SECY-22-0062, Enclosure 3 - Final Rule Comment Response Document for NuScale Small Modular Reactor Design Certification

SECY-22-0062: Enclosure 3 - Final Rule Comment Response Document for NuScale Small Modular Reactor Design Certification

ML22004A007
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Issue date:	07/01/2022
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Malave, Yanely
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References
10 CFR Part 52, NRC-2017-0029, NuScale, RIN 3150-AJ98 SECY-22-0062
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NRC Responses to Public Comments Final Rule: NuScale Small Modular Reactor Design Certification NRC-2017-0029; RIN 3150-AJ98

U.S. Nuclear Regulatory Commission Office of Nuclear Material Safety and Safeguards Office of Nuclear Reactor Regulation

Month 2022

ABBREVIATIONS AND ACRONYMS

ADAMS Agencywide Documents Access and Management System CFR Code of Federal Regulations COL combined license DCA design certification application DCD design control document DCR design certification rule EA environmental assessment ECCS emergency core cooling system ER environmental report FR Federal Register FSER final safety evaluation report GDC general design criterion/criteria ITAAC inspections, tests, analyses, and acceptance criteria NRC U.S. Nuclear Regulatory Commission NuScale NuScale small modular reactor PRA probabilistic risk assessment SAMDA severe accident mitigation design alternative TMI Three Mile Island

ii U.S. NUCLEAR REGULATORY COMMISSION RESPONSE TO PUBLIC COMMENTS RECEIVED ON THE PROPOSED RULE NUSCALE SMALL MODULAR REACTOR DESIGN CERTIFICATION

Introduction

This document presents the U.S. Nuclear Regulatory Commissions (NRCs) responses to written public comments received on the proposed rule, NuScale Small Modular Reactor Design Certification (NuScale). The NRC published the proposed rule and notice of the proposed rule in the Federal Register on July 1, 2021 (86 FR 34999), for public comment with a 60-day public comment period. On August 24, 2021 (86 FR 47251), the NRC extended the public comment period by 45 days, resulting in a total comment period of 105 days.

The proposed rule on NuScale is available from the Federal e-Ru lemaking Web site at https://www.regulations.gov/ (Docket ID No. NRC-2017-0029) and through the NRCs Agencywide Documents Access and Management System (ADAMS) (Acce ssion No. ML21147A432).

In developing the final rule, the NRC considered all the commen ts provided in response to the proposed rule. If, as a result of its review of a public commen t, the NRC changed the rule text, the final rule preamble (also referred to as the statements of consideration), or the supporting documents, the NRCs response to the comment indicates where th e change occurred.

Overview of Public Comments

The NRC received comments from nine individuals and organizatio ns, as shown in Table 1. Of those comments, six were in favor of the design certification r ule (DCR), one was opposed, and the other two comments stated no preference for the outcome of the rule but included questions. One of the submissions was received after the close of the public comment period.

As stated in the proposed rule, comments received after the com ment close date are considered by the NRC when it is practical to do so; the NRC de termined it was practical to consider the late-filed comment submission.

The NRC reviewed and annotated the comment submissions to ident ify separate comments within each submission. Accordingly, a single submission may ha ve several individual comments associated with it. The NRC gave each individual comme nt within a submission a unique identifier. The NRCs summaries include this unique iden tifier to indicate which individual comments are addressed by each response. Public comment submiss ions are available online in the NRC Library at http://www.nrc.gov/reading-rm/adams.html. From this page, the public can access ADAMS, which supplies text and image files of the NRCs public documents. If you do not have access to ADAMS, or if there are problems in accessing the documents located in ADAMS, contact the NRCs Public Document Room at 1-800-397-4209, 301-415-4737, or by e-mail to pdr.resource@nrc.gov. In addition, public comments and supporting materials related to this final rule can be found at https://www.regulations.gov by searching for Docket ID NRC-2017-0029.

3 Table 1: Comment Submissions Comment Submission ID Commenter ADAMS Accession Number 1 Private Citizen, Keith Welch ML21189A248 2 Private Citizen, James A. Hoerner ML21189A249 3 Private Citizen, Diana Wulf ML21196A531 4 Nuclear Energy Institute ML21288A130 5 Union of Concerned Scientists ML21288A131 6 NuScale Power Inc. ML21288A189 7 The United Association of ML21288A273 Plumbers and Pipe Fitters and The Mechanical Contractors Association of America 8 The Breakthrough Institute ML21288A274 9 Private Citizen, Nick Wagner ML21288A275

Comment Categorization

This comment response document separates the comments into the nine categories identified below. Within each category, the NRC summarizes and responds to the comments. In general, the NRC addresses each individual comment. However, when simila r comments can be readily grouped together, the NRC has binned those comments and treated them as a single comment.

The agencys response addresses the binned comments. The annota ted comment number or numbers appear in parentheses at the end of each comment summar y to provide a cross-reference to aid the reader.

The comment summaries are grouped in the following categories:

A. General Comments on the Proposed Rulemaking B. Severe Accident Mitigation Design Alternatives C. Unresolved Technical Issues in the Design Certification Appl ication D. Departures, Changes, or Exemptions E. Gas Combustion F. Reference Corrections G. Inadvertent Actuation Block Valves H. Definitions I. Compliance

A. General Comments on the Proposed Rulemaking

Comment A-1: Two comments support the proposed rule. One comment states that the innovative design will also play an important role in providing relatively clean, safe, reliable, and cost-competitive base-load electricity and ensuring America rem ains a leader global nuclear technology. The second comment states that NuScale is designe d to provide a safer, more cost-effective clean option for meeting future energy needs and is particularly well suited to replacing aging U.S. coal plants. (2-1, 7-1)

4 NRC Response: The comments support the proposed rule and suggest no changes. No changes were made in response to these comments.

Comment A-2: The comment states that the review process was clear and well communicated in a manner that provides a high-level of public confidence. In addition, the comment states that lessons learned from the NRCs review of the NuScale design cer tification application (DCA) should be documented and disseminated for general knowledge and improvement of future DCA submissions and to assist combined license (COL) applicants to proceed more effectively with their applications. (8-1)

NRC Response: The NRC agrees with the comment. The Office of Nuclear Reactor Regulation (NRR) issued a report on lessons learned from NRRs review of t he NuScale DCA on March 20, 2022 (ADAMS Accession No. ML22088A160). No changes we re made in response to this comment.

