considerations The applicant should demonstrate that ALARA principles have been incorporated into the DSS design to the extent practical. As part of this demonstration, the SAR should describe the bases for the selection and design of DSS features, including geometric and materials aspects, and include appropriate radiation protection, technological, and economic considerations, as applicable. The SAR should show that the applicant considered ALARA principles as part of the following design elements:

  • geometric design (e.g., physical sizes of design features, surface features and shapes that minimize accumulation of contamination, features that minimize or simplify needed maintenance activities, labyrinthine inlet and outlet vents to reduce radiation streaming)
  • arrangement of design features (e.g., placement of vent paths with respect to the SNF contents)
  • materials design (e.g., type and density of concrete selected to minimize dose rates, application of corrosion- and abrasion-resistant coatings to prevent accumulation of contamination in surface pores of SSCs)

In these and any other appropriate aspects of the design, the applicant should consider how general licensees will need to operate the DSS. Such considerations include means to minimize necessary decontamination efforts, minimize generation of radioactive wastes, and minimize the time required for personnel to perform necessary operations (e.g., provide sufficient space to easily perform all expected operations). Considerations should also include actions necessary to recover from anticipated occurrences. For DSS designs where personnel performing recovery 10B-6

actions may be exposed to significantly higher dose rates as compared to normal operations, the applicant may need to provide further justification that such designs are adequate from an ALARA perspective as well as from a general radiation protection perspective. In the case of design changes (e.g., in an amendment), the applicant should describe how the design changes maintain or improve the DSSs effective implementation of ALARA principles. The SAR should, as appropriate and applicable, describe how the applicant has used its experience with past DSS designs to develop the proposed DSS and improve implementation of ALARA principles.

10B.4.4.2 Procedures and Engineering Controls The SAR should describe plans and procedures that have been developed in accordance with applicable guidance of SRP Chapter 11. These plans and procedures should adequately demonstrate the implementation of ALARA principles. This includes describing, in the SARs operating procedures chapter, the use of appropriate engineering controls or equipment that licensees should employ to maintain doses ALARA for DSS operations. The appropriate procedures should also include cautions and warnings regarding streaming paths or other potential radiological hazards (e.g., higher dose rates from radioactive material such as CRUD that is entrained in water being drained from the DSS) for operations that may involve such hazards. The sequencing of procedures should also reflect consideration of ALARA principles as well. The engineering controls and procedures described in the SAR should be founded upon sound engineering design criteria and radiation protection principles.

10B.5 Review Procedures This section describes review procedures for evaluating DSS designs and descriptions of operations with regard to radiation protection requirements, doses to workers and to members of public, and implementation of ALARA principles in the designs and operations of DSSs. The radiation protection review includes evaluation of compliance with all regulatory requirements and acceptance criteria given in this SRP and other applicable NRC documents and accepted codes and standards. The reviewer should always assume that such a comprehensive scope of the review applies, even though it is not further detailed or repeated in this section. Figure 10B-1 shows the interrelationship between the radiation protection evaluation and the other areas of review described in this SRP. Based on its review, as described in the following sections, the reviewer should work with the technical specifications (SRP Chapter 17) reviewer to ensure that any CoC condition regarding preoperational testing includes testing of design features and procedures that are significant to radiation protection, as appropriate.

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Chapter 1 - General Chapter 3 - Chapter 6 - Chapter 8 - Chapter 9 -

Information Principal Design Shielding Materials Confinement Evaluation Criteria Evaluation Evaluation Evaluation

  • DSS Description and
  • Classification of SSCs
  • Shielding Design
  • Material Properties
  • Confinement Design Operational Features
  • Design Bases for SSCs Description
  • Environmental Characteristics
  • Engineering Drawings Important to Safety
  • Radiation Source Degradation; Chemical
  • Confinement Monitoring
  • Contents
  • Design Criteria for Definition and Other Reactions Capability Safety Protection
  • Shielding Model
  • Nuclides with Potential Systems Specification and Retrievability for Release
  • Shielding Analyses
  • Confinement Analyses Chapter 10B - Radiation Protection Evaluation (CoC)

