ML17325A616

Public Comments Resolution Table DG-1330

ML17325A616
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Issue date:	04/30/2018
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RG-1.232, Rev. 0 DG-1330
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Response to Public Comments on Draft Regulatory Guide DG-1330, Guidance for Developing Principal Design Criteria for Non-Light Water Reactors Proposed Revision 0 of Regulatory Guide 1.232 On April 5, 2017, the NRC published in the Federal Register (82 FR 16636) that Draft Regulatory Guide, DG-1330 (Proposed Revision 0 of RG 1.2332), was available for public comment. The public comment period ended on April 20, 2017. The NRC received comments from the organizations listed below. The NRC has combined the comments and NRC staff responses in the following table.

Comments were received from the following:

Michael D. Tschiltz, Director Edward Burns New Plant, SMRs & Advanced Reactors 7701 Greenbelt Road, Suite 320 Nuclear Energy Institute (NEI) Greenbelt, MD 20770 1776 I Street NW, Suite 400, Washington DC 20006 eburns@x-energy.com ADAMS Accession No. ML17101A543 ADAMS Accession No. ML17097A241 Mark Holbrook Peter Smith Idaho National Laboratory 815 Bruce Street, Marine City, MI (DOE/LAB) Pwsmith11416@gmail.com ADAMS Accession No.ML17097A240 ADAMs Accession No. ML17096a266 Michael Keller Hans Gougar, Co-National Technical Director 14713 Woodyard Advanced Reactor Technologies Overland Park, KS 66223 Idaho National Laboratory m.keller@hybridpower.com ADAMS Accession No. ML17097A242 ADAMS Accession No. ML17086A110 John Kirby Anonymous kirbyjp@hotmail.com ADAMS Accession No. ML17097A238 ADAMS Accession No. ML1748A162 Herbert Burke, Energize Northwest (NNW) Tanju Sofu 1206 S. Cushman Avenue (Tsofu@anl.gov)

Tacoma, WA ADAMS Accession No. ML17255A220 (Herb@energizeenw.com)

ADAMS Accession No. ML17101A542 Page 1 of 105

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NRC should clarify the language NRC does not agree with this comment.

throughout the document regarding the NRC believes that the DG clearly states regulatory basis for Principal Design that it is guidance for non-LWR Criteria and the use of the regulatory applicants and designers, and is not a guide once issued. Principal Design regulatory requirement for advanced Criteria (PDCs) are required to be reactors. NRC does understand the included in an application for construction industrys point regarding the first permit, design certification, combined sentence in the purpose section of the license, design approval, or DG and proposes to change it to:

manufacturing license. (see 10 CFR This regulatory guide (RG) describes 50.35, the NRCs proposed guidance on how 52.47, 52.79, 52.137, and 52.157). the general design criteria (GDC) in 10CFR50 Appendix A States: Appendix A, General Design Criteria The principal design criteria for Nuclear Power Plants, of Title 10 of establish the necessary design, the Code of Federal Regulations, Part fabrication, construction, testing, 50 Domestic Licensing of Production 1 Industry/NEI General and performance requirements for and Utilization Facilities (10 CFR Part structures, systems, and 50) (Ref. 1) may be adapted for non-components important to safety; light water reactor (non-LWR) designs.

that is, structures, systems, and In addition, a similar modification was components that provide made to the Role of the General reasonable assurance that the Design Criteria for Non-LWRs, section facility can be operated without of the Discussion in the RG. No other undue risk to the health and safety changes were made in response to this of the public. These General comment.

Design Criteria establish minimum requirements for the principal design criteria for water-cooled nuclear power plants similar in design and location to plants for which construction permits have been issued by the Commission.

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The General Design Criteria are also considered to be generally applicable to other types of nuclear power units and are intended to provide guidance in establishing the principal design criteria for such other units.

It is industrys position, based on the above, that the GDCs of Appendix A do not establish regulatory requirements for use with non-LWR designs but provide guidance in developing and submitting PDCs with an application. Industry believes that this RG document will essentially replace Appendix A of 10 CFR Part 50 as guidance for advanced reactors in developing PDC to be included with an application. There are a number of statements in the draft guidance document that appear to presume the GDC in Appendix A are regulatory requirements for advanced reactors. For example, the Purpose Section of the DG states, this regulatory guide (RG) describes the NRCs proposed guidance on how the general design criteria (GDC) in Appendix A..apply to non-light water reactor (non- LWR) designs. Industry believes it is unnecessary and inappropriate to attempt to make the GDC of Appendix A apply to non- LWRs through this guidance document but rather to simply Page 3 of 105

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state the objective as guidance to an applicant develop PDCs as is done in the second sentence of the section. This is also consistent with the section entitled Intended Use of This Regulatory Guide in Section C.

There are a number of other places in the DG that imply conformance or alignment with Appendix A. It is recommended that a search for reference to Appendix A be performed and language appropriately clarified.

Suggested Change Clearly state that the objective is to provide guidance to an applicant develop PDCs and not to meet the GDCs as they are regulatory requirements for non-LWR reactors. This should be clear and consistent through- out the document.

For example the purpose section should state:

This regulatory guide (RG) describes the NRCs proposed guidance on how the general design criteria (GDC) in Appendix A, General Design Criteria for Nuclear Power Plants, of Title 10 of the Code of Federal Regulations, Part 50 Domestic Licensing of Production and Utilization Facilities (10 CFR Part

50) (Ref. 1) apply to non-light water reactor (non-LWR) designs. This guidance may be used by non-LWR Page 4 of 105

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reactor designers, applicants, and licensees to may develop principal design criteria (PDC) for any non-LWR designs, as required by the applicable NRC regulations. LWR general design criteria (GDC) in Appendix A, General Design Criteria for Nuclear Power Plants, of Title 10 of the Code of Federal Regulations, Part 50 Domestic Licensing of Production and Utilization Facilities (10 CFR Part 50)

(Ref. 1) are intended to only provide guidance to non-LWR designs. This RG derives Advanced Reactor Design Criteria (ARDC) from the intent of the GDC to provide more specific guidance.

The RG also derives additional design-specific criteria describes the NRCs proposed guidance for modifying and supplementing the GDC to develop PDC that address two specific non-LWR design concepts: sodium-cooled fast reactors (SFRs), and modular high temperature gas-cooled reactors (MHTGRs). PDCs for other designs can be developed using the more generic ARDC with design-appropriate changes.

Security Design Considerations Security for SMRs and non-LWRs is an As acknowledged in the preliminary draft ongoing activity and it is not appropriate 2 Industry/NEI General guidance on non-light water reactor to include a discussion in the RG. The security design (83 FR 13511; March 13, NRCs effort to incorporate security 2017), the Commissions Policy design considerations has been put on Page 5 of 105

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Statement on the Regulation of Advanced hold and the NRC is instead focusing Reactors, (73 FR 60612; October 14, on whether consequence-based 2008) states security requirements should be that the design of advanced reactors developed for small modular reactors should include considerations for safety (SMRs) and non-LWRs.

and security requirements together in the design process such that security issues (e.g., newly identified threats of terrorist attacks) can be effectively resolved through facility design and engineered security features, and formulation of mitigation measures, with reduced reliance on human actions. NRC goes on to observe that, as we have previously commented, design considerations and associated regulatory requirements related to security are currently addressed outside of 10 CFR 50 Appendix A. We appreciate the staffs attention to distinguishing security design considerations from general design criteria. This structure should be maintained, and design considerations related to security should not be incorporated into the advanced reactor design criteria.

Suggested Change Without incorporating security design considerations in the advanced reactor design criteria, add a brief discussion of the relationship and expectations for Page 6 of 105

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security in design, i.e., advanced reactor design criteria and security design considerations should be addressed by advanced non-light water reactor developers in parallel.

Security Design Considerations Security for SMRs and non-LWRs is an The degree to which integration of safety ongoing activity and it is not appropriate and security design requirements remains to include a discussion in the RG. The a challenge for designers of advanced NRCs effort to incorporate security reactors. Guidance on this will be design considerations has been put on 3 X-Energy General beneficial. hold and the NRC is instead focusing on whether consequence-based Suggested Change security requirements should be None provided. developed for small modular reactors (SMRs) and non-LWRs.

Discussion, Harmonization with NRC staff agrees with this comment.

International Standards, Page 10 Reference to the IAEA Coordinated IAEA is also developing safety design Research Activity on MHTGR safety criteria and safety design guidelines for design criteria was added to this 4 Industry/NEI General MHTGRs. section.

Suggested Change NRC should coordinate with MHTGR activities at IAEA in addition to SFRs.

Page 9 - Discussion, Key Assumptions NRC does not agree with this comment.

and Clarifications Regarding the non- The Key Assumptions provide insight LWR Design Criteria into what the staff did and did not The draft regulatory guide states: consider while developing the DG. The 5 Industry/NEI General It is the responsibility of the purpose of including this statement in applicant to demonstrate the RG is to ensure that applicants are compliance with applicable severe aware that severe accidents and accident and BDBE regulations BDBEs should be considered in the and orders, demonstrate why any Page 7 of 105

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that are not applicable do not design of the plants, even though there apply, and demonstrate why other are not explicit ARDCs for them.

design specific severe accidents or BDBE that can occur will be mitigated.

Since ARDC/SFR-DC/MHTGR-DC apply to normal, AOOs, and design-basis events, and do not pertain to BDBE regulations, this sentence is outside the scope of this report.

Suggested Change It is recommended that this key assumption be deleted.

Page 9 - Discussion, Key Assumptions The NRC staff agrees with this and Clarifications Regarding the non- comment. This change was LWR Design Criteria incorporated.

The seventh bullet states: The NRC intends the ARDC to apply to the six advanced reactor technology types identified in the DOE report; however, in some instances, the SFR-DC or MHTGR-DC may be more applicable to a design 6 Industry/NEI General or technology than the ARDC.