Comment A-3: The comment states opposition to the design certification appr oval. In addition, the commenter requested the NRC to focus on water and abolish t he Price-Anderson Act. (3-1)

NRC Response: The NRC licenses and regulates the Nation's civilian use of rad ioactive materials to provide reasonable assurance of adequate protectio n of public health and safety and to promote the common defense and security and to protect t he environment. The comment is out of scope, and no changes were made in response to this c omment.

B. Severe Accident Mitigation Design Alternatives

Comment B-1: The comment states that the NRC failed to consider severe accid ent mitigation design alternatives (SAMDAs) associated with the potential for boron redistribution/dilution transients that could lead to core damage. Specifically, the co mment states that the NRCs environmental assessment (EA) referenced in the proposed rule f ails to evaluate potential SAMDAs that could reduce the risk of core damage and radiologic al release associated with boron redistribution events. The comment further states that a s the result of Chapter 15 deficiencies, the ECCS design is incomplete. In addition, the comment notes that the latest NuScale design changes have improved the boron mixing before th e emergency core cooling system (ECCS) actuation, but additional design modifications a re needed for NuScale to mitigate post ECCS actuation boron dilution and demonstrate tha t the system capabilities to bring the system back to normal with no adverse impacts on the core cooling. (5-1)

NRC Response: The NRC disagrees with this comment. As described in Final Sa fety Evaluation Report (FSER) Section 19.1.4.6.4, the NRC thoroughly reviewed the possible phenomena and processes that could lead to rapid flow incursion s that could lead to core damage. The NRC performed an independent evaluation (ADAMS Acce ssion Nos. ML20191A069 and ML20205L317) into the physical processes a ffecting the boron dilution and how those processes might impact the likelihood for core da mage to occur during postulated events. The NRC found in FSER Section 19.1.4.6.4 tha t the applicant adequately addressed the impact of the boron redistribution phenomena in t he DCA probabilistic risk assessment (PRA), and the PRA adequately reflects the design an d operation as described in the DCA. The NRC found reasonable assurance that there are no k nown significant risk

5 contributors that are unaccounted for and that the identified r isk insights are acceptable to support the Commissions objectives for use of PRA at the desig n stage. Because boron redistribution is unlikely to lead to core damage and is not a significant risk contributor, the NRC concludes that further consideration of a SAMDA is not warrante d. No changes were made in response to this comment.

Comment B-2: The comment states that the NuScale PRA has identified a cask d rop during refueling as the internal initiating event with the highest fre quency of core damage (on the order of 1x10-6/plant-year for a 12-module plant). Nevertheless, the NRC, desp ite being unable to reach a finding on SAMDAs associated with a cask drop during refueling (Release Category 8 in the NRC EA), approved the NuScale environmental report (ER) on the basis that any SAMDA addressing this risk would be associated only with improvements to the reactor building crane, which is not considered part of the design certification. The comment states that this is false because the crane has a critical function in the operation of t he plant and plays an outsized role in plant risk, and it is highly likely that other SAMDAs could be identified to help mitigate the risk of a cask drop. (5-2)

NRC Response: The NRC believes the comment is referring to events involving a dropped NuScale power module because Release Category 8 is associated w ith a dropped NuScale power module during refueling operations. A potential cause of a dropped NuScale power module could be failures associated with the reactor building c rane. With that understanding, the NRC disagrees with this comment.

However, based on this comment the NRC clarified in the final E A that the NRCs environmental evaluation included aspects of the reactor building crane becau se the crane was considered during the review of the DCA. The staff documented its review o f the reactor building crane in the following FSER Sections: 3.2.1, Seismic Classification; 3.7.3, Seismic Subsystem Analysis; 3.8.4, Seismic Category I Structures; 9.1.2, New and Spent Fuel Storage; 9.1.4, Light Load Handling System (Related to Refueling); 9.1.5, Ov erhead Heavy Load Handling System; 18.1, Human Factors Engineering Program Management; and 19.1.4.6.3, Reactor Building Crane Failure Resulting in Postulated Module Drop.

Based on its review, the NRC staff concluded that the reactor b uilding crane is single-failure proof consistent with cranes used in currently operating plants and the guidance for Type I cranes in NUREG-0554, Single-Failure-Proof Cranes for Nuclear Power Plants, issued May 1979 (ADAMS Accession No. ML110450636), and American Societ y of Mechanical Engineers (ASME)-NOG-1, Rules for Construction of Overhead and Gantry Cranes (Top Running Bridge, Multiple Girder). The design includes limit sw itches to protect against the reactor building crane experiencing overtravel, overspeed, over load, and unbalanced load events. The major risks for a dropped module are due to human e rrors of commission and failures of instrumentation. Two additional inspections, tests, analyses, and acceptance criteria (ITAAC) for the reactor building crane were added to Tier 1 of the design control document (DCD) for rated load tests of the module-lifting fixtures and m odule-lifting adapter and for inspection of the as-built module-lifting fixtures and module-l ifting adapter.

Also, as the staff noted in its FSER, a COL applicant that refe rences the NuScale design will describe the process for the handling and receipt of critical l oads, including NuScale Power

6 Modules, to satisfy COL Item 9.1-5. A licensee that references the NuScale design must satisfy all ITAAC, including those associated with the reactor building crane.

The NRC evaluated three SAMDAs related to the reactor building crane. As discussed in the draft EA, one SAMDA is related to automation of the power modul e transport process to reduce operator errors of commission, one SAMDA is related to providin g a railway system on the reactor pool floor to assist in transporting the power module t o the refueling area, and the final SAMDA is related to improving testing and maintenance procedure s for the reactor building crane. The SAMDA related to the addition of a railway system wa s eliminated because the cost exceeded the benefit. The two SAMDAs related to automation of t he power module transport process and improving testing and maintenance procedures are de pendent on site-specific information provided by a COL applicant referencing the NuScale design.

The final EA clarifies that the two SAMDAs to be analyzed by a COL applicant referencing the NuScale design are related to a design element to reduce human errors of commission. A COL applicant could consider additional design elements addressing human errors of commission in addition to training and procedures. Thus, the dropped module s evere accident risks and maximum benefit discussed in Revision 5 of the NuScale DCA ER ( ADAMS Accession No. ML20224A512) could change as a result of the COL applicants cl osure of the two SAMDAs and COL Item 9.1-5. Any COL applicant referencing the NuScale desig n would need to close the related COL items, assess the risks according to site-specific conditions regarding dropping a module during any moves, and address SAMDAs for reducing or avo iding adverse environmental effects in the COL application.