Radiation Protection Occupational Exposures Exposures at or Beyond the ALARA Design Features Controlled Area Boundary

  • Design Considerations
  • Normal Operations and
  • Procedures and Engineering Occurrences Controls
  • Accident Conditions and Natural Phenomenon Events Chapter 11 - Operation Chapter 16 - Chapter 17 - Technical Procedures and Accident Analysis Specifications Systems Evaluation Evaluation Evaluation
  • Operation Description
  • Event Consequences
  • Functional and Operating
  • Storage Container Loading and Regulatory Limits, Monitoring Instruments,
  • Storage Container Handling Compliance and Limiting Control Settings and Storage Operations
  • Corrective Course of
  • Surveillance Requirements
  • Storage Container Unloading Action
  • Design Features
  • Administrative Controls Figure 10B-1 Overview of Radiation Protection Evaluation 10B.5.1 Radiation Protection Design Features Carefully review the general description and functional features of the DSS and the technical drawings presented in the SAR. In addition, review information on SSCs and design criteria as well as any additional details regarding radiation protection provided in the SAR. Based on this review, the staff should identify those DSS SSCs and features that are relevant to radiation protection. Since some of this information may be in the shielding and confinement chapters of the SAR; coordinate this review with the reviewers of the SAR shielding and confinement chapters to (1) obtain a sufficient understanding of the relevant features and the evaluations of those features, (2) ensure the features are described and analyzed adequately to evaluate their overall effectiveness for radiation protection purposes for all applicable configurations and conditions, and (3) ensure that any identified inadequacies or inconsistencies are appropriately addressed.

Verify that the SAR demonstrates and confirms that the DSS design adequately meets the following criteria:

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  • The DSS features are designed to facilitate decontamination to the extent practicable in accordance with 10 CFR 72.236(i). This includes minimizing contamination or preventing accumulation of contamination.
  • The DSS design includes adequate consideration of ALARA principles and is sufficient to facilitate compliance (by the general licensee) with applicable public and occupational exposure requirements of 10 CFR Part 20 and ALARA objectives.

Evaluation of the adequacy of radiation protection features necessarily includes consideration of dose rates and doses, in terms of both the public and workers. Sections 10B.5.2 and 10B.5.3 below address the evaluation of the doses. For purposes of this section, consider factors such as comparisons between the proposed DSS and DSSs that the NRC has already certified. While some allowance should be given for differences in the SNF contents and the design features, similarities in DSS dose rates, dose estimates for personnel, distances to meet 10 CFR 72.104(a) limits for sample DSS array sizes, and estimates of doses at the controlled area boundary for accident conditions can provide a good indicator about the adequacy of the DSS design in terms of radiation protection. If the values of the preceding items are significantly larger than those of currently certified systems, or if they seem to be large considering the proposed SNF contents compared with those of currently certified systems, consider whether the proposed design is sufficiently protective and seek further justification of the designs adequacy to protect personnel and the public.

For a DSS design that necessitates operations methods that are unusually different or are significant departures from those methods and descriptions that are common for certified systems in order to maintain personnel or public doses at reasonable levels, which are similar to those of the certified systems, consider seeking further justification regarding the designs adequacy. In some cases, design changes may be necessary so that the design will adequately protect personnel and the public.

Consider the regulatory limits and requirements in 10 CFR Part 20 and 10 CFR Part 72 beyond those directly applicable to CoCs to inform these kinds of evaluations. For these scenarios, consider the need for conditions in the CoC, in the technical specifications, regarding the DSSs design features or operational controls and methods, and coordinate the review with the technical specification reviewer (SRP Chapter 17). These conditions and technical specifications may include items such as clear descriptions of any extra shielding items as part of the DSS design and specifications (e.g., thicknesses), specifications of any remote operations equipment, requirements regarding use of these extra shield features and remote operations equipment, monitoring of operations and SSCs conditions, requirements for recovery actions for off-normal events, preoperational testing of any remote operations and equipment, limits on the duration of high dose-rate configurations, and added considerations for general licensees to address in their evaluations for using the DSS. 2 Also review the SAR chapter for operating procedures and coordinate with that reviewer (SRP Chapter 11) to evaluate the relevant aspects of the DSS design and operations that are covered in that chapter.