Clarification would be useful that a mix and match approach is entirely appropriate - i.e., an entire set of criteria for a given design wont necessarily apply.

Suggested Change Change to: The NRC intends the ARDC to apply to the six advanced reactor Page 8 of 105

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technology types identified in the DOE report; however, in some instances, one or more of the criteria from the SFR-DC or MHTGR-DC may be more applicable to a design or technology than the ARDC.

Page 9 - Discussion, Key Assumptions NRC staff agrees with this comment.

and Clarifications Regarding the non- Sentence was modified to remove, are LWR Design Criteria intended to apply to all designs of these The eighth bullet states, in part: The technologies. The phrase, were SFR-DC and MHTGR-DC are intended to developed because the designs were apply to all designs of these mature and the design features diverse technologies, which could leave the for these technologies, was added to impression that the criteria in the RG are clarify why the SFR and MHTGR DCs 7 Industry/NEI General inviolate, irrespective of specific design were developed.

attributes.

Suggested Change Caveat with a statement indicating that, as with all criteria, design-specific exceptions may be proposed (and defended) by the applicant.

Page 3- Communications, and Policy The NRC staff agrees with this Statements comment. Change was incorporated The draft regulatory guide includes the following citation in its Related Guidance, Communications, and Policy Statements listing: NRC, Next Generation Nuclear 8 DOE/Lab General Plant - Assessment of Key Licensing Issues, dated July 17, 2014, provides the NRC staffs review and insights on the Next Generation Nuclear Plant MHTGR design (Ref. 11).

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Suggested Change The NGNP interactions did not include NRC review of a specific modular HTGR design, but rather a series of proposals to address policy and key technical issues associated with MHTGR technology. The word design should be deleted and replaced with proposed licensing approach.

Page 6- Role of GDC in Regulatory The NRC staff agrees with this Framework comment. Change was incorporated The draft regulatory guide states: The GDC are also intended to provide guidance in establishing the PDC for non-LWRs. The GDC serve as the fundamental criteria for the NRC staff when reviewing the SSCs that make up a nuclear power plant design particularly when assessing the performance of their safety functions in design basis events 9 DOE/Lab General postulated to occur during normal operations, anticipated operational occurrences (AOOs), and postulated accidents.

Suggested Change Our understanding is that SSC safety functions are only relied on during plant response to postulated accidents. This sentence, which also refers to normal operations and AOOs, should be revised to more clearly reflect this. A suggested Page 10 of 105

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revision is to change safety functions to intended functions.

Page 7- Role of GDC for Non-LWRs The NRC staff agrees with this The draft regulatory guide states: comment. Sentence was changed to, Together, these requirements recognize Together, these requirements that different requirements may be recognize that different requirements necessary for non-LWR designs. may need to be adapted for non-LWR designs.

10 DOE/Lab General Suggested Change Based on the generally applicable statement from Appendix A in the previous paragraph, requirements should be revised to adapted requirements.

Page 7- Role of GDC for Non-LWRs The paragraph continues on, It is the The draft regulatory guide states: The applicants responsibility to develop the non-LWR design criteria developed by the PDC for its facility based on the NRC staff and included in Appendices A specifics of its unique design, using the to C of this regulatory guide, are intended GDC, non-LWR design criteria, or other to provide stakeholders with insight into design criteria as the foundation.

the staffs views on how the GDC could Further, the applicant is responsible for be interpreted to address non-LWR considering public safety matters and design features; however, these are not fundamental concepts, such as defense 11 DOE/Lab General considered to be final or binding in depth, in the design of their specific regarding what may eventually be facility and for identifying and satisfying required from a non-LWR applicant. necessary safety requirements.

This additional information explains Suggested Change what is meant by, what may This statement is not adequately clear eventually be required from a non-LWR and predictable for industry. The staff applicant.

appears to be saying that the guidance in this draft regulatory guide may not be the complete list of design requirements that Page 11 of 105

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apply. However, the last phrase of the cited text implies that the items being addressed in the draft regulatory guide may be incomplete and not a fully acceptable approach for developing the associated principal design criteria. It is recommended that the phrase however, these are not considered to be final or binding regarding what may eventually be required from a non-LWR applicant be deleted.

Page 7- Role of GDC for Non-LWRs The NRC staff agrees with this The draft regulatory guide states: The comment. Change was incorporated.

NRC recognizes the benefits to risk informing the non LWR design criteria to the extent possible, depending on the design information and data available.

12 DOE/Lab General Suggested Change Suggest changing benefits to future benefits to make it clear that this initial set has not been risk-informed beyond the general consideration of risk consistent with the LWR-based GDCs in Appendix A.

Much work has been undertaken to risk- The NRC staff does not agree with this inform the regulatory requirements and comment. This is noted in the guidance for large LWRs. More work will paragraph about the Vision and be needed for advanced non-LWRs. As Strategy: Safely Achieving Effective and 13 X-Energy General DG-1330 is finalized, a statement needs Efficient Non-Light-Water Reactor to be included that acknowledges the Mission Readiness, page 7, maturity of these efforts and the Implementing the mid- and long-term expectation for future enhancements. Implementation Action Plans as part of the Vision and Strategy activities will Page 12 of 105

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help NRC determine whether risk informed non-LWR design criteria should be included as part of a new regulatory framework.

The guidance that results from this DG-1330 effort should be noted as subject to further refinements as advanced non- The RG is a document that can be LWR designs are brought into the revised on an as-needed basis. This is marketplace. repeated multiple times in the guide and the office instructions for developing Suggested Change RGs. There is no need to add another None provided. statement to this effect in the final regulatory guide.

Page 8 - DOE-NRC Initiative Phase 1 The NRC staff agrees. Change was The draft regulatory guide states: The incorporated.

ARDC are intended to be technology neutral and, therefore, could apply to any type of non LWR design.

Suggested Change 14 DOE/Lab General A better term would be technology inclusive to align with the list of six technologies above, and to exclude LWRs. The DOE proposal was based on the six advanced reactor technologies summarized in the previous paragraph, and not any type.

Page 9 - Key Assumptions The NRC staff does not agree with this The draft regulatory guide states: It is the comment. Although beyond design responsibility of the applicant to basis events are not part of the scope of 15 DOE/Lab General demonstrate compliance with applicable the Reg. Guide, applicants must still severe accident and BDBE regulations address them. This is consistent with and orders, demonstrate why any that are Page 13 of 105

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not applicable do not apply, and other statements in the RG that mention demonstrate why other design specific additional considerations for applicants.

severe accidents or BDBE that can occur will be mitigated.

Suggested Change Since ARDC/SFR-DC/MHTGR-DC apply to normal, AOOs, and design basis events, and do not pertain to BDBE regulations, this sentence is outside the scope of this report. It is recommended that this key assumption be deleted.

Page 9 - Key Assumptions The NRC staff does not agree with The draft regulatory guide states: While deleting the key assumption.

developing the non-LWR design criteria, the staff assumed that a core disruptive ASLBP Issuance LBP-84-4 (ADAMS accident will be demonstrated to be a Accession No. ML16357A782) severe accident or a BDBE by the describes the manner in which a core applicant. disruptive accident should not be considered a design basis accident. As Suggested Change part of the review of the Clinch River This text implies that non-LWR designs Breeder Reactor (CRBR) plants 16 DOE/Lab General must designed for a core disruptive Construction Permit, the applicant accident that is a deterministic holdover evaluated the potential failures of from the past that current risk-informed equipment and potential accidents design approaches will likely eliminate which could cause fuel disruption. The from consideration. For some applicant also evaluated the technologies, the terms severe accident consequences of the disrupted core.

or core disruptive accident are not The analyses demonstrated that there technically meaningful. A goal of non- existed sufficient defense-in-depth in LWR designs would be to eliminate core the CRBR plant design and that the disruptive accidents from consideration by consequences of core disruption was reducing their likelihood to less than the acceptable relative to the risk of a LWR Page 14 of 105

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lower frequency threshold for beyond design. As discussed in the ASLPB design basis events. It is recommended memorandum, the core disruptive that this key assumption be deleted. accident was a potential consequence of a design basis accident only if multiple defense-in-depth components failed for the CRBR plant design.

The key assumption explicitly denotes that the all advanced reactors will need to demonstrate sufficient protection from and/or low consequence of core disruptive accidents in order to demonstrate that a core disruptive accident is not a probable event during a design basis accident.

If a core disruptive accident is a probable outcome of a design basis accident, the applicant may need to develop specialized PDCs which would ensure that the reactivity control systems and heat removal systems are sufficient considering the unknown core configuration. PDC changes may include requiring the secondary shutdown mechanism to be safety-related and may require additional detail on effective cooling.

Page 9 - Key Assumptions The NRC staff agrees with this The draft regulatory guide states: Safety comment. Change was incorporated.

design objectives for non-LWRs can differ (replaced objectives with approach).

17 DOE/Lab General substantially from those associated with This was included as an assumption LWRs. because staff considered this when developing the non-LWR design criteria.

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Suggested Change The statement is correct (replace objectives with approach) but its not clear why it is listed as an assumption.

Page 9 - Key Assumptions The NRC does not understand this The draft regulatory guide states: comment because it does not specify Proposed GDC adaptations were where such changes are needed or in focused on those needed for improved what context.

18 DOE/Lab General regulatory certainty and clarity.

Suggested Change This is the better choice of language -

NRC should use adaptation throughout.

Page 9 - Key Assumptions The NRC staff agrees with this Currently, the following items are located comment. Change was incorporated.

in the text of the NRC rationales:

• Prior to issuing this regulatory guide as final, it appears that Commission agreement will be needed on the functional containment performance requirements for the MHTGR.

19 DOE/Lab General

• In addition, staff acceptance of the SARRDL will also be needed.

Suggested Change It seems reasonable to state these in the assumptions to highlight that there are key policy items discussed in the regulatory guide that are still unresolved.