Finally, the NRC disagrees with the comment that the NRC appro ved the NuScale ER. The NRC does not approve an applicants ER; rather the agency prepa res an independent EA based on the applicants ER and other sources. The conclusion o f the NRCs EA is that the proposed action will not have a significant effect on the quali ty of the human environment. The proposed action is to certify the NuScale design in Appendix G to Title 10 of the Code of Federal Regulations (10 CFR) Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants.

No change to the rule was made in response to this comment.

Comment B-3: The comment states that site-specific SAMDAs, multi-unit aspe cts, procedural and training SAMDAs, and the reactor building crane design woul d need to be assessed when a specific site is proposed for constructing and operating a NuSc ale power plant. The term multi-unit in the context of a multi-module reactor design is ambigu ous, as each reactor module could be considered a unit. The EA considered multi-module aspe cts; it appears this phrase was meant to instead refer to multi-plant aspects (i.e., more t han one 12-module facility at a site). The comment suggests replacing the term multi-unit wit h multi-plant. (6-10)

NRC Response: The NRC agrees with this comment that the term multi-unit co uld be confusing for the NuScale design. The NRC staff addressed the d istinction between multi-module and multi-unit review issues in its response to NuScale Power dated October 25, 2016 (ADAMS Accession No. ML16229A522). In its response, the NRC sta ff referenced the definitions of Nuclear Power Unit and Modular Design found in 10 CFR 52.1, Definitions, as shown below:

7 Nuclear power unit. A nuclear power unit means a nuclear power reactor and associa ted equipment necessary for electric power generation and includes those structures, systems, and components required to provide reasonable assuranc e the facility can be operated without undue risk to health and safety of the public.

Modular design means a nuclear power station that consists of two or more esse ntially identical nuclear reactors (modules) and each module is a separ ate nuclear reactor capable of being operated independent of the state of completio n or operating condition of any other module co-located on the same site, even though th e nuclear power station may have some shared or common systems.

For the purposes of the NuScale design, each nuclear power modu le is a single unit, but because the design has been reviewed for up to 12 units in a si ngle reactor building, some multi-unit site issues have been reviewed and resolved for the NuScale design. Footnote 1 in the final EA clarifies that the SAMDA candidates for multi-uni t sites for NuScale are evaluated in the context of multiple 12 unit plants at the same site.

The NRC added a discussion of these issues to the preamble to t he final rule and revised the EA to clarify how multi-unit site considerations were handled. The NRC has also added to the final rule a definition of the term nuclear power unit as app lied to NuScale.

C. Unresolved Technical Issues in the Design Certification Appl ication

Comment C-1: The comment states that for the plant-specific DCD, a COL appl icant may also have to include considerations for multi-module facilities in t he plant-specific DCD that were not previously evaluated as part of the design certification rule. The comment states that it is unclear what the NRC intends by this statement because the NuSc ale final safety analysis report is based on a 12-module plant. The comment proposes clar ification or deletion of the statement that a COL applicant may need to address additional m ulti-module considerations in the plant-specific DCD. (6-8)

NRC Response: The NRC confirms that it evaluated the NuScale power plant, in cluding up to 12 modules and the associated balance-of-plant support systems and structures. Accordingly,Section V.B of the preamble to the final rule reads, A COL app licant will also have to include considerations for a multi-unit site in the plant-specific DCD that were not previously evaluated as part of the design certification rule, e.g., construction im pacts on operating units. For example, an applicant proposing to add modules to an operating NuScale power plant would need to address the potential impacts on the operating modules from the addition of the new modules that were not reviewed as part of the DCA..

8 Comment C-2: The comment states that the NRC identified the following issues as unresolved open items in the DCA: shielding wall design, containment leaka ge from the combustible gas monitoring system, and steam gener ator stability during density wave oscillations. The comment also states that these unresolved issues create regulatory unce rtainty for COL applicants. The comment proposes that the industry make the outstanding issues generic to allow effective resolution by the research community. In addition, the comment states that the NRC should clarify what the potential outstanding multi-module considerati ons and provide guidance on how they may be resolved. (8-3)

NRC Response: The NRC takes no position on this comment to the extent the com ment recommends that the industry consider genericizing the unresolv ed issues. The NRC disagrees with the comment requesting clarification on how to resolve out standing issues because this discussion is referring to issues that may arise from site-or application-specific considerations that were not evaluated in the design certification process (e. g., later addition of modules to an operating NuScale power plant), and as such, the NRC cannot det ermine prospectively what the issues might be. The preamble to the final rule clarifies that the NuScale design is certified for up to 12 modules in a single reactor building (i.e., multi-modu le considerations for construction and operation of up to 12 modules in a single reactor building), and no changes were made in response to this comment.

D. Departures, Changes, or Exemptions

Comment D-1: The comment states that although 10 CFR 50.109(a) applies to st andard design approvals, it does not apply to design certifications. The NRC has addressed this issue in Section VIII.C.1 of the final rule, which provides that the bac kfitting requirements in 10 CFR 50.109 apply to changes to NuScale design certification generic technical specifications and other operational requirements that were completely reviewe d and approved in the design certification rules (and do not require a change to a design fe ature in the generic DCD). The change processes described in Section VIII.C.1 are specific to the NuScale design certification rule. (4-1)

NRC Response: The NRC agrees that Section VIII.C.1 directs the NRC to apply t he requirements in 10 CFR 50.109 when making a change to generic t echnical specifications or other operational requirements that were completely reviewed an d approved in the design certification rulemaking and that did not require a change to a design feature in the generic DCD. The NRC further agrees that Section VIII.C.1 of Appendix G applies only to the NuScale design. Each design certification rule has an equivalent provi sion that applies only to the specific design certification rule of which it is a part. No ch anges were made as a result of this comment.

Comment D-2: The comment states that an interpretation of Sections VI.C, VI II.C.1, and VIII.C.4 of the final rule that would withhold issue resolution but grant backfit protecti on and require exemptions for departures from unresolved matters seems inconsistent. The comment cites minimum operator staffing and containment leakage rate te sting as examples of operational requirements that were completely reviewed and appr oved in the NRC staffs FSER.