Evaluate the adequacy of ALARA considerations in the DSS design and operations. The regulatory guides cited in Section 10B.5.4 provide information that may be useful for this evaluation. Consider the physical design features together with the descriptions of DSS 2 The shielding design features are important for ensuring compliance with regulatory dose limits, including the limits in 10 CFR 72.104(a) and 10 CFR 72.106(b). See also Footnote 1 on page 10B-5.

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operations in the SAR. For example, designs that require unique methods of operation in order to protect personnel and the public may not have adequately incorporated ALARA. While doses during normal operations may be acceptable, the need to implement these unique operational methods and the equipment to facilitate them could introduce scenarios, including off-normal events, that typically do not need to be evaluated or considered but that could, for these designs, have potentially significant dose consequences to personnel and members of the public. Such a situation may not be consistent with ALARA principles and should be carefully evaluated. Flag and be attentive to whenever the SAR identifies the need for methods of operations or use of specialized equipment that is significantly or unusually atypical compared with operations that are common for existing, certified DSSs.

Coordinate the review with the technical specification reviewer (SRP Chapter 17) to ensure that the CoC, including the technical specifications, adequately describes the DSS, including any important features that the analysis relies on for demonstrating compliance with regulatory limits.

In addition to what is commonly considered part of the DSS, the CoC should include the description of (1) shield berms that support the results of dose analyses described in Section 10B.5.3; (2) significant shielding components required for personnel protection to enable personnel to perform operations on or around the DSS; and (3) parameters for ensuring that the shielding remains adequate for normal, off-normal, and accident conditions where normal activities (e.g., ISFSI expansion) could otherwise remove materials relied on for shielding.

Evaluate the applicants proposed operational controls to ensure that exposures and doses to workers and members of the public are controlled, within NRC dose limits, and consistent with ALARA objectives. Ensure that the technical specifications include any necessary operational controls and limits as discussed in Section 10B.4.1 above. Also ensure that the SAR descriptions of operating procedures include implementation of needed operational controls and limits.

Coordinate this effort with the operating procedures reviewer (SRP Chapter 11). Also coordinate with the shielding, confinement, and conduct of operations reviewers to ensure that the SAR includes acceptance tests and maintenance programs that are sufficient to ensure that the DSS will perform as designed in terms of radiation protection for the duration of its certified life and use.

10B.5.2 Occupational Exposures Ensure that the SAR includes occupational dose estimates for DSS operations as well as descriptions of the methods and parameters (e.g., inputs, assumptions) used to develop those estimates. Verify that the estimates and descriptions adequately address the operations sequences and considerations identified in Section 10B.4.2.

Review the SAR chapters that describe systems operations. Coordinate with the reviewers of these chapters to ensure that all descriptions are consistent with and adequately detailed to support the occupational dose estimates and the bases for those estimates. These descriptions should include the necessary actions and cautions to ensure operations are conducted in a manner that is consistent with the bases of the occupational dose estimates. In addition, ensure that dose estimates for periodic or routine maintenance as well as surveillance activities include reasonable assumptions regarding dose contributions from adjacent DSSs or the DSS array depending on the storage configuration and the expected personnel actions and positions the applicant described.

Coordinate with the shielding reviewer (SRP Chapter 6) to ensure that the SAR includes dose rates at adequate locations and numbers of locations around the DSS for all of the different configurations that arise during normal operations and anticipated occurrences for systems where 10B-10

such evaluations are needed (see Section 10B.4.2). Ensure that the SAR includes dose rates on and near DSS surfaces where personnel will be performing operations on or close to the DSS.

Verify that the SAR also includes dose rates at appropriate distances from the DSS for operations that involve personnel positioned at such distances from the DSS.