Page 10 - Harmonization with The NRC staff agrees with this International Standards comment. The NRC will follow its 20 DOE/Lab General The draft regulatory guide states: The standard procedures for public NRC will continue to monitor and participation in the development of Page 16 of 105

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collaborate on these documents and future NRC documents that reference or consider using them to the extent endorse international standards.

practical in developing SFR design criteria.

The last sentence states that NRC will consider use of international standards.

Will the US industry get to review and comment on these international standards-based criteria?

Suggested Change The last sentence states that NRC will consider use of international standards.

Will the US industry get to review and comment on these international standards-based criteria?

Page 10 - Harmonization with The NRC staff agrees with this International Standards comment. The Coordinated Research Its not clear why this section is included, Activity on MHTGR safety design and if its retained, why it doesnt include criteria was added.

other international efforts, such as the IAEA CRP on safety design criteria for 21 DOE/Lab General MHTGRs.

Suggested Change Include other international efforts, such as the IAEA CRP on safety design criteria for MHTGRs.

Page 10 - Harmonization with The NRC staff agrees with this International Standards comment. Change was incorporated.

22 DOE/Lab General The draft regulatory guide states: The International Atomic Energy Agency (IAEA), in collaboration with the Page 17 of 105

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International Project on Innovative Nuclear Reactors and Fuel Cycles and the Generation IV International Forum, established the Sodium-Cooled Fast Reactor Task Force.

Suggested Change This last paragraph focuses solely on the SFR. There is a similar activity underway for modular HTGRs that should be cited.

Page 11 - Intended Use of this The NRC staff agrees with this Regulatory Guide comment. Change was incorporated.

The draft regulatory guide states: For example, FHRs are liquid-metal reactors that use tristructural isotropic (TRISO) fuel, which is the same fuel used for 23 DOE/Lab General MHTGR technologies.

Suggested Change FHRs are not liquid-metal reactors. FHRs are a type of molten-salt cooled high-temperature reactors that use a fixed core rather than liquid fuel.

Page 11 - Intended Use of this The NRC staff agrees with this Regulatory Guide comment. Change was incorporated.

The draft regulatory guide states:

Applicants may use this RG to develop 24 DOE/Lab General all or part of the PDC and are free to choose among the ARDC, SFR-DC, or MHTGR-DC to develop each PDC.

Suggested Change Page 18 of 105

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Should add something like after considering the underlying safety basis for the criterion and evaluating the rationale for the adaptation described in this Reg. Guide to the end of this sentence.

Page 11 - Intended Use of this NRC staff does not agree with this Regulatory Guide comment. A change was incorporated The draft regulatory guide states: Finally, in the location specified in comment no.

the non-LWR design criteria as developed 24, but was not repeated in this second by the NRC staff are intended to provide location because it would be repetitive stakeholders with insights into the staffs and therefore unnecessary views on how the GDC could be interpreted to address non-LWR design features; however, these are not considered to be final or binding on what 25 DOE/Lab General may eventually be required from a non-LWR applicant.

Suggested Change Should add something like after considering the underlying safety basis for the criteria and evaluating the rationale for the adaptation described in this Reg. Guide to the end of this sentence.

Intended Use of this Regulatory Guide The NRC staff does not agree with this It is unclear to me why "applicants would comment. An exemption is not needed not need to request an exemption from since the GDC in 10 CFR 50 Appendix 26 Peter Smith General the GDC in 10 CFR Part 50 when A are not requirements for non-LWRs.

proposing PDC for a specific design." Is it They are considered to be generally the intention of the Staff that the RG applicable to non-LWRs and are Page 19 of 105

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represents an interpretation of how the intended to provide guidance in GDC can be satisfied? establishing the principal design criteria.

Suggested Change None provided Intended Use of this Regulatory Guide The NRC staff does not agree with this What is the legal basis for materially comment. The GDC in 10 CFR 50 altering Appendix A to 10CFR50 using a Appendix A are not requirements for low tier regulatory guidance document? non-LWRs. They are considered to be Specifically, I am referring to the generally applicable to non-LWRs and 27 Michael Keller General exemptions proposed for gas reactor (m- are intended to provide guidance in HTGR) - e.g. removing the requirements establishing the principal design criteria.

for a containment.

Suggested Change None provided Page 14 - Table 1, Multiple Barriers The NRC agrees with this comment.

The draft regulatory guide states: Change was incorporated.

MHTGR-DC 18 - Same as GDC 28 DOE/Lab General Suggested Change Should say Same as ARDC Page 22 - Acronyms The NRC staff does not agree with this The draft regulatory guide states: comment. See resolution to comment SARRDL - specified acceptable system No. 43 radionuclide release design limit 29 DOE/Lab General Suggested Change Not what was proposed; should be specified acceptable core radionuclide release design limit. The detailed basis for this comment is provided with comments on modular HTGR-DC 10.

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Page 25 - References NRC staff does not agree with this The draft regulatory guide states: 32. comment. Reference 35 is specific to DOE, Tanju Sofu, Argonne National the rationale for why a pressure Laboratory, Sodium-cooled Fast reactor retaining containment is needed for 30 DOE/Lab General (SFR) Technology Overview SFR designs. No change.

Suggested Change The NGNP - modular HTGR training material also should be referenced.

America needs regulations that promote The NRC considers the comment to be thorium reactor research and outside the scope of this regulatory development. Smaller and safer reactors guide.

may well add to the safety of America's citizenry not only by reducing carbon foot prints and reducing money funneled into the middle east, but the major reason to 31 John Kirby General promote new research is to protect against natural disasters by providing a robust and redundant energy solution that could even survive nuclear winters from volcanos, meteors, or man.

Suggested Change None provided.

Tax me more please :):):) The NRC considers the comment to be outside the scope of this regulatory 32 Anonymous General Suggested Change guide.

None provided.

A general *comment on this and other The NRC considers the comment to be Herbert Burke project Design criteria. Go easy! outside the scope of this regulatory 33 - Energize General Experience from failures show that we guide.

Northwest cannot anticipate every problem with a complex system. So, don't try! GE, but I Page 21 of 105

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guess not Westinghouse, are big boys with deep pockets. If a big and complex program has troubles they just fix them.

Boeing spent 32 billion on the 787 Dream liner. It had problems and was late and over budget. The same is true for the GEnx engines that power the Dreamliner.

Both companies spent billions on development, had the best engineers and company experience but the both has serious problems. But they just fiedt them and went on to produce and. excellent products. These companies do not need hundreds of regulations. General ones like build the plant underground so nothing can get in or out will do. They can do it for underground nuclear tests, why not for nuclear power plants?

Remember, you can't beat Murphy's Law.

Keep it simple and let the contractor handle the design details (with supervision).

Suggested Change None provided.

Page A Appendix A The NRC staff agrees with this The draft regulatory guide states: The comment. Change was incorporated.

NRC staff then determined what if any 34 DOE/Lab General adaptation was appropriate for non-LWRs.

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The if any part should be separated from the rest of the sentence with commas: The NRC staff then determined what, if any, adaptation was appropriate for non-LWRs.

Page C Appendix C The NRC staff agrees with this Reference is made to the Glossary comment. Change was incorporated.

section of the guide for a definition of the modular HTGR, but no Glossary section 35 DOE/Lab General is provided in the draft.

Suggested Change Remove reference to the glossary.

Appendix C The staff partially agrees with this Much effort has been undertaken for comment. The title of Section IV, Fluid MHGTRs in establishing top-level Systems, is not applicable to MHTGRs.

regulatory criteria. These criteria can be The staff will change the title of MHTGR summarized in terms of reactivity control, Section IV to Heat Transport Systems.

heat removal, and radionuclide retention functions. The draft Appendix C (DG- No changes to the titles for Section I, II, 1330) retains many of the existing terms III, V, VI, and VII. The staff notes that all that have been derived for LWRs. of the MHTGR-DCs within Section VI, 36 X-Energy General Reactor Containment are not Suggested Change applicable to the MHTGR design.

As the guidance is finalized, consideration should be given to rephrasing (at least at The MHTGR-DCs themselves are the level of the recommended generally applicable to MHTGR GDC groupings) to better align with these designs. Therefore, a vendor can top-level functions. propose design specific terms for their design specific PDCs.

The terminology used within the MHTGR-DCs and rationale are Page 23 of 105

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generally applicable to MHTGR designs. A vendor can propose design-specific terms for their design-specific PDCs.

Appendix B General The NRC staff does not agree with this In several cases, SFR-DCs indicate comment. The rationale currently same as ARDC. Some others do not describes the changes made to each indicate this, when the only change is design criteria. No change.

from reactor coolant boundary to Appendix B 37 Industry/NEI primary coolant boundary.

General Suggested Change Consider indication, where applicable, that only difference from ARDC is coolant boundary designation.

Appendix B General The NRC staff agrees with this In many cases, the SFR-DC rationale comment. Change was incorporated.

include: The use of the term primary indicates that the SFR-DC are applicable only to the primary cooling system, not the intermediate cooling system. In Appendix B 38 Industry/NEI several instances, however, indicates is General replaced with implies, which connotes less certainty as to applicability.

Suggested Change Replace implies with indicates for consistency.

Appendix C Appendix C General The NRC staff does not agree with this General Many of the proposed MHTGR GDC comment. The non-LWR design criteria 39 Industry/NEI (MHTGR-DC 17, retain the statement assuming a single are an important first step to address 34, 44) failure. This inclusion makes no the unique characteristics of non-LWR reference to SECY-03-0047 and the technology. The NRC recognizes the Page 24 of 105

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Commission SRM that described the future benefits to risk informing the non-replacement of the single failure criterion LWR design criteria to the extent with a probabilistic (reliability) criterion. possible, depending on the design information and data available.