The comment states that the NRC should clarify or revisit, on a generic basis, the portion of Appendix G,Section VI.C, that states [t]he Commission does no t consider operational requirements for an applicant or licensee who references this a ppendix to be matters resolved

9 within the meaning of § 52.63(a)(5). This statement appears to be in tension with Appendix G,Section VIII.C.1, which applies the requirements of 10 CFR 50.1 09 to certain operational requirements that were comple tely reviewed and approved by the design certification rulemaking. The comment states that this is resolved by reading Section VI.C to apply to operational requirements that were not completely reviewed and approved as part of the design certification rulemaking. Issue resolution should be afforded w hen the NRC has completed its safety review and the public was afforded the opportunity to co mment. If the NRC disagrees, it should revisit these provisions for operational requirements on a generic basis and in a manner that does not impact the NuScale design certification rule sche dule. (4-2)

NRC Response: The NRC agrees with this comment in part and disagrees in part. Operational requirements and design information are afforded finality by di fferent provisions of the NRCs regulations. Among other things, an application for design cert ification may be requested only for essentially complete designs (see Sections 10 CFR 52.41, Scope of subpart, and 10 CFR 52.47(c)(1)-(2)), which ensures that applicants provide sufficient information to allow the NRC to make comprehensive findings on the design. As Sectio n VI.A of Appendix G states,

[a] conclusion that a matter is resolved includes the finding that additional or alternative structures, systems, and components, design features, design cr iteria, testing, analyses, acceptance criteria, or justifications are not necessary for Nu Scale. In contrast, operational requirements are generally afforded finality through 10 CFR 50. 109, Backfitting; for example, changes to the generic minimum staffing requirements in 10 CFR 50.54(m) would be subject to the requirements in 10 CFR 50.109. Therefore, when these provis ions for operational requirements were first used, in the Advanced Boiling Water Rea ctor design certification rule, the Commission determined the following (62 FR 25805; May 12, 1 997):

The Commission does not support extension of § 52.63 to technical specifications and other operational requirements as requested by NEI, rather the Commission supports the proposal to treat the technical spe cifications in Chapter 16 of the DCD as a special category of information. T he purpose of design certification is to review and approve design informatio n. There is no provision in Subpart B of 10 CFR Part 52 for review and approva l of purely operational matters. After the COL is issued, the set of tech nical specifications for the COL (the combination of plant-specific and DCD derived) would be subject to the backfit provisions in § 50.109 (assuming no Tier 1 or Tier 2 changes are involved).

Thus, the generic technical specifications and operational requ irements that have been completely reviewed and approved during the design certificatio n rulemaking process are afforded finality by Section VIII.C of Appendix G. However, Sec tion VI.C properly provides that operational requirements, even those completely reviewed and ap proved in the design certification rulemaking, are not subject to the issue finality requirements in 10 CFR 52.63, Finality of standard design certifications, because those req uirements are intended for design information, which must be essentially complete, whereas operat ional requirements do not need to be essentially complete because they are not the primary sub ject of design certification.

However, the NRC recognizes that operational requirements can b e affected by aspects of the design and are appropriate for review in the design certificati on process. The NRC therefore affords completely reviewed and approved operational requiremen ts the finality established through 10 CFR 50.109 by Section VIII.C, as discussed in respon se to Comment D-1.

10 The NRC agrees that the design-specific operator staffing requi rements in Appendix G are the kind of completely reviewed and approved operational requiremen ts that are the subject of Section VIII.C. However, although NuScales exemption request f or Type A containment leakage rate testing was reviewed in the FSER, the proposed rul e did not include an exemption from Type A testing for licensees that reference Appendix G, an d therefore, the NRC disagrees with the comment that, as proposed, the containment leakage rat e testing matter was a completely reviewed and approved operational requirement. Never theless, in the final rule, the NRC has included provisions for both operator staffing and Type A testing, and thus both provisions have been completely reviewed and approved in the co urse of the design certification rulemaking process and will be subject to the requirements of 1 0 CFR 50.109 as stated in Section VIII.C.

It is important to note that these are design-specific regulati ons (i.e., rules that apply only to licensees that reference Appendix G). The NRC can generically a ddress any need for design-specific alternatives to NRC regulations that apply only during operation, but must do so through rulemaking, as has been done in Appendix G for operator staffing and in the final rule for Type A testing provisions. Rulemaking for such matters is n ecessary because Section VIII.C cannot relieve future licensees that reference Appendix G from complying with applicable regulations.Section VIII.C applies without additional rule pro visions only when, for example, the NRC has concluded that an operational method or approach discus sed in the DCA does or does not meet an NRC requirement. A conclusion that alternative regulations would be appropriate for a particular design is not, therefore, a comple tely reviewed and approved operational requirement subject to Section VIII.C. Thus, had the NRC not included design-specific regulations in the final rule for minimum staffing and Type A testing, these matters would not be approved within the meaning of Section VIII.C.

No changes were made as a result of this comment.

Comment D-3: The comment states that the 10 CFR 50.54(m) exemption is listed among exemptions for the NuScale design, which is not applicable to the design, but rather to a licensee referencing the design certification. In addition, the comment states that Section V.D may also warrant a brief discussion of this new approach to exe mptions from 10 CFR 50.54(m) and 10 CFR Part 50, Domestic Licensing of Production and Utili zation Facilities, Appendix J, Primary Reactor Containment Leakage Testing for Water-Cooled P ower Reactors, for licensees referencing the NuScale design certification. The com ment proposes the creation of a new Section V.C in the rule to list and clarify the exemptions for licensees referencing the NuScale design certification, such as 10 CFR 50.54(m) and the A ppendix J exemption. (6-14)

NRC Response: The NRC agrees that the preamble and rule text should address the design exemptions separately from the design-specific regulations for licensees referencing the NuScale DCR. Therefore, a new section was created under Section IV, Additional Requirements and Restrictions, to list the exemptions applicab le to future applicants and licensees referencing the design certification (i.e., alternati ve staffing requirements and the Appendix J Type A testing exemption). In addition, the alternat ive staffing requirements of the proposed rule were removed from paragraph V.B and moved to the new Section IV.C in the rule text.