Verify that the applicant presents sufficient information in describing the methods, bases, and assumptions used for the dose assessment. This information should include the rationale used to justify the bases for various exposure times, personnel locations relative to the DSSs (including hot spots), number of personnel required for each operation, and appropriate gamma and neutron dose rates at all assumed locations. Verify that calculated doses and applied assumptions are consistent with these estimates and SAR descriptions of operating procedures. Also verify the reasonableness of these assumptions. Comparisons with other NRC-certified systems may provide useful insights for this evaluation. Confirm that the SAR provides dose estimates that consider all configurations that will occur during operations. The dose estimates should be refined adequately enough to appropriately capture the differences in personnel positions (e.g., personnel positioned at the canister lid vs. standing near the DSS base) and changes in DSS configurations (e.g., water present in the DSS canister vs. drained out of the canister). Regarding method, it may simply involve multiplying the dose rates (calculated in the shielding analysis) for different locations for each operation by the number of individuals and the time duration associated with that operation and summing the totals for each operation over each operation sequence. If a more complex method is chosen that involves computer codes (beyond that used for the dose rates in the shielding analysis), consult Chapter 6, Sections 6.4.4.1, Computer Codes, and 6.5.4.1, Computer Codes, of this SRP for applicable review guidance.

Determine the reasonableness of the estimated doses for the different operations. To do this, consider the doses estimated for other NRC-certified systems in line with the considerations described above in Section 10B.5.1 as well as consideration of implications for a licensees ability to meet 10 CFR Part 20 exposure and dose limits and requirements when using the DSS. In evaluating the estimated doses, keep in mind that a general licensee using the DSS will conduct DSS operations under the licensee's radiation protection program, which includes personnel dose monitoring to ensure compliance with 10 CFR Part 20 limits and any licensee administrative limits.

Regulatory Guide 8.34, Monitoring Criteria and Methods To Calculate Occupational Radiation Doses, which was developed to implement revisions to 10 CFR Part 20, contains information to consider in evaluating the acceptability of the applicants occupational exposure evaluation and monitoring recommendations.

10B5.3 Exposures at or Beyond the Controlled Area Boundary As described in Section 10B.4.3, demonstration of compliance with the requirements in 10 CFR 72.104 and 10 CFR 72.106 is ultimately the responsibility of the general licensee that uses the DSS because that demonstration considers factors that are specific to the licensees site.

However, 10 CFR 72.234(a) requires the CoC holder and applicant for a certificate to ensure the DSS design complies with 10 CFR 72.236, which includes sufficient shielding and confinement features to meet the requirements in 10 CFR 72.104 and 10 CFR 72.106 (see 10 CFR 72.236(d)).

Confirm that the applicant has provided analyses that are adequate to demonstrate that the DSS is sufficiently designed to meet these requirements in accordance with 10 CFR 72.236(d). 3 These analyses also facilitate the general licensees evaluations for its sites compliance with 10 CFR 72.104 and 10 CFR 72.106.

3 See Footnote 1 on page 10B-5.

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Ensure that the SAR adequately describes the methods, assumptions, and bases used in the analyses and that these methods, assumptions, and bases are appropriate for the analyses and the conditions being evaluated. The analyses should include the contributions both from direct radiation and from any effluents for the DSS, including, as appropriate, surface contamination (at the levels allowed by the technical specifications). Since the evaluation for surface contamination would be similar to that for effluents, coordinate with the confinement reviewer to evaluate any contamination contributions, which technical specifications limits should make negligible for offsite doses. The analyses should also account for all appropriate exposure pathways for effluents.

The SAR should include dose calculations for a single DSS and a theoretical array of DSSs, assuming design-basis source terms and full-time occupancy. Other aspects of the analyses are described in the sections that follow for the different conditions of operation. It should be noted that, because of the design requirements for DSSs, direct radiation is expected to be the major contributor to exposures and doses. Also, because of these design requirements, radioactive materials are not expected to be released from DSSs during normal, off-normal, or accident conditions. However, as noted elsewhere, the analyses will include effluent dose contributions for DSSs with specified, analyzed leak rates (i.e., the DSS is not designed and tested to be leak-tight or meet the noncredible leakage criteria).

Coordinate with the shielding, confinement, and accident evaluation reviewers to obtain the dose and dose rate results from those evaluations and to ensure that they are sufficient to support the evaluations in this part of the review. Also consider the results of staffs confirmatory analyses for those reviews in evaluating compliance with requirements in this review, particularly if those confirmatory analyses indicate significant differences in comparison with the applicants analyses.