Replacing single failure criterion with a probabilistic (reliability) criterion will be Suggested Change considered in the future as part of the The single failure requirement should be NRCs Vision and Strategy activities.

replaced with a probabilistic (reliability) No change.

criterion.

Appendix B SFR-DC 1 and 10 The NRC staff does not agree with this As regard quality standards and records, comment. Design codes for specific and reactor design, no specific SFR non-LWR technologies are beyond the criteria are proposed scope of the design criteria. No change.

Appendix B 40 Industry/NEI SFR-DC 1 and 10 Suggested Change It is suggested to add that design codes adapted to SFR specificities (high temperature) must be defined.

Flexibility to Apply SARRDL The NRC staff agrees with this Some fast reactor designs utilize vented comment and notes that this flexibility is fuel concept that release the fission gas stated in Section C, Intended Use of to the primary coolant during normal This RG.

operation. SARRDL concept may be more applicable than SAFDL for such ARDC 10, SFR- designs. SARDDL would also apply more 41 DOE/Lab DC 10 readily to liquid fueled molten salt reactor concepts.

Suggested Change It would be very useful if the ARDC-10 rationale offered the flexibility to adopt the MHTGR-DC 10 approach in such cases.

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The NRC staffs incorporation of the Positive comment, no change required.

SARRDL as a replacement for the SAFDL is a very important step forward in the development of the modular HTGR 42 DOE/Lab MHTGR-DC 10 design criteria.

Suggested Change Positive comment, no change suggested.

SARRDL Definition SARRDL Definition The NRC staffs incorporation of the The NRC staff does not agree with this SARRDL as a replacement for the comment. The word system refers to SAFDL is a very important step forward the primary He coolant circuit and all in the development of the modular connected systems that are not isolated HTGR design criteria. However, the and may potentially contribute to dose change in the definition of the SARRDL, during an AOO.

replacing core with system, is problematic. The NRC apparently The first paragraph of the rationale expanded SARRDL applicability to the MHTGR-DC 10states:

entire reactor helium pressure boundary The concept of specified acceptable rather just applying it as a measure of fuel design limits, which prevent 43 DOE/Lab MHTGR-DC 10 particle fuel coating effectiveness. In additional fuel failures during addition to the concerns expressed anticipated operational occurrences below, use of system could be (AOOs), has been replaced with that of misinterpreted in the future to include the specified acceptable system systems such as the helium purification radionuclide release design limits system. (SARRDL), which limits the amount of radionuclide inventory that is released The rationale for this criterion, and the by the system under normal and AOO NRC staff presentation of 02/22/17 to conditions. The term system refers to the ACRS Subcommittee, indicates that the fuel, the helium coolant circuit and this change is intended to capture the all connected systems that are not idea that radionuclides that deposit, or isolated and may contribute to dose.

plate out, on the internal surfaces of the Page 26 of 105

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reactor helium pressure boundary can Design features within the reactor be re-entrained during normal operations system must ensure that the SARRDLS or AOOs, and that such re-entrainment are not exceeded during normal needs to be taken into account in operations and AOOs.

assessing whether the SARRDL is exceeded.

While this is conceptually true, in fact the amount of re-entrainment that occurs during an AOO is negligible.

Experiments to measure re-entrainment under depressurization conditions have shown that re-entrainment is a function of shear ratio. Shear ratio is the ratio of the maximum helium shear force during a transient event to the shear force of the flowing helium at any given location during normal, full power operation. As described in the NGNP Mechanistic Source Terms White Paper, which is listed as a reference in-situ measurements of re-entrainment vs.

shear ratio indicate that re-entrainment of radionuclides greater than 1% does not occur until the shear ratio reaches 5.

As discussed in the Preliminary Safety Information Document (PSID) for the General Atomics MHTGR, the peak shear ratio expected for the design basis depressurization event is 1.15. This design basis event entails a breach of the reactor vessel pressure relief line, Page 27 of 105

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resulting in an opening of 13 in2 and a depressurization in a period of minutes.

For the largest breach in the helium pressure boundary that would be expected to fall within the spectrum of the AOOs (failure of an instrumentation line equivalent to a breach of less than one square inch, resulting in depressurization over a period of hours),

the changes in helium flow velocity and in the shear forces on the reactor helium pressure boundary surfaces result in shear ratios less than one.

When the reactor is started up from cold shutdown, the shear forces around the helium pressure boundary are lower than those during normal, full power operation, so the shear ratios in this case are also less than one.

Insignificant re-entrainment is expected to occur when shear ratios are less than one.

It should be noted that essentially all fission product radionuclides on the reactor helium pressure boundary surfaces are originally released from the core. The release of activation products from reactor helium pressure boundary surfaces is expected to be minimal compared to release from the core. Core Page 28 of 105

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radionuclide release values are measured by grab samples (plateout activity) and plateout probes (condensed activity) for comparison with the SARRDL. Gross circulating activity is also monitored continuously. It is not possible to distinguish radionuclides that have been re-entrained from other circulating activity that is monitored or collected in a grab sample. The SARRDL value is set taking into account the fact that the plateout inventory of long-lived radionuclides will increase over time to an end of life maximum.

Due to all of the above considerations, the definition of the SARRDL should be that which was proposed by DOE/INL:

Specified Acceptable Core Radionuclide Release Design Limit.

Suggested Change Due to all of the SARRDL Definition considerations, the definition of the SARRDL should be that which was proposed by DOE/INL: Specified Acceptable Core Radionuclide Release Design Limit. SARRDL Approval MHTGR-DC 10 rationale refers to SARRDL Approval MHTGR-DC 16 because SARRDL is The Rationale states that the NRC has intertwined with functional containment.

not yet approved the SARRDL concept The staff is preparing a SECY paper for for replacement of the SAFDL and refers the Commission to discuss functional to the rationale for modular HTGR DC containment performance requirements, Page 29 of 105

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16 for information. However, the DC 16 as well as topics integral to functional rationale has no link back to DC 10 and containment (e.g., specified acceptable the SARRDL, so it is not clear what this system radionuclide release design means. limit, mechanistic source term, etc.).

The staff expects to issue the SECY Suggested Change paper in early 2018. The RG may be The paragraph that states that the NRC modified to incorporate the has not yet approved the SARRDL Commissions position if needed.

concept should be revised so that the relationship between the referenced DC 16 discussion and this issue is clarified.

Clarification is also needed regarding whether release of the Regulatory Guide will constitute approval of the SARRDL, and if release does not constitute approval, what further steps would be needed to obtain approval.

SARRDL definition was changed from NRC staff does not agree with this specified acceptable core radionuclide comment. See resolution to comment release design limits to specified No. 43 acceptable system radionuclide release 44 DOE/Lab MHTGR-DC 12 design limits.

Suggested Change See DOE Lab comment on MHTGR-DC-10 The definition of the primary coolant The NRC staff does not agree with this boundary includes the cover gas comment.

boundary. Therefore, the Criterion 14 Appendix B SFR-45 Industry/NEI requiring an extremely low probability of Similar to LWR designs, the NRC DC 14 abnormal leakage for cover gas leakage allows the use of safety valves in is not necessary. A cover gas leakage sodium fast reactors to prevent the would lead to very limited safety system from exceeding the design Page 30 of 105

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consequences (no impact on the fission pressure. The staff does not believe process, no impact or limited radiological that the SFR-DC language or the GDC consequences). This allows for safety language prohibits the use of safety valves on the cover gas system to limit valves.

abnormal pressure on the reactor vessel.

On the other hand, the failure of the The cover gas system can be designed reactor vessel could have very severe and constructed to meet this design consequences (e.g. reactivity insertion, criteria without undue burden. The failure of the core coolability). cover gas system should be operating at a low pressure that inherently Suggested Change reduces the probability of propagating It is therefore proposed to state that failure and gross rupture. The design Each part of the primary coolant criteria specifies abnormal leakage. The boundary shall be designed, fabricated, criteria should not be interpreted to erected, and tested so as to have a prohibit a small amount of cover gas prevention level of abnormal leakage, of loss due to diffusion, seal leakage, rapidly propagating failure, and of gross leakage pathways due to fabrication rupture, commensurate with the gaps, etc., all of which can only be consequences of such failures. prevented at significant and unjustifiable cost. The staff does not require the system to have perfect seals. The normal system leakage should be estimated and described by the applicant in the licensing basis as the baseline condition of the system.

The staff envisions that small leakage of the cover gas (which is greater than the normal system leakage) that does not challenge the safety of the plant would be permitted in a similar manner as the LWR Plant Technical Specifications.

Additionally, the cover gas system for a sodium fast reactor should not have an Page 31 of 105

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active degradation mechanism based upon the environment in the piping system.

Considering the consequences of leakage into the primary coolant system, the staff believes that the SFR-DC should remain as written.

The requirements as written imply the The NRC staff does not agree with this primary helium pressure retention is a comment.

safety function similar to LWRs.

The helium pressure boundary may still Suggested Change be part of the functional containment De-emphasize the pressure retention and credited in limiting radionuclide function of the helium pressure boundary. release. The quality and leak tightness of the reactor helium pressure boundary However, it is important to note that may still serve a safety function.

although the leak tightness and high quality of the helium pressure boundary is Furthermore, as currently written in the necessary for commercial operation of ASME Boiler and Pressure Vessel MHTGR-DC 14, 46 Industry/NEI MHTGRs, the pressure retaining function Code,Section III, ASME class 30, 31, 32 of the helium pressure boundary is not a components forming the primary helium required safety function. boundary are required to be pressure tested prior to ASME BPVC Suggested Change certification. Therefore, quality and leak MHTGR-DC 70 correctly emphasizes tightness for the helium pressure seismic stability and geometric stability of boundary are required for the the reactor vessel system. construction of the system as part of meeting GDC 1. A vendor can provide The safety function of the reactor vessel justification that the helium pressure and its support system is to maintain core boundary does not perform any safety coolable geometry and provide sufficient function and that the design would still meets regulatory requirements.