Comment D-4: The proposed rule states that in proposing a contention on comp liance with the Tier 2 departure provisions as part of an ongoing adjudicatory proceeding, the intervenor must

11 demonstrate that the change stands on an asserted noncompliance with an ITAAC acceptance criterion. The comment states that it is unclear what it mean s for a change to stand on an asserted ITAAC noncompliance and that previous design certifica tion rules have used the term bears on, which appears correct in this context. The comment s proposed resolution is to replace the term change with departure to enhance clarity a nd revise the provision to state, Further, the petition must demonstrate that the departure bear s on an asserted noncompliance with an ITAAC acceptance criterion in the case of a § 52.103 pr eoperational hearing. (6-18)

NRC Response: The NRC agrees with the comment. Previous design certifications have used the term bears on; therefore, revising the sentence in paragr aph VIII.B.5.g would increase clarity and consistency among 10 CFR Part 52 appendices. The re vised sentence in paragraph VIII.B.5.g of the rule generally will read as follows :

Further, the petition must demonstrate that the change bears on an asserted noncompliance with an ITAAC acceptance criterion in the case of a §52.103 preoperational hearing, or that the departure bears directly on the amendment request in the case of a hearing on a license amendment.

Comment D-5: The comment states that the preamble and proposed rule do not address the inapplicability of 10 CFR Part 50, Appendix J, to a licensee re ferencing the NuScale design.

NuScale DCA Part 7, Section 7, sought an exemption from 10 CFR Part 50, Appendix A, General Design Criteria for Nuclear Power Plants, General Des ign Criterion (GDC) 52, Capability for containment leakage rate testing, for the NuSc ale design and an exemption from Appendix J Type A testing for licensees referencing the NuScale design. The comment states that the staffs FSER, Section 6.2.6.4, approved both exemption requests and that neither the preamble nor the applicable regulations portion of the propos ed rule discusses the exemption for licensees from the Type A testing requirements of Appendix J. The comment proposes the addition of Appendix J Type A testing to the list of exemptions granted by the final rule. (6-7)

NRC Response: The NRC agrees with the comment that the exemption to Appendix J Type A testing should be identified in the rule and preamble to the fi nal rule. FSER Section 6.2.6.4, Technical Evaluation for Exemption Request No. 7, issued July 2020 (ADAMS Accession No. ML20205L406), documents the NRC staffs review of the reque st to exempt licenses referencing the NuScale design from 10 CFR Part 50, Appendix J, Type A tests. Therefore, the new Section IV.C of the final rule and Section IV.H of the prea mble to the final rule will include Appendix J to 10 CFR Part 50 for Type A testing for licenses re ferencing the NuScale design.

Comment D-6: The comment states that current regulations are written in a sp ecific, prescriptive manner, which is based on large light-water reacto r operational experience and incidents. As an example, 10 CFR Part 50 provides control room staffing requirements based on a set of assumptions applicable to large light-water reactors. This will require COL applicants that seek to deploy the NuScale reactor to seek exemptions if t hey wish to use the number of operators recommended by NuScale.

The comment states that the prescriptive nature of the regulati ons also required NuScale to seek exemptions from a standard, rather than simply describing how the NuScale design meets safety objectives. The comment recommends allowing use of the I mplementation of the Proposed Risk-Informed Technology Inclusive Regulatory Framewor k Approach for COL

12 Applicants Referencing the NuScale DCA because it includes elem ents that would improve regulatory certainty for COL applicants. Specifically, the fram ework includes provisions for performance-based demonstrations that would enhance the ability of COL applicants to demonstrate that safety objectives have been met, without seeki ng exemptions. (8-2)

NRC Response: The NRC disagrees with this comment. The final rule provides al ternative staffing requirements that will be applicable to any licensee o perating a NuScale power plant under Appendix G to 10 CFR Part 52. The NRC is currently develo ping a rule and guidance for implementing a technology-inclus ive regulatory framework. However, work on developing the new framework is ongoing and not available yet for use. No cha nges were made as a result of this comment.

E. Gas Combustion

Comment E-1: The comment notes that the preamble states that the combustibl e gas monitoring leakage issue may be resolved by performing radiati on dose calculations and demonstrating that doses would remain within applicable dose li mits in 10 CFR part 20. As the preceding sentence notes, this issue does not affect norma l plant operation or non-core damage accidents. The dose limits of 10 CFR Part 20, Standard s for Protection against Radiation, apply to normal plant operations. The comment state s that the staffs FSER, Section 12.3.4.1.3, invokes the control room habitability asses sment of 10 CFR 50.34(f)(2)(xxviii) and the important area access requ irement of 10 CFR 50.34(f)(2)(vii) as relevant to the potential onsite dos es associated with this core damage accident-related release; it also cites the accident dos e limits of 10 CFR 52.47(a)(2)(iv) as applicable to offsite doses. The comment proposes the deleti on of the reference to 10 CFR Part 20 as an applicable requirement for a COL applicant to resolve the combustible gas monitoring leakage issue. (6-5)

NRC Response: The NRC disagrees with the comment. The dose limits of 10 CFR P art 20 apply at all times, not only to normal operating conditions. To resolve the combustible gas monitoring leakage issue, the COL applicant will need to ensure that post-accident leakage from these systems does not result in the total main control room do se exceeding the dose criteria (i.e., 50 millisieverts (5 rem)) for the surrogate event with s ignificant core damage or include design features in accordance with 10 CFR 50.34(f)(2)(xxvi) and (f)(2)(xxviii) to ensure that the dose criteria are not exceeded, or both. To demonstrate that th ese requirements are met, the COL applicant can submit an analysis showing the 10 CFR Part 20 limits are not exceeded. No changes were made in response to this comment.

Comment E-2: The comment states that the Federal Register notice summarizes the NRCs position that there was insufficient information available reg arding NuScale combustible gas monitoring system and the potential for leakage from this syste m outside containment. The NRC was unable to determine whether this leakage could impact analyses performed to assess main control room dose consequences, offsite dose conseq uences to members of the public, and whether this system can be safely re-isolated.... The comment states that the NRC conclusions are mistaken because this issue comes down to leaka ge from a system (combustible gas monitoring), which is addressed in 10 CFR 50.4 4, Combustible gas control for nuclear power reactors, for the express purpose of monitoring combustible gases in a beyond-design-basis core damage event. The comment also states that this type of

13 beyond-design-basis event is not required to meet the offsite d ose criteria of 10 CFR 52.47(a)(2)(iv). No design-basis event in the NuScale de sign damages fuel cladding, let alone causes severe core damage. The NRC seems to be mixing the design-basis offsite dose requirementwhich includes a h ypothetical major fission product release inside containment postulated only for that purposewith the functional requiremen t for combustible gas monitoring under real (although extremely unlikely) core damage scenarios, which are beyond design basis.