If the confinement analysis only provides effluent dose results at 100 meters (328 feet) or only for a single DSS, coordinate with the confinement reviewer to evaluate how effluents may contribute to doses at additional distances and for DSS arrays in order to determine if additional analyses are needed in the SAR to address these scenarios. Also coordinate with other reviewers, such as the structural and shielding reviewers, to understand the impacts of different events (anticipated occurrences and accident conditions) to ensure the SAR analysis adequately addresses the dose impacts for all relevant events for all relevant DSS operating configurations.

10B.5.3.1 Normal Conditions and Anticipated Occurrences Ensure that the applicants evaluations for these conditions include analyses for a single DSS and for a sample array of DSSs on an ISFSI pad. These analyses have typically only considered the DSS in its storage configuration on the ISFSI pad. The NRC has accepted this practice for most systems because the other operation configurations (e.g., loading and transfer) are of very short duration so that dose contributions beyond the controlled area boundary are expected to be very small to negligible. Also, the limits in 10 CFR 72.104(a) include doses from both normal conditions and anticipated occurrences. For DSSs in their storage configuration, anticipated occurrences typically do not affect the DSSs. So, the dose rates and doses are not affected by anticipated occurrences. Anticipated occurrences have also not typically affected DSS dose rates for other operation configurations, and dose rates, though high in some cases, have not necessitated consideration of anticipated occurrences for those operations either.

The guidance in this section is generally based on these practices. However, be aware of instances where the impacts of anticipated occurrences or these other operation configurations should be considered in the analysis. These instances include DSSs where design features, dose rates, or operations methods for these other operations configurations are significantly different from those that are typical of certified DSSs. An example would be a DSS with significantly higher dose rates in a particular operation configuration that, if the DSS remained in this configuration for 10B-12

a reasonable duration (resulting from either normal conditions or an anticipated occurrence), could have a nonnegligible effect on doses beyond the controlled area. In such cases, ensure that the SAR adequately considers the impacts of these operations in demonstrating the adequacy of the DSSs shielding for meeting the limits in 10 CFR 72.104(a). 4 Also ensure that dose analyses for normal conditions and anticipated occurrences adequately consider variations in the storage configuration(s) that may occur for DSS designs where normal, though likely infrequent, actions may alter the DSSs shielding in its storage configuration. Such actions include construction activities associated with expansion of an operating ISFSI that removes material relied on for DSS shielding or otherwise exposes this material when it would not otherwise be exposed. In cases when their consideration is necessary, the applicants analysis should include the dose impacts from the bounding anticipated occurrence, assuming a reasonable event duration that includes the necessary time to recover from the event. Coordinate with the shielding, structural, and other relevant reviewers, as appropriate, in evaluating these scenarios.

Ensure that the SAR includes analyses for a single DSS and for a hypothetical array of DSSs.

The hypothetical array should consist of at least 20 DSSs in a 2 x 10 array configuration. The SAR analyses should include dose or dose rate versus distance curves to facilitate site-specific evaluations for general licensees. The NRC has accepted the use of dose (rate) versus distance curves for a single DSS and a DSS array as a means to demonstrate the DSS design is sufficient to meet the 10 CFR 72.104(a) limits.

Ensure that the applicants analyses assume appropriately bounding conditions. Such conditions include design-basis source terms, no intervening shielding between the DSS or DSS array and location of the dose receptor, and full-time yearly occupancy at each analyzed distance. Ensure the distances for which doses are provided include the doses at 100 meters (328 feet) from the single DSS and the DSS array since 100 meters is the minimum distance to the nearest ISFSI controlled area boundary, as noted in 10 CFR 72.106(b). Analyses that only include distances that are larger than 100 meters may be acceptable if the longer distance is made a condition of use in the CoC. In addition, ensure that the SAR determines the degree to which dose rates under normal conditions could change for other identified operating conditions and anticipated occurrences. For the array analyses, the applicant may account for shielding among DSSs, but should provide sufficient details regarding how that is done. Ensure that the analyses appropriately address this inter-DSS shielding when credited. If the analyses credited some engineered feature (e.g., a shield wall or berm), then ensure the CoC includes this feature as part of the DSS design and that the SAR includes appropriate descriptions and technical drawings for this feature.