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conduction and convection heat transfer At present, the NRC staff has not properties in the core region. reviewed sufficient fuel qualification testing and DG-1330 does not Suggested Change sufficiently limit the fuel of MHTGRs for However, emphasis on T/H properties of the staff to state in regulatory guidance the reactor vessel at uninsulated the core that the primary coolant boundary is not region is lacking. required to protect the public health and safety for all MHTGR designs. The staff position may evolve as the Commissioners review policy papers on regulatory issues for advanced non-LWRs.

The public comment over-emphasizes the function of helium coolant boundary as it applies to 10 CFR Part 100 limits (accident dose consequences). The staff emphasis on the leak-tight design, fabrication, and testing reflects the need to ensure operability of the helium coolant boundary compared to the as-constructed condition.

The addition of the reference to modular Positive comment, no change required.

HTGR DC 30, and the associated changes to modular HTGR Criteria 14 47 DOE/Lab MHTGR-DC 14 and 30, are both excellent improvements.

Suggested Change Positive comment, no change suggested.

The addition of "heat removal systems" The NRC staff does not agree with this appears to be limited solely to connected comment. The heat removal system 48 Industry/NEI MHTGR-DC 15 systems, i.e., the steam generator. refers to any system, such as the steam Clarification is needed as to the role of generator, which serves as part of the Page 33 of 105

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the RCCS for heat removal under normal helium pressure boundary. The RCCS operations and AOOs. is not assumed to be part the helium pressure boundary in the MHTGR Suggested Change design and hence not addressed by Clarify the role of the RCCS for heat MHTGR DC 15. The heat removal removal under normal operations and capability of the RCCS may be, if AOOs. necessary, credited to demonstrate that helium pressure boundary acceptance criteria is not exceeded during normal operation including an AOO.

The changes to the text in the body of this Positive comment, no change required.

criterion made by the NRC staff relative to the proposed text in the DOE/INL report 49 DOE/Lab MHTGR-DC 15 are an improvement.

Suggested Change Positive comment, no change suggested.

The addition of "heat removal systems" The NRC staff does not agree with this appears to be limited solely to connected comment. The heat removal system systems, i.e., the steam generator. refers to any system, such as the Clarification is needed as to the role of stream generator, which serves as part the RCCS for heat removal under normal of the helium pressure boundary. The operations and AOOs. RCCS is not assumed to be part the helium pressure boundary in the 50 Industry/NEI MHTGR-DC 15 Suggested Change MHTGR design and hence not Clarify the role of the RCCS for heat addressed by MHTGR DC 15. The heat removal under normal operations and removal capability of the RCCS may be, AOOs. if necessary, credited to demonstrate that helium pressure boundary acceptance criteria is not exceeded during normal operation including an AOO.

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Removal of the Word System The NRC staff agrees with this The changes to the text in the body of this comment. Change was incorporated.

criterion made by the NRC staff relative to the proposed text in the DOE/INL report are an improvement. However, the word System should be removed from the title of the criterion. The reactor helium pressure boundary is not an individual system, but rather is constituted from 51 DOE/Lab MHTGR-DC 15 parts of several systems, which are listed and referred to in the body of the criterion.

Removal of the word System from the title will make the title consistent with modular HTGR terminology.

Suggested Change Remove the word System from the title of the criterion.

Appendix A ARDC 16 Page A-4 The NRC staff does not agree with this The draft guidance for ARDC 16, comment. The staff believes that the Containment design, retains the original Commission may wish to assess the GDC language, thereby carrying forward reactor technologies and possible design criteria intended for a pressure- approaches to functional containment retaining light water reactor. This results that are different from those previously in limiting the applicability of the functional presented for MHTGRs. The staff is 52 Industry/NEI ARDC 16 containment concept to applicable non- preparing a SECY paper for the LWR designs, and appears to be Commission to discuss functional inconsistent with the Commissions containment performance requirements, position on alternatives to a leak tight as well as topics integral to functional containment, as discussed in SECY 93- containment (e.g., specified acceptable 092 and the associated SRM. Advanced system radionuclide release design reactor containment design guidance limit, mechanistic source term, etc.).

should flow logically from ARDC 16 to the The staff expects to issue the SECY Page 35 of 105

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SFR and MHTGR design criteria. ARDC paper in early to mid-2018. The RG 16 should be a high-level technology- may be modified to incorporate the neutral design criterion from which Commissions position if needed. Until technology-specific design criteria are this time, ARDC 16 will remain as derived. same as GDC 16.

Suggested Change ARDC 16 language should include technology neutral containment requirements which can be subsequently applied to a specific technology. The original DOE/INL language for ARDC 16 is provided below.

Containment design A reactor functional containment consisting of a structure surrounding the reactor and its cooling system or multiple barriers internal and/or external to the reactor and its cooling system, shall be provided to control the release of radioactivity to the environment and to assure that the functional containment design conditions important to safety are not exceeded for as long as postulated accident conditions require.

The concept of a functional containment would be of interest for application to other technologies.

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Applying this recommendation would provide a high-level technology-neutral ARDC which could be used to obtain Commission approval of containment performance criteria. SFR and MHTGR DC 16 would then serve to illustrate how technology-specific design criteria can be derived from ARDC 16.

Add Functional Containment The NRC staff does not agree with this Language comment. The staff believes that the ARDC 16 language should include Commission may wish to assess the technology neutral containment reactor technologies and possible requirements which can be subsequently approaches to functional containment applied to a specific technology. The that are different from those previously original DOE/INL language for ARDC 16, presented for MHTGRs. The staff is which was written with the objective of preparing a SECY paper for the being technology neutral, is provided Commission to discuss functional below. containment performance requirements, as well as topics integral to functional Containment design. A reactor containment (e.g., specified acceptable 53 DOE/Lab ARDC 16 functional containment consisting system radionuclide release design of a structure surrounding the limit, mechanistic source term, etc.).

reactor and its cooling system or The staff expects to issue the SECY multiple barriers internal and/or paper in early to mid-2018. The RG external to the reactor and its may be modified to incorporate the cooling system, shall be provided Commissions position if needed. Until to control the release of this time, ARDC 16 will remain as radioactivity to the environment same as GDC 16. The last sentence and to assure that the functional of the ARDC 16 rationale will be containment design conditions deleted. The second to the last important to safety are not sentence will remain until the exceeded for as long as Commissions position is clarified in its response to the SECY paper.

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postulated accident conditions require.

The concept of a functional containment would be of interest for application to other technologies. Applying this recommendation would provide a high-level technology-neutral ARDC which could be used to obtain Commission approval of containment performance criteria. SFR and MHTGR DC 16 would then serve to illustrate how technology-specific design criteria can be derived from ARDC 16.

Suggested Change ARDC 16 language should include technology neutral containment requirements which can be subsequently applied to a specific technology. The original DOE/INL language for ARDC 16, which was written with the objective of being technology neutral, is provided below.

Containment design. A reactor functional containment consisting of a structure surrounding the reactor and its cooling system or multiple barriers internal and/or external to the reactor and its cooling system, shall be provided to control the release of Page 38 of 105

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radioactivity to the environment and to assure that the functional containment design conditions important to safety are not exceeded for as long as postulated accident conditions require.

Functional Containment Policy Issue Discussions of Commission policy decisions on functional containment need to be worded carefully. For the modular HTGR, a policy decision is not needed regarding the general acceptability of applying a functional containment (radionuclide retention) approach that differs from a conventional LWR high-pressure, low-leakage structure.

However, based on the SRM to SECY-03-0047, a policy decision is needed regarding the performance criteria to be applied to a functional containment. The information located in the MHTGR-DC 16 rationale correctly states that a policy decision regarding functional containment performance requirements and criteria will be needed. Its noted that containment performance criteria for LWRs are provided in 10 CFR 50 Appendix J, rather than in the GDC of Appendix A.

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The last two sentences in the rationale for ARDC 16 should be deleted.

Appendix A ARDC 16 Page A-4 The NRC staff does not agree with this Clarify that use of ARDC 16 [per industry comment. If a reactor is able to comment ##] for non-LWR designs other demonstrate safety margins and/or than MHTGRs may be subject to a policy consequences below regulatory limits decision Making a justification, similar using barriers other than an essentially to that for research reactors and non- leak-tight structure, a functional power reactors has basis in NRC policy containment may be justified. However, and should not require a Commission- the Commission may wish to assess the level policy decision. reactor technologies and possible Discussions of Commission policy approaches to functional containment decisions on functional containment need that are different from those previously to be worded carefully. For the modular presented for MHTGRs. The staff HTGR, a policy decision is not needed expects to consider functional regarding the general acceptability of containment concepts and associated applying a functional containment performance criteria within the broader 54 Industry/NEI ARDC 16 (radionuclide retention) approach that development of a risk-informed, differs from a conventional LWR high- performance-based, technology-pressure, low-leakage structure. inclusive regulatory framework for non-However, based on the SRM to SECY- LWRs. Those efforts will include 03-0047, a policy decision is needed making proposals and regarding the performance criteria to be recommendations to the Commission, applied to a functional containment. The which will clarify the potential use of information located in the MHTGR-DC 16 functional containment for various rationale correctly states that a policy reactor technologies.

decision regarding functional containment performance requirements and criteria will be needed. Its noted that containment performance criteria for LWRs are provided in 10 CFR 50 Appendix J, rather than in the GDC of Appendix A.

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Suggested Change Revise rationale to state, However, it is also recognized that characteristics of the coolants, fuels, and containments to be used in other non-LWR designs could share common features with SFRs and MHTGRsUse of the ARDC 16 for non-LWR designs other than MHTGRs-DC 16 will may be subject to a policy decision by the Commission. If a reactor is able to demonstrate safety margins and/or consequences on the order of those demonstrated by non-power and research reactors, a functional containment may be justified, and the reactor may be able to use ARDC 16 without a Commission level policy decision. See rationale for MHTGR-DC 16 for further information on the policy decision.