In addition, the comment states that the Three Mile Island (TMI ) rules do address beyond-design-basis accidents, but they do so by requiring addi tional functions to help mitigate those events, not by imposing dose limits. NuScale addressed th ese rules in its Lessons-Learned from the Design Certification Review of the Nu Scale Power, LLC Small Modular Reactor, dated February 19, 2021 (ADAMS Accession No. ML21050A431). As noted in that report, 10 CFR 50.34(f)(xxvi) does not apply a dose lim it for leakage control, but just requires that leakage be as low as practical. The regulation in 10 CFR 50.34(f)(2)(vii) is explicit (see NUREG-0737, Clarification of TMI Action Plan Requirements, issued November 1980, ADAMS Accession No. ML051400209) that it does not address leaka ge from systems outside containment, because those systems already have leakage as low as practical under 10 CFR 50.34(f)(2)(xxvi). Lastly, 10 CFR 50.34(f)(2)(xxviii) do es not require the control room habitability to address new beyond-design-basis events; instead, it requires licensees to re-verify their control rooms for the DCA Chapter 15 events.

The comment states that the NRC seems to be combining the combu stible gas monitoring requirement with other unrelated rules to yield a result that, for the first time, applies dose criteria to beyond-design-basis events. This is akin to requiri ng a plant to analyze doses for a station blackout or anticipated transient without scram event. The comment states that the NRCs issue not resolved positi on is correct, and this would set a bad precedent for future applicants. (9-1, 9-2, 9-3)

NRC Response: The NRC disagrees with this comment. To perform hydrogen monito ring following a significant accident in the NuScale design, manual actions outside of the control room may be required.

The regulation in 10 CFR 50.34(f)(2)(vii) requires, in part, th at licensees perform radiation and shielding design reviews of spaces around systems that may, as a result of an accident, contain accident source term and design as necessary to permit adequate access to important areas.

The TMI requirement includes a footnote specifying that the fis sion product release should be based on a major accident that is hypothesized for purposes of site analysis, or postulated from considerations of possible accidental events, that would result in potential hazards not exceeded by those from any accident considered credible. It als o indicates that such accidents have generally been assumed to result in substantial meltdown o f the core with subsequent release of appreciable quantities of fission products. In the N uScale design, the site analysis includes an assumed maximum hypothetical accident that includes core damage.

NUREG-0737 specifies the 5-rem whole body dose limit for meetin g 10 CFR 50.34(f)(2)(vii).

Therefore, the NRCs position is that a radiation and shielding design review is required for the actions that may be necessary to perform hydrogen monitoring.

Regarding offsite dose, 10 CFR 52.47(a)(2)(iv) includes a footn ote similar to the TMI item, and

14 this requirement is assessed against an assumed core damage acc ident. Since the hydrogen monitoring system may be in use following a severe accident and since it is uncertain that the system can be safely reisolated, the NRCs position is that leakage from the hydrogen monitoring loop must be shown not to result in doses in excess of the regulatory limits. This is similar to how engineered safety feature system leakage is cons idered in the offsite dose analysis in other reactor designs.

As a result, no changes were made in response to this comment.

F. Reference Corrections

Comment F-1: Preamble Section IV.F identifies FSER Sections 12.2, 12.3, 3.1 1, and 15.0.3 as discussing TR-0915-17565, Revision 3, Accident Source Term Met hodology, dated April 2019 (ML19112A172). The staffs FSER Section 15.0.2 also discusses that report. The comment proposes that FSER Section 15.0.2 be added to the list of sections that discuss TR-0915-17565. (6-2)

NRC Response: The NRC agrees with this comment and Section IV.F of the pream ble to the final rule generally will read as follows:

The NRCs review and findings of topical report TR-0915-17565, Revision 3, are documented in the topical report final safety evaluation report issued on October 29, 2019 (ADAMS Accession No. ML19297G520). The approved versio n TR-0915-17565-NP-A, Revision 4 (ADAMS Accession No. ML20057G132) i s discussed in the DCA safety evaluation report Section 12.2, Ra diation Sources, Section 12.3, Radiation Protection Design Features, Section 3.11 Environmental Qualification of Mechanical and Electrical Equip ment, Section 15.0.2, Review of Transient and Accident Analysis Methods, an d Section 15.0.3, Radiological Consequences of Design Basis Accidents.

Comment F-2: Preamble Section IV.A identifies FSER Chapter 3 as Design of S tructures, Components, Equipment, and Systems. The title of FSER Chapter 3 is Design of Structures, Systems, Components, and Equipment. The comment proposes corre ction of the title of FSER Chapter 3. (6-3)

NRC Response: The NRC agrees with this comment, and the FSER title in Secti on IV.A of the preamble to the final rule will be changed to FSER Chapter 3, Design of Structures, Systems, Components and Equipment.

Comment F-3: Preamble Section IV.A states, With the exception of the steam generator tube and inlet flow restrictor issue discussed previously. Identif ying that previous discussion asSection III.C.3 would increase clarity. The comment proposes re placement of previously with in Section III.C.3. (6-4)

NRC Response: The NRC agrees with this comment and will update the last sent ence in Section IV.A in the preamble to the final rule to state, With the exception of the steam generator tube and inlet flow restrictor issue discussed in Section III.C.3, the NRC found the comprehensive vibration assessment program adequate to ensure t he structural integrity of the NuScale power module components.

15 Comment F-4: Preamble Sections V.B, V.F, and V.H state that the generic tech nical specifications for the NuScale design are in Chapter 16 of the generic DCD. Chapter 16 of the NuScale final safety analysis report describes the process for developing the technical specifications, but the generic technical specifications are fo und in Part 4 of the DCA. The comment proposes that references to Chapter 16 of the DCD be ch anged to instead refer to DCA Part 4. (6-9)

NRC Response: The NRC agrees with the comment and deleted references to Chap ter 16.

Therefore,Section V.B. and V.H of the preamble to the final ru le generally will read as follows:

Section V.B The NRC is treating the technical specifications in Part 4, Ge neric Technical Specifications, of the DCA as a special category of informatio n and designating them as generic technical specifications in order to facilitate the spe cial treatment of this information under appendix G to 10 CFR part 52.

Section V.H The process in paragraph VIII.C.1 for making generic changes to the generic technical specifications or other operational requirements in the generic DCD is accomplished by rulemaking and governed by the backfit standards in § 50.109.

Comment F-5: The comment states that the proposed 10 CFR Part 52, Appendix G, Section VI.B, list of matters resolved does not include referen ced information in public documents. Nuclear safety and safeguards issues associated with referenced information intended as requirements in nonpublic reports are explicitly re solved, but not safety issues in public reports. Several of the reports referenced in the generi c DCD are exclusively public reports, with no equivalent nonpublic report that would be with in the scope of issue resolution.