Identify the minimum distance that the applicants analysis indicates is required to meet the dose regulation in 10 CFR 72.104(a) for both the single DSS and the array of DSSs. Past applications have shown this distance to be typically within 200 meters (656 feet) for a single DSS. Consider the minimum distance for the single DSS and the DSS array and evaluate whether the distance indicates that the DSS includes shielding and confinement features that are sufficient to meet the dose regulation in 10 CFR 72.104. Compare how the distances for this DSS compare with those of certified DSSs, accounting for the relevant considerations identified in Section 10B.5.1. Also consider, to the extent practical, typical general licensee site features. These may include typical distances to owner-controlled area boundaries, typical distances to locations of the public around the licensees site (e.g., residences, recreation areas), and residency times. Consider that a general licensee may store as much SNF as it generates at its 10 CFR Part 50 or 10 CFR Part 52 reactor facility. For DSSs for which significant distances are needed to meet the 4 See Footnote 1 on page 10B-5.

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10 CFR 72.104(a) limits, accounting for the considerations listed above, the SAR may need to include additional information to justify how the DSS design is sufficient to enable a general licensee to reasonably meet those limits. Typically, the DSS design and the SAR analyses do not include engineered features such as shield berms. However, the general licensee may choose to use such features at its ISFSI, which is permissible, to mitigate doses to individuals near the site.

Thus, verify that the CoC includes a condition to ensure that a general licensee that chooses to use these features will adequately manage these features. This condition should be similar to the following:

Supplemental Shielding: In cases where engineered features (e.g., earthen berms, shield walls) are used to ensure that the requirements of 10 CFR 72.104(a) are met, such features are to be considered important to safety, evaluated to determine the applicable quality assurance category, and appropriately evaluated under 10 CFR 72.212(b).

Be aware that the general licensee that uses the DSS must perform a written evaluation to demonstrate that the requirements in 10 CFR 72.104 are met, as required in 10 CFR 72.212(b)(5)(iii), for any real individual (as defined in Section 10B.4.3.1) located beyond the controlled area of the licensees site. The licensee may use information provided in the DSS SAR as well as site-specific information in performing this evaluation. Although evaluations the general licensee performs are not submitted to the NRC for approval, they are subject to NRC inspection and should be recorded and maintained by the general licensee.

The CoC should include a condition that ensures that a general licensee using the DSS will implement the necessary monitoring to ensure compliance with the limits in 10 CFR 72.104(a) during the operations of its ISFSI. By virtue of being a general licensee, the licensee already has programs including its radiological protection program and environmental monitoring program to meet its 10 CFR Part 50 or 10 CFR Part 52 license requirements, which may also be applied toward monitoring for compliance with 10 CFR 72.104. Thus, a CoC condition that directs the general licensee to update its radiological protection and environmental monitoring programs to incorporate its SNF operations and to address compliance with 10 CFR 72.104(a) limits may be sufficient. Consult the CoC conditions (likely in the CoC technical specifications) of currently certified DSSs in developing an appropriate monitoring condition for the DSS being reviewed.

The monitoring program should address both direct radiation and effluents, as appropriate, as well as have operating procedures to identify and reevaluate potential increases in exposures to individuals located beyond the sites controlled area.

As noted in Section 10B.5.1, the reviewer should consider the need to include operational controls and limits in the CoC conditions (in the technical specifications). As noted in Section 10B.4.1, controls and limits to be considered include dose rate limits and associated measurements. The determination for the need for these limits is discussed in those sections and is contingent upon a variety of factors. These factors include, but are not limited to, DSS dose rates for the different operations and configurations, the nature of the DSS design, potential dose impacts of changes to that design, and the need for such limits to ensure continued compliance with 10 CFR 72.236(d).