It is indicated that the reactor The NRC staff does not agree with this containment is a pressure retaining comment. The NRC staff considers structure surrounding the reactor and its pressure retention essential to cooling systems. In case of SFR, it is accommodate the impact of sodium possible to limit the pressure loadings on reactions with air or water that could the containment structure in accident release significant energy inside the 55 Industry/NEI SFR-DC 16 conditions. For example the rooms with containment structure. The language in sodium circuits can be designed so that the rationale makes it clear that the the effect of a sodium leak or fire would pressure retention is not like that in an not result in significant pressure on the LWR containment. Several references containment structure and the pressure have also been cited to clarify this point.

effect could be limited to the room where [Ref: SRM to SECY-03-0047 Page 41 of 105

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the leak occurs. Also, the reactor cooling (ML031770124), and VR to SECY systems could include secondary cooling 0047 (ML031770333) outline the systems which are partially outside the Commissions view on pressure containment structure where this can be retaining containment]

particular concern is cooling systems with air as the heat sink, for which sodium/air The last paragraph of SFR-DC 16 heat exchanger must be placed outside clarifies the expectation for pressure of the containment. retaining characteristics of SFR designs.

Suggested Change It is therefore proposed to modify the first Also, the words in the first sentence of sentence of the criterion as: A reactor the criterion and its primary cooling containment consisting of a high strength, system, are appropriate and need not low leakage, pressure retaining structure be changed or removed.

surrounding the reactor and its cooling systems shall be provided to control the release of radioactivity to the environment and to assure that the reactor containment design conditions important to safety are not exceeded for as long as postulated accident conditions require. Additionally, remove the phrase and its primary cooling system.

Under rationale, statement that all past, The NRC staff agrees with this current, and planned SFR designs use a comment. Change was incorporated.

high strength, low-leakage, pressure-retaining containment concept seems 56 Industry/NEI SFR-DC 16 broader than can be substantiated without knowledge of all planned designs.

Suggested Change Delete and planned 57 DOE/Lab MHTGR-DC 16 Functional Containment Policy Issue Functional Containment Policy Issue Page 42 of 105

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Discussions of Commission policy The NRC staff does not agree with this decisions on functional containment need comment. The rationale for MHTGR-to be worded carefully. For the modular DC 16 states that the Commission has HTGR, a policy decision is not needed found the concept of functional regarding the general acceptability of containment acceptable, The NRC staff applying a functional containment has brought the issue of functional (radionuclide retention) approach that containment to the Commission, and differs from a conventional LWR high- the Commission has found it generally pressure, low-leakage structure. acceptable, as indicated in the staff However, based on the SRM to SECY- requirements memoranda (SRM) to 03-0047, a policy decision is needed SECY 93 092 (Ref. 8) and SECY 03 regarding the performance criteria to be 0047 (Ref. 9). The rationale goes on to applied to a functional containment. The indicate that the Commission instructed information located in the MHTGR-DC 16 the staff to develop performance rationale correctly states that a policy requirements and criteria working decision regarding functional containment closely with industry experts (e.g.,

performance requirements and criteria will designers, EPRI, etc.) and other be needed. Its noted that containment stakeholders regarding options in this performance criteria for LWRs are area, taking into account such features provided in 10 CFR 50 Appendix J, rather as core, fuel, and cooling systems than in the GDC of Appendix A. The last design, and directed the staff to two sentences in the rationale for ARDC submit options and recommendations 16 should be deleted. to the Commission for a policy decision. This is language taken directly from the SRM to SECY 03-0047. The last sentence of ARDC 16 will be deleted.

Functional Containment Language Functional Containment Language ARDC 16 should discuss functional The NRC staff does not agree with this containment with the MHTGR-DC comment. See NRC resolution to ARDC referring to the ARDC. See ARDC 16 16 comment No. 55 team comment.

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Functional Containment Performance Functional Containment Standard Performance Standard The NRC staff notes in the next-to-last The NRC staff does not agree with this rationale paragraph that the staff has comment. Although the NRC staff provided feedback to DOE on the use of noted that the DOE proposed a functional containment as part of its performance standard for the modular review of the NGNP. The rationale should HTGR functional containment is also note that the NRC staff also stated in reasonable in the NGNP feedback, this its assessment report that it finds the was never brought before the DOE proposed performance standard for Commission.

the modular HTGR functional containment to be reasonable. This performance standard ensures the integrity of the fuel particle barriers rather than to allow significant fuel particle failures and then to rely extensively on other mechanistic barriers.

Suggested Change Reword the rationale to clarify what policy decisions have been made and what decisions need to be made. Delete last two sentences of the rationale.

Appendix A, ARDC 17, Page A-4 The NRC staff agrees with this Clarify A reliable power system is comment. Change was incorporated.

required for SSCs during postulated accident conditions to apply to SSCs 58 Industry/NEI ARDC 17 whose safety performance relies on electric power Suggested Change Modify to:

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A reliable power system is required for SSCs during postulated accident conditions when those SSCs safety functions require electric power.

The following text is confusing: The NRC staff agrees with this The existing single switchyard allowance comment and has removed the text remains available under ARDC 17. If a related to the single switchyard particular advanced design requires the allowance.

use of GDC single switchyard allowance wording, the designer should look to GDC 17 for guidance when developing PDC.

59 Industry/NEI ARDC 17 Suggested Change Suggest rewording to: The single switchyard allowance under GDC 17 is not eliminated because of the changes in ARDC 17; if a particular advanced design ARDC 17 states the safety function for The NRC staff agrees with this the electrical systems shall be to provide comment. ARDC 17 was modified to sufficient capacity, capability, and clarify functions that rely on electric reliability to ensure thatvital functions power during postulated accidents and that rely on electric power are maintained to address the use of electrical power in the event of postulated accidents. The for the performance of the prescribed scope of vital functions is unclear. For safety functions.

60 Industry/NEI ARDC 17 example, it is unclear if the independent and diverse means of shutdown prescribed by ARDC 26 paragraph 2 is considered such a vital function.

Further, the Rationale for ARDC 17 states If electrical power is not required to permit functioning of SSCs important to Page 45 of 105

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safety, the requirements in the ARDC are not applicable to the design. In this case, the functionality of SSCs important to safety must be fully evaluated and documented in the design bases. The requirements of ARDC 17 are related to performance of the prescribed safety functions (e.g., sufficient redundancy to perform their safety functions).

Accordingly, it appears the appropriate test for applicability of ARDC 17 is whether electrical power is required to perform the specifically prescribed safety functions, not the functioning of SSCs important to safety more generally.

Suggested Change Revise ARDC 17 with respect to the postulated accident safety function, or clarify the scope of vital functions with the Rationale.

Revise the Rationale discussion on applicability of ARDC 17 to address the use of electrical power for the performance of the prescribed safety functions.

Editorial: The existing single switchyard The NRC staff agrees with this allowance remains available under comment. Change was incorporated.

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As indicated. However, also refer to comment on ARDC 17 suggesting rewording of this rationale discussion.

Editorial: The existing single switchyard The NRC staff agrees with this allowance remains available under ARDC comment. Change was incorporated.

MHTGR-DC 17 62 Industry/NEI MHTGR-DC 17 Suggested Change As indicated. However, also refer to comment on ARDC 17 suggesting rewording of this rationale discussion.

The team commends the NRC for this Positive comment, no change required.

criterion adaptation. The adaptation provides increased flexibility for designers and license applicants as they pursue enhanced margins of safety and the use of simplified, inherent, passive, or other innovative means to accomplish safety and security functions, consistent with the ARDC, SFR-DC, 63 DOE/Lab Commissions policy on advanced MHTGR-DC 17 reactors.

This positive comment also applies to the corresponding SFR-DC-17 and modular HTGR-DC-17.

Suggested Change Positive comment, no change suggested.

Use of the Word Systems The NRC staff agrees with this ARDC, SFR-DC, Based on the ACRS discussion of comment. These design criteria and 64 DOE/Lab MHTGR-DC 17 02/22/17, we might wish to request the rationale were modified to clarify increased clarity on what is intended systems.

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when the plural systems is used with respect to duplicate and independent power supply. As written now, multiple independent systems are more implied rather than explicitly stated in the design criterion.

Suggested Change None provided.

Rationale Wording Inconsistency The NRC staff does not agree with this Paragraph two of the rationale refers to comment. The last sentence of ARDC, the deletion of words in GDC 18 SFR-DC and MHTGR-DC 18 was pertaining to additional system examples, modified from the original GDC as ARDC, SFR-DC, but there do not appear to be any such shown below:

65 DOE/Lab MHTGR-DC 18 deletions from the text of the criterion. and the transfer of power among the Suggested Change nuclear power unit, the offsite power Remove the second sentence in the system and the onsite power systems.

rationale. This is what the second paragraph of the rationale is referring to.

Appendix A, ARDC 19, Page A-6 The NRC staff does not agree with this This criterion presumes that operator comment. Remote operation of action is required and that operator reactors requiring no operator action or actions, including monitoring, must be monitoring may be subject to a policy performed from a single location (i.e., a decision by the NRC. Applicants can control room). propose a non-traditional means of controlling the plant as part of the PDC 66 Industry/NEI ARDC 19 Suggested Change and application.

Consideration should be given to an applicant demonstrating that operator action, including monitoring, is not required for safety, and/or that any necessary actions, including monitoring, could be demonstrated to be feasible Page 48 of 105

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from additional and/or redundant and/or remote locations.

Appendix A, ARDC 19, Page A-6 The NRC staff does not agree with this The way the text is written still appears to comment. Elimination of the assume some fundamental, legacy needs requirement for a control room may be in a power plant. None of this makes subject to a Policy decision by the NRC.

sense if operators have literally zero Applicants can propose a non-ability to influence the safety of the plant traditional means of controlling the plant because it is physically inherent (note: not as part of the PDC and application.