While issue resolution for the FSER, Tier 2, and the rulemaking record implies resolution of referenced public reports, the design certification for the eco nomic simplified boiling-water reactor, Appendix E (ESBWR DC) to 10 CFR Part 52, includes the 20 documents approved for incorporation by reference by the Director of the Office of the Federal Register (i.e., the public documents) within the scope of Issue Resolution paragraph B.1. A clearer approach for the NuScale design certification may be to revise paragraph B.2 or include a new paragraph. The comment proposes the revision of the issue resolution provision s to include nuclear safety issues associated with referenced information in public documen ts which, in context, are intended as requirements in the generic DCD for the NuScale des ign. (6-16)

NRC Response: The NRC disagrees with this comment. The design certification i ncorporates by reference all documents that are necessary to meet the appli cation content requirements in 10 CFR 52.47(a)-(c) (except for conceptual design information a nd the ER), and for which either the NRC or the design certification applicant would like to establish finality. Therefore, no additional clarification is warranted, and no changes were made to the rule language in response to this comment.

Comment F-6: The comment states that the proposed rule provides a 15-year du ration from October 29, 2021. Other proposed design certification rules (a side from the direct final rule approach for the Advanced Power Reactor (APR) 1400) have includ ed a placeholder for the

16 final rule effective date; NuScale wants to call attention to t his to ensure that the final rule includes the correct duration start date. The comment proposes the revision of the duration provision to begin with the effective date of the final rule. ( 6-17)

NRC Response: The NRC agrees with the comment, and an effective date of 30 days after publication of the rule in the Federal Register will be provided in the preamble to the final rule.

G. Inadvertent Actuation Block Valves Comment G-1: The comment strongly agrees with the NRC staffs recommendati on in SECY-19-0036 [, Application of the Single Failure Criterion to NuScale Power LLC's Inadvertent Actuation Block Valves, dated April 11, 2019 (ADAM S Accession No. ML19060A081),] and Commissioner Barans dissenting vote to reje ct NuScales assertion that the critically important inadver tent actuation blocks, which mu st close rapidly and fully seal to prevent premature opening of the main ECCS valve should be reg arded as passive components that are not subject to the single-failure criterion. The comment states the Commissions majority vote to accept NuScales illogical conte ntion is irresponsible, dangerous, violates common sense, and should be overturned in t he final rule. (5-3)

NRC Response: The Commissions direction to the staff does not regard, redes ignate, or reclassify the inadvertent actuation block valves as passive co mponents. Rather, the Commission narrowly directed the staff not to apply the single-failure criterion only to the inadvertent actuation block valve closing function. The decisio n in the July 2, 2019, staff requirements memorandum to SECY-19-0036 (ADAMS Accession No. ML 19183A408) was not changed in response to this comment. No changes were made to th e rule text in response to this comment.

Comment G-2: Preamble Section IV.C states that the inadvertent actuation bl ock valve is safety-significant. In this context, safety significant is undefined and creates ambiguity. The comment states that NuScale has not undertaken risk-informed ca tegorization of structures, systems, and components pursuant to 10 CFR 50.69, Risk-informe d categorization and treatment of structures, systems and components for nuclear pow er reactors, which categorizes SSCs by their safety significance. Risk insights in dicate that the inadvertent actuation block is not risk significant. The comment proposes that the phrase safety-significant be deleted from the preamble. (6-6)

NRC Response: The NRC glossary defines the term safety-significant as foll ows:

used to qualify an object, such as a system, structure, compon ent, or accident sequence, this term identifies that object as having an impact on safety, whether determined through risk analysis or other means, that exceeds a predetermined significance criterion.

For the NuScale design, the NRC characterized the inadvertent a ctuation block valves as safety-significant because of the important role they play in ensuring that the fuel integrity and containment barriers remain intact and because they are necessa ry for satisfying safety requirements such as 10 CFR Part 50, Appendix A, GDC 10, 35, an d 38. The use of the term safety-significant in the preamble is consistent with its use in the FSER and SECY-19-0036.

No changes were made to the rule text in response to this comme nt.

17 H. Definitions Comment H-1: The comment states that the generic DCD is defined as the docu ment containing Tier 1 information, Tier 2 information, and generic technical specifications. This definition may cause confusion because the NuScale DCA does not include a discrete document containing that information; the generic technical spe cifications are in Part 4 of the DCA. The comment proposes revision of the final definition to r ead...means the Tier 1 and Tier 2 information (including the technical and topical reports referenced in Chapter 1) and generic technical specifications that are incorporated by refer ence into this appendix. (6-11)

NRC Response: The NRC agrees that the DCD is not contained in a single docume nt but disagrees that the term generic DCD refers only to a single d ocument containing Tier 1 information, Tier 2 information, and generic technical specific ations. The generic DCD as a whole is a singular official record that NuScale Power, LLC, as the design certification applicant, is required to maintain. The NRC notes that the same definition has been used for other design certification rules for which the generic DCD comprises multipl e documents but agrees that the definition can be clarified.

In the final rule, the definition of generic DCD reads as follo ws:

Generic design control document (generic DCD) means the documents containing the Tier 1 and Tier 2 information (including the tec hnical and topical reports referenced in Chapter 1) and generic technical specific ations that are incorporated by reference into this appendix.

The NRC does not intend this varia tion from other design certification rules to indicate a substantive difference from those other design certification ru les, but merely clarifies that this DCD comprises multiple documents.

Comment H-2: The comment states that the plant-specific DCD definition is de fined to include plant-specific changes to generic DCD information. Under desi gn certification rule nomenclature, changes are generic, while departures are pla nt-specific. The comment proposes that changes be replaced with departures. (6-12)

NRC Response: The NRC disagrees with the comment. The qualifier plant-specif ic clarifies that the changes are, indeed, plant specific (i.e., departures). The term departure is generally defined as plant-specific changes, so the language plant-spe cific change is equivalent to departure, but the term plant-specific change is clearer in this context because it contrasts with generic. No changes were made in response to this comme nt.