Such dose rate limits should be derived from the applicants dose rate and dose analyses for normal and off-normal conditions. The limits should be developed for appropriate DSS configurations and should be compared against the maximum measured dose rates. The dose rate limit condition should include an appropriate number of measurements at appropriate DSS surface locations to adequately ensure compliance with the limits.

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10B.5.3.2 Accident Conditions and Natural Phenomenon Events Ensure that doses are calculated for all relevant accident conditions for all relevant DSS configurations. Thus, the SAR should include accident condition doses for DSS configurations in addition to the final storage configuration, such as the loaded transfer cask for canister-based DSSs. Refer to the accident analysis evaluation chapter (SRP Chapter 16), which includes a list of accidents that are typically analyzed. Consider whether the DSS design may be susceptible to other types of accidents for which doses should be analyzed. Also consider whether the design introduces the possibility of other, atypical, configurations for which such accidents should be analyzed. For example, accident dose analyses should include doses for accidents that occur when material relied on for shielding may be removed or exposed under some normal, though temporary, operating conditions when that material otherwise would not be removed or exposed.

Typically, accident condition doses are analyzed for only a single DSS; however, consider whether there may be scenarios for the DSS design when an accident could affect the entire array. Ensure that the applicant analyzed doses for a DSS array in such a case.

Ensure that the applicants analyses assume appropriate bounding conditions. These conditions include assumptions such as no intervening shielding between the DSS and the individual at the analyzed distance(s), full-time occupancy at the analyzed distance(s), and a reasonably bounding duration of the event. The event duration should include the time to recover from the event and its impacts. A typically assumed duration is 30 days. The sum of the doses from each applicable contributing factor (direct radiation, effluents, surface contamination) should not exceed the limits in 10 CFR 72.106(b).

Verify that the applicant calculated doses at 100 meters (328 feet) from the DSS, the minimum distance allowed in regulations from the ISFSI storage and handling facilities to the nearest boundary of the controlled area. Applicants may calculate doses or dose rates at a discrete distance(s) or may develop a curve that shows dose versus distance. Ensure that the analysis shows that the DSS will not exceed the 10 CFR 72.106(b) dose limits at 100 meters from the DSS. For those DSSs for which a greater distance or engineered design features, such as berms, are needed to meet these dose limits, ensure that the CoC includes this distance or the engineered feature(s) as a condition of DSS use. Ensure also that the CoC includes the engineered feature(s) in the description of the DSS and that the SAR includes adequate descriptions of the engineered feature(s), including technical drawings.

10B.5.4 As Low As Is Reasonably Achievable Design Evaluate the DSS with regard to implementation of ALARA, both in the physical design features and descriptions of operating procedures. To perform this evaluation, consider the DSSs design features and the operating procedures described in the SAR. Ensure that the DSS design and operations address ALARA for both occupational and public exposures. Also consider how ALARA is incorporated into other NRC-approved DSSs, as appropriate, and the state of technology to inform the evaluation of the proposed DSS. Consider consulting available regulatory guides (e.g., Regulatory Guide 8.8, Information Relevant to Ensuring that Occupational Radiation Exposures at Nuclear Power Stations Will Be As Low As Is Reasonably Achievable) that contain information regarding ALARA that may also be useful to inform the review. Reference to these regulatory guides may be useful to inform the evaluation of the DSSs adequacy for meeting and for facilitating licensees compliance with regulatory requirements and ALARA objectives.

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Consider actions that general licensees would be reasonably expected to take to implement ALARA during DSS operations. Among others, these actions include the use of lead blankets, other types of temporary shielding, preoperation planning, prestaging of equipment, and preassembly of equipment and components. The DSS design and operations should ensure that these and other reasonable actions will be sufficient to ensure licensee compliance with ALARA requirements.