67 Industry/NEI ARDC 19 to be confused with inherent safety as defined by the IAEA, which requires no decay heat)

Suggested Change As with some other sections, frame with As applicable to plant design:

Delete "as defined in § 50.2" as this is NRC does not agree with this comment.

implicit in all of the GDC statements. Reference to § 50.2 appears in GDC 19 68 Industry/NEI ARDC 19 and was included in ARDC 19 as well.

Suggested Change Delete "as defined in § 50.2" Appendix A, ARDC 25 through 28, The NRC staff does not agree with this Page A-7 comment. ARDC 26, which replaces It appears assumed that GDCs 26 and 27, has been rewritten to control/protection systems are required state that either a system or means of for reactivity control. It also assumes that reactivity control can be used.

the ultimate reactivity protection 69 Industry/NEI ARDC 25-28 No changes to ARDCs 25 and 28 are mechanism is still an active function. This assumption is not necessarily true for all necessary. The suggested change, as designs. The term system indicates applicable to plant design is true of all active/designed to us. ARDCs because the applicant can As with some other sections, frame with submit plant specific PDCs. Also, the As applicable to plant design: staff envisions that many non-LWR Page 49 of 105

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designs will have an active means of Suggested Change adding positive reactivity to account for As with some other sections, frame with fuel depletion (e.g., control rods, As applicable to plant design: dilution, reflector movement, or fuel additions) whose failure could add positive reactivity, which is addressed by ARDC 25 for AOOs and ARDC 28 for postulated accidents.

Appendix A, ARDC 26, Page A-7 With the exception of the section (1) Capability (1) is specific to having a regarding design basis events, the NRC means to shut down the reactor in staff does not agree with this comment.

regularly occurring situations. The No change is proposed because the move from specified acceptable fuel staff is not able to define appropriate design limits to fission product barriers margin based on the possible number is a significant improvement towards of different advanced reactor reactivity technology neutrality, enabling mechanisms and their associated accurate safety assessment of both failure modes.

more conventional fuel forms with more complex fuel forms including The staff notes GDCs 26 and 27 are liquid fuel forms on the same basis. explicit in stating that the appropriate 70 Industry/NEI ARDC 26 margin is a stuck rod(s). Also, GDC 26 states that one of the reactivity That being said, there was concern that mechanisms shall use control rods.

there are some possible components Therefore, determination of a specific considered as fission product barriers failure mechanism, a stuck rod(s), could could fail without significant impact to be defined. Appropriate margin should safety. Therefore, words were added to be addressed on a case by case basis.

ensure that the focus is on only those fission product barriers that are safety (1) The current GDCs do not address related. safety versus non-safety related.

Therefore, for consistency with the (2) Many industry comments current GDCs, the ARDCs will not included reasoning that two specify safety versus non-safety related Page 50 of 105

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independent means for shutting and no change to the ARDC 26 will be down the reactor and maintaining made. The staff does agree that safety-shutdown may not be needed, related fission product barriers are especially for reactor types that those credited in determination of offsite have natural or passive means for dose consistent with the definition of 10 shutdown as the primary means. CFR 50.2.

In addition, the requirement for two fully independent means both (2) Staff agrees with using anticipated capable of achieving and operational occurrences and postulated maintaining shutdown does not accidents instead of design basis seem to be the standard for LWRs. events. This change will be made in the RG.

This presents the simplest wording that GDC 26 states that two independent allows for reactors with inherent or reactivity control systems of different passive shutdown fundamental to the design principles shall be provided.

physics of the system to make a ARDC 26 is written consistent with GDC justification that a second means would 26 as two systems are still required be superfluous. It also allows for reactors even if one is inherent or passive. The to make a probability risk assessment to staff notes that both the PRISM and make a similar justification. MHTGR had inherent reactivity mechanisms but also retained a The wording change from design basis secondary means of achieving events to anticipated operational shutdown through the use of control occurrences and postulated accidents is rods. Therefore, the proposed taken from the NRCs Rationale and modification to ARDC 26 to allow for ensures that what is being referred to is one reactivity mechanism will not be clearly outlined terminology in the incorporated.

regulation.

ARDC 26 (1) has been revised such (3) The requirement of that safe shutdown is achieved and subcriticality may not be the most maintained during an AOO. ARDC 26 appropriate measure of safe (3) states that following an AOO or shutdown. For example, it has postulated accident shutdown is Page 51 of 105

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been demonstrated in various maintained. The combination of ARDC reactor types that a safe, long term 26 (1) and (3) allow for a temporary re-shutdown could be achieved criticality during a postulated accident naturally without rods or coolant consistent with the licensing basis of even if brief moments of criticality some PWRs. ARDC 26 (1) and (3) occurred. (see Secondary should be accomplished using safety-shutdown systems of Nuclear related SCCs.

Power Plants, ORNLNSIC-7, January 1966). Wording was (3) The reactor should be subcritical, taken directly from the NRC with appropriate margin, under Rationale to expand the capability conditions so corrective actions on to account for such a capability in SCCs or normal refueling operations certain designs. can occur. The assumption is that the requirements of ARDC 26 (3) are met With the addition of the phrase before preceding to conditions of lower appropriate margin for malfunctions, it is pressure and/or temperatures (i.e.,

important that the subjective phrase be conditions) at which corrective actions defined by NRC. This wording is an or refueling can take place. Therefore, attempt to define appropriate margin no change to ARDC 26 (4) will be with options for both deterministic and made.

risk-informed scenarios for malfunction.

Depending on the reactor type, it may be preferred to utilize the simplicity of a deterministic approach. There also may or may not be enough data to utilize a risk- informed approach. For others, a risk-informed approach may more accurately determine appropriate margin.

The previous metric of maintaining fission product barriers is kept as the primary metric in this measurement of margin.

The definition could be:

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(1) A single active failure must not result in exceeding design limits for safety related fission product barriers, or (2) The probability for a malfunction of the means must not be greater than the frequency for AOOs. If the probability is greater than the frequency for postulated accidents by an order magnitude or more, that malfunction must not result in exceeding design limits for safety-related fission product barriers.

Suggested Change Define Appropriate Margin AND Change wording to the below (red italics indicates changed wording, red indicates added wording)

Reactivity control systems shall include the following capabilities:

(1) A means of shutting down the reactor shall be provided to ensure that, under conditions of normal operation, including anticipated operational occurrences, and with appropriate margin for malfunctions, design limits for safety-related fission product barriers are not exceeded.

(2) A means of shutting down the reactor and maintaining a safe shutdown in Page 53 of 105

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anticipated operational occurrences and postulated accidents, with appropriate margin for malfunctions, shall be provided. If the primary means for shutdown is not inherent, passive, or shown to have a probability of failure an order of magnitude less than that of postulated accidents, a second means of reactivity control shall be provided that is independent, diverse, and capable of achieving and maintaining safe shutdown both for anticipated operational occurrences and postulated accidents. (3)

A system for holding the reactor subcritical in the long term or in an equilibrium condition naturally achieved by the design under cold conditions shall be provided.

The second to last paragraph of the The NRC staff does not agree with this ARDC 26 rationale states: comment. ARDC 26 (2) is consistent with the second reactivity system of The second sentence of ARDC GDC 26. The requirement of having a 26(2) refers to a means of second, backup shutdown system has achieving and maintaining been eliminated. Only one safety-related shutdown that is important to or a combination of safety-related 71 Industry/NEI ARDC 26 safety but not necessarily safety means (SSCs) are needed to achieve related. The second means of and maintain shutdown for AOOs reactivity control serves as a (ARDC 26 (1)) and maintain shutdown backup to the safety-related following an AOO or postulated accident means and, as such, margins for (ARDC 26 (3)).

malfunctions are not required but the second means shall be highly Page 54 of 105

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reliable and robust (e.g., meet ARDC 1 -5).

The distinction between the terms important to safety and safety-related is not properly defined. To avoid confusion, the statement should be revised.

Suggested Change Recommend restating the rational to say:

The second sentence of ARDC 26(2) refers to a means of achieving and maintaining shutdown that is important to safety but not necessarily safety related.

The second means of reactivity control which serves as a backup to the safety-related primary means and, as such, margins for malfunctions are not required but the second means shall be highly reliable and robust (e.g., meet ARDC 1-5).

ARDC 26, Reactivity Control Systems. The NRC staff does not agree with this ARDC 26 replaces "specified acceptable comment. This change is specific to fuel design limits" with design limits for ARDC 26. It would not be appropriate fission product barriers." Why is "specified to change this throughout the ARDC.

72 Peter Smith ARDC 26 acceptable fuel design limits" not similarly replaced throughout the ARDC?

Suggested Change None provided.

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GDC 26 and GDC 27 requirements are: The NRC staff does not agree with this

• Two independent reactivity control comment Since SFR-DC 26 is the same systems of different design as ARDC 26, this comment will principles shall be provided. therefore refer to ARDC 26.

• One of the systems shall use The first bullet is addressed by ARDC control elements and be capable 26 (2), which states that an independent of reliably controlling reactivity and diverse method is used for changes to assure that under reactivity control. As stated in the ARDC conditions of normal operation, 26 (2) rationale discussion, The term including anticipated operational independent and diverse indicates no occurrences (AOOs), and with shared systems or components and a appropriate margin for design which is different enough such malfunctions such as stuck control that no common failure modes exist elements, specified acceptable between the system or means in ARDC fuel design limits are not 26 (2) and safety-related systems in exceeded.

SFR-DC 26 ARDC 26 (1) and (3).