I. Compliance Comment I-1: The comment states that all nuclear power plants should incorpo rate onsite backup generators. The comment also states that there is no rea son (other than cost) not to equip these reactors with backup generation. Even if the plant itself can withstand the accident conditions, why force operators to deal with an accident in the absence of onsite power? The midst of a serious problem is not the time to be managing basic issues such as power. While

18 utilities could install backup power if they wished, the only w ay to ensure that they do so is to require it. (1-1)

NRC Response: The NRC agrees with the comment that, generally, nuclear power plants should have onsite backup generators. The NuScale onsite electr ical power system includes a backup power supply system consisting of backup diesel generato rs, as stated in DCD Part 2, Tier 2, Section 8.1, and the NRCs safety evaluation report in Section 8.1.1, Introduction. No changes were made in response to this comment.

Comment I-2: The comment notes that the proposed rule states that the requi rements of 10 CFR Part 20 have not been demonstrated with respect to steam generator tube integrity. The radiation protection standards of 10 CFR Part 20 pertain to dos es to plant workers and members of the public as a result of expected plant operations. Failure of steam generator tubes is an accident condition, as noted in the preamble (The failure of multiple steam generator tubes resulting from failure of an inlet flow restric tor has not been included within the scope of the NuScale accident analyses in DCA Part 2, Tier 2, C hapter 15.). The comment proposes the deletion of references to 10 CFR Part 20 requireme nts with respect to steam generator integrity. (6-1)

NRC Response: The NRC disagrees in part and agrees in part. The 10 CFR Part 20 dose limits apply at all times, including during beyond-design-basis accide nts. However, the comment is correct that demonstration of compliance with 10 CFR Part 20 do se limits is not required for design certification rule applications. Instead, 10 CFR 52.47(a )(2)(iv) requires applicants to show in the safety analysis that the dose will not exceed certa in criteria for this accident. The NRC agrees that a license application will need to meet the sta ndard in 10 CFR 52.47(a)(2)(iv),

rather than demonstrating that the 10 CFR Part 20 limits will n ot be exceeded. Therefore,Section III.C.3 of the preamble and final rule paragraph IV.A.2.i do not refer to 10 CFR Part 20, and the applicable dose criteria regulation was added to read a s follows:

Preamble,Section III.C.3 Therefore, the NRC concludes that NuScale Power has not demonst rated compliance with 10 CFR 52.47(a)(2)(iv) and appendix A to 10 CFR part 50, General Design Criterion (GDC) 4 and GDC 31, relative to potent ial impacts on steam generator tube integrity from inlet flow restrictor failu re

Rule, Paragraph IV.A.2.i Information demonstrating that the requirements of 10 CFR 52.47 (a)(2)(iv) and General Design Criterion (GDC) 4 and GDC 31 of appendix A to 10 CFR part 50 are met with respect to the structural and leakage integrity of the steam generator tubes that might be compromised by effects from densi ty wave oscillations in the secondary fluid system

Comment I-3: The comment noted that the proposed 10 CFR Part 52, Appendix G,

Section VI.B.1.d, states that GDC 10, Reactor design, applies to the steam generator integrity issue, implying that the COLA must demonstrate conformance with GDC 10 to resolve the staffs concerns. Two other provisions of the proposed rule add ressing steam generator integrity do not cite GDC 10. As GDC 10 concerns the reactor design, it i s not relevant to steam

19 generator integrity and is not cited by the FSER in this respec t. The comment proposes the deletion of references to GDC 10 with respect to steam generato r integrity. (6-15)

NRC Response: The NRC agrees with the comment because GDC 10 was erroneously listed and is not applicable to the steam generator integrity issue. T herefore,Section VI.B.1.d of the final rule reads: consistent with the other design informatio n regarding steam generator integrity described in DCA Part 2, Tier 2, Sections 3.9.1, 3.9. 2, 5.4.1, and 15.6.3, and in accordance with 10 CFR part 50, GDC 4 and 31.

Comment I-4: The comment states that the proposed 10 CFR Part 52, Appendix G,

Section IV.A.2.g, would require the COL applicant to include sh ielding design information to meet the radiation zones specified in DCA Part 2, Tier 2, Figur e 12.3-1. This requirement effectively controls that Tier 2 radiation zone map equivalentl y to Tier 1 information, because a COL applicant would have no ability to depart from the radiatio n zone map without first getting an exemption from this requirement. In other words, if a COL ap plicant were to depart from the radiation zone map in a manner otherwise acceptable under the T ier 2 departure provisions (because it meets the 10 CFR 50.59-like criteria), the applican t would still need an exemption from this provision because it would not provide shielding sati sfying the generic DCDs radiation zone map.

The comment states that this is an unnecessary new control on T ier 2 information. The regulatory history of Tier 1, standardization, and the change c ontrol provisions do not support an exemption requirement for this radiation zone map. The comment also notes that the radiation zone map, while supporting the operational dose limits and equi pment qualification, does not rise to the level of a fundamental basis for the staffs review and is not essential to standardization of the plant design, and thus departures from t he map do not justify an exemption requirement.

The comment states that the COL applicant can adequately addres s the NRCs expectation to address shielding of major penetrations by providing the shield ing details necessary to meet the radiation zones specified in the applicants plant-specific DCD ; the applicant then maintains the ability to depart from the generic DCD radiation zone maps to t he same extent they otherwise would be able to if the shielding details were provided in the generic DCD. The comment proposes changing this provision to refer to the plant-specific DCD radiation zone map instead of the DCA radiation zone map. (6-13)

NRC Response: The NRC disagrees with the comment. The radiation shielding be tween the power module bays and steam gallery areas minimizes radiation s treaming from the reactor power modules to the steam galleries and other outside areas. T he shielding is important not only for controlling radiation exposure to individuals but is a lso credited in the environmental qualification analysis. While these shield walls include large penetrations, NuScale did not analyze the radiation streaming through the penetrations and in dicated that the penetration shielding design had not been finalized and would be completed in a future phase of the design.

NuScale indicated that this would be the responsibility of the COL applicant.

The rule provision requires that a COL applicant provides penet ration shielding information demonstrating that shielding is provided to limit dose equivale nt to those values specified in the DCA. If the COL applicants approach is approved and the COL is issued, the COL applicant will

20 not be required to maintain doses to the radiation zone maps af ter COL issuance. If the penetration shielding is inadequate to limit doses to those spe cified in DCA Part 2, Tier 2, Figure 12.3-1, then different aspects of the radiation protecti on design would be unresolved, including findings related to 10 CFR Part 20 and environmental qualification compliance. For these reasons, it is appropriate to reference the DCA radiation zone map in the provision. The NRC did not change the rule language in response to this commen t.

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