10B.5.4.1 Design Considerations Coordinate with the shielding and confinement reviewers to ensure that the DSS design features adequately incorporate ALARA principles to the extent practical as described in Section 10B.4.4.1 of this SRP. These principles should also be reflected in any design criteria the applicant described in the SAR to support the materials, geometric, and dimensional aspects of the DSS design. Ensure that the applicant has adequately justified that the proposed DSS design incorporates ALARA to the extent practical and necessary, or reasonable. Credit for incorporation of ALARA should be limited to features that are part of the DSS design that are adequately described in the SAR, including the technical drawings and schematics. Designs that necessitate operations that are atypical of approved DSSs in order to maintain reasonable personnel doses for normal operations or that could result in potentially significant exposures to personnel involved in actions to recover from an off-normal condition may not meet ALARA objectives. There may also be implications for public doses and ALARA considerations for those doses. In such cases, seek further justification from the applicant regarding the adequacy of ALARA incorporation into the DSS design and consider whether any CoC condition is needed in this regard.

10B.5.4.2 Procedures and Engineering Controls Confirm that the descriptions of proposed DSS operations adequately demonstrate implementation of ALARA principles into operating procedures as described in Section 10B.4.4.2 above. Confirm that the description of operating procedures includes necessary controls and actions to minimize dose and minimize contamination. Identify operations where elevated dose rates may occur, and ensure that operations descriptions include proper cautions and warnings and, where appropriate, personnel actions. Examples of these operations include those that necessitate personnel to perform actions near streaming paths or where radioactive particulates may be entrained in water draining from SSCs of the DSS. Some of the actions may include recommendations to use temporary, portable shielding such as lead blankets, recommendations on positioning of personnel involved in the procedures, or wetting the DSS surfaces exposed to SNF pool water to minimize adherence of radioactive particles (contamination control). Ensure that the proposed procedures and controls include those that are necessary for the DSS to meet 10 CFR 72.236(d) and support licensee compliance with 10 CFR 72.104(b), 10 CFR 72.104(c),

and 10 CFR 72.126(a) as well as relevant 10 CFR Part 20 requirements.

10B.6 Evaluation Findings The NRC reviewer should prepare evaluation findings upon satisfaction of compliance with the regulatory requirements in Section 10B.4, as determined through a review conducted in accordance with the information in this SRP chapter. Such a review includes coordination with other reviewers as described in the guidance in this chapter. If the documentation submitted with 10B-16

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

F10B.1 The [DSS designation, specify] provides radiation shielding and confinement features that are sufficient to meet the requirements of 10 CFR 72.104 and 10 CFR 72.106, in accordance with 10 CFR 72.236(d).

F10B.2 The design and operating procedures of the [DSS designation, specify]

provide acceptable means for controlling and limiting occupational radiation exposures within the limits given in 10 CFR Part 20 and for meeting the ALARA objective with respect to exposures, consistent with 10 CFR 20.1101(b).

F10B.3 The [DSS designation, specify] is adequately designed to facilitate decontamination in accordance with 10 CFR 72.236(i) and includes, to the extent practical and appropriate, adequate features, operating procedures, and controls that are designed to assist a general licensee to meet the radiological protection criteria in 10 CFR 72.126(a) and 10 CFR 72.126(d).

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

The staff finds, with reasonable assurance, that the radiation protection design of the

[DSS designation, specify] is in compliance with 10 CFR Part 72 and that the applicable design and acceptance criteria have been satisfied. The evaluation of the radiation protection design provides reasonable assurance that the [DSS designation, specify] will allow safe storage of SNF. The staff reached this finding based on a review that considered applicable NRC regulations and regulatory guides, codes and standards, accepted health physics practices, statements and representations contained in the SAR, and the staffs confirmatory analyses.

10B.7 References 10 CFR Part 20, Standards for Protection Against Radiation.

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.

Regulatory Guide 8.8, Information Relevant to Ensuring that Occupational Radiation Exposures at Nuclear Power Stations Will Be as Low as Is Reasonably Achievable.

Regulatory Guide 8.34, Monitoring Criteria and Methods To Calculate Occupational Radiation Doses.

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U.S. Nuclear Regulatory Commission (NRC), Expand Applicability of Part 72 to Holders of, and Applicants for, Certificates of Compliance, Federal Register, Vol. 64, No. 199, October 15, 1999, pp. 56114-56128.

NRC, Criticality Control of Fuel Within Dry Storage Casks or Transportation Packages in a Spent Fuel Pool, Federal Register, Vol. 71, No. 221, November 16, 2006, pp 66648-66657.

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