73 Industry/NEI

• The second reactivity control SFR-DC 27 system shall be capable of reliably Regarding the second bullet, no controlling the rate of reactivity changes are necessary as ARDC 26 (1) changes resulting from planned, preserves the same requirements as normal power changes to assure GDC 26.

acceptable fuel design limits are not exceeded. Regarding the third bullet, the revised

• One of the systems shall be ARDC 26 (2) reverts back to the original capable of holding the reactor reactivity requirement of the second core subcritical under cold reactivity system in GDC26, namely, conditions. controlling the rate of reactivity changes

• The reactivity control systems resulting from planned, normal power shall be designed to have a changes to assure acceptable fuel combined capability of reliably design limits are not exceeded.

controlling reactivity changes to Addressing the fourth bullet, the word assure that under postulated cold has been replaced with accident conditions and with Page 56 of 105

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appropriate margin for stuck conditions which allow for interventions control elements the capability to such as fuel loading, inspection and cool the core is maintained. repair as there is no consistent

• definition of cold conditions for non-Current BWRs and PWRs in the US have LWR designs.

two independent systems for controlling reactivity through movement and Addressing the fifth bullet, the revised positioning of control rods. ARDC 26 (1) and (3) replace the words reliably controlling reactivity in GDCs To attain the desired core power level and 26 and 27 with a quantitative power distribution during normal requirement of achieving and operation, one reactivity control system is maintaining shutdown during and used to position control rods to following AOOs and maintaining compensate for reactivity due to changes shutdown following AOOs and in temperature and fuel burnup. BWRs postulated accidents. Consistent with also used core flow and PWRs also use the current licensing basis of some boration to help control reactivity during PWRs, re-criticality during a postulated normal operation. To ensure all safety accident would be allowed if adequate criteria are met during AOOs and DBAs, a heat removal capability is available. The second reactivity control system is used to rationale associated with ARDCs 26 (3) provide rapid, full insertion of all control describes the conditions which define rods (scram). The circuitry and hardware following an AOO or postulated used to move the control rods are accident.

completely independent for the two reactivity control systems.

The reactivity worth of the control rods is sufficient to ensure reactor shutdown when the rods are fully inserted by either control system for BWRs. For PWRs, control rod insertion and boration ensure reactor shutdown.

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US LWRs have implemented design features to provide an alternate method for reactor shutdown in the event that the reactivity shutdown system (scram) fails.

For PWRs, alternate control rod insertion methods in the event of scram failure have been implemented (same control rods as normal scram, but an independent method for inserting the rods). For BWRs, standby liquid boron injection systems are used to provide an alternate method for reactor shutdown.

These alternate means to shut down the reactor are required to meet 10CFR50.62 requirements. Note, these alternate means of shutdown are for a beyond design basis event and the requirements are not addressed in the GDC.

Requirement differences with NRC SFR-DC 26:

• Item (1) of SFR-DC 26 changes specified acceptable fuel limits to design limits for fission product barriers. Challenges to primary coolant boundary or containment boundary are addressed in other GDCs. Change is not necessary, but does not add new requirement.

• Item (2) of SFR-DC 26 changes the requirement to provide capability to cool the core during Page 58 of 105

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postulated accidents to maintaining a safe shutdown under design basis events. The reactivity control system requirement has been extended from ensuring core damage does not prevent core cooling to including other aspects (e.g. heat removal from primary system) of safe shutdown. Additional requirements to achieve safe shutdown are addressed by other GDCs. The term design basis events is not used in the GDCs.

• Item (2) of SFR-DC 26 adds the requirement to have a second independent shutdown system for design basis events. 10CFR does not require a second independent shutdown system for design basis events. 10CFR requires an alternate means of shutdown for beyond design basis events (10CFR50.62).

• SFR-DC 26 eliminates the requirement that the reactivity control system for normal operation reactivity control be independent from the reactivity control system used for shutdown (scram).

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Recommend retaining GDC 26 and 27 unchanged as SFR-DC 26 and SFR-DC

27. GDC 26 and 27 are applicable for currently licensed and operating LWRs.

The reactivity control requirements currently in place for LWRs are sufficient for SFRs.

The existing GDC includes the wording The NRC staff agrees with this "specified acceptable fuel design limits", comment. MHTGR-DC 26 (1) was while the proposed MHTGR-DC does not modified to reflect the AOO design include the replacement "specified criteria associated with MHTGR DC 10 acceptable system radionuclide release (SARRDLs) and MHTGR DC 15 design limits" wording. The wording that (Helium pressure boundary).

"design limits for fission product barriers are not exceeded" is imprecise and moves the intent from maintaining fuel design limits to fission product barriers.

The rationale describes: "Additionally, 74 Industry/NEI MHTGR-DC 26 specified acceptable fuel design limits is replaced with design limits for fission product barriers to be consistent with the AOO acceptance criteria." This appears to be inconsistent with other design criteria which include SARRDLs. See proposed MHTGR-DC 10, 17, 20, and 25.

Suggested Change Recommend establishing consistency between MHTGR-DC 26 and other design criteria mentioned.

The original GDC 26 language was Positive comment, no change required.

ARDC, SFR-DC 75 DOE/Lab unnecessarily confusing and the staffs MHTGR-DC 26 proposed revision of ARDC 26-27 offers Page 60 of 105

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greater clarity of underlying safety intent.

Generally speaking, the team agrees that the revised structure of ARDC 26 is a significant improvement.

This positive comment also applies to the corresponding SFR-DC 26 and MHTGR-DC 26.

Suggested Change Positive comment, no change suggested.

Important to Safety Important to Safety The term important to safety is almost The NRC staff does not agree with this universally understood to mean safety- comment. Safety-related is a subset of related in the context of the GDC and Important to Safety. Specific examples ARDC. ARDC 1-5, referenced in the include the PWR chemical and volume phrase highly reliable and robust (e.g., control system (CVCS) system that is meet ARDC 1-5) most often refer to used to satisfy GDC 33, Reactor safety functions, strongly implying safety Coolant Makeup, and the second systems. The DOE/INL ARDC report reactivity control system of GDC 26 that, (December 2014) defined important to shall be capable of reliably controlling safety as follows: the rate of reactivity changes resulting ARDC, SFR-DC 76 DOE/Lab Based on existing 10 CFR 50 from planned, normal power changes MHTGR-DC 26 Appendix A language, this (including xenon burnout) to assure designation refers to structures, acceptable fuel design limits are not systems, and components (SSCs) exceeded.. Many plants CVCS that provide reasonable assurance systems are not credited to prevent or the facility can be operated without mitigate an AOO or postulated accident.

undue risk to the health and safety In these cases, the CVCS system and of the public. SSCs with this the second reactivity control system are designation are safety related and important to safety, but not safety-are relied upon to remain related.

functional during design basis accidents. Undue risk is ARDC Scope changes Page 61 of 105

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associated with the inability to The NRC staff does not agree with this ensure the capability to prevent or comment. ARDC 26 (2) has been mitigate the consequences of revised to maintain consistency with the accidents which could result in second reactivity system of GDC 26.

offsite radiological consequences The requirement of having a second, exceeding the limits set forth in 10 backup shutdown system has been CFR 50.34 (or 10 CFR 52.79). eliminated. Only one safety-related or a combination of safety-related means Suggested Change (SSCs) are needed to achieve and Within the scope and context of the GDC, maintain shutdown for AOOs (ARDC 26 important to safety is equivalent to (1)) and maintain shutdown following an safety related. Therefore, it is AOO or postulated accident (ARDC 26 recommended that the subject paragraph (3)).

in the rationale be reworded to avoid potential contradiction with the common The term fission product barriers was usage of the term throughout the GDC added to ARDC 26 to address reactors and ARDC. without mechanical fuel limits (i.e.,

SAFDLs) such as liquid fueled reactors.

ARDC Scope Changes Adding fission product barriers is not an Item (1) seems to have a narrower focus extension of requirements, because than the GDC, focusing more on ARDC 15 requires that the reactor shutdown capability than on reactivity coolant system be designed withstand control and does not appear to reflect the normal operations including AOOs.

requirement of GDC 26 to have two MHTGR-DC 15 and MHTGR-DC 26 reactivity control systems for controlling refer to maintaining the helium pressure reactivity for normal operations and boundary for initiating events that may AOOs. In addition, Item (2) of this not include small failures of the pressure combined design criteria requires two boundary (e.g., failure of instrument or independent and diverse means of sample lines). Rather, these DC refer to achieving and maintaining safe shutdown preventing failures which substantially under design basis conditions whereas reduces the helium pressure boundarys GDC 27 seems to allow a collective and effectiveness as a fission product combined capability. barrier, if so credited, and would prevent Page 62 of 105

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preserving a passive coolable geometry.

The existing rationale does not explicitly Determining the event classification of explain the apparent scope changes that breaks associated with small lines occurred in the transition from the original connected to the helium coolant GDC language to the current ARDC 26 boundary is outside the scope of this language. task.

ARDC 26 Item (1) also included the The term fission product barrier does replacement of specified acceptable fuel not include functional containment. The design limits with design limits for fission rationale of ARDC 26 notes that ARDC product barriers. The discussion in the 15, Reactor coolant system design, rationale and the NRC staff presentation provides the appropriate design limits of February 22, 2017, indicate that the for the reactor coolant boundary. Also, focus of this change is on both the fuel the staff does not envision an AOO and the reactor coolant boundary. event of sufficient severity which would Addition of the reactor coolant boundary challenge functional containment.

is an increase in scope from GDC 26 SFR-DC 26 is identical to ARDC 26. In relative to what needs to be protected MHTGR-DC 26, the term design limits from failure during normal operation and of fission product barriers was replaced AOOs. This change is inconsistent with with terms from MHTGR design criteria the fact that some AOOs could involve MHTGR-DC 10 (SARRDL) and failure of fission product barriers (e.g., MHTGR-DC 15 (helium pressure failure of instrumentation lines, sample boundary). The helium pressure lines, etc.). Furthermore, nothing is boundary criteria was explicitly included provided in the rationale to prevent future because the pressure boundary should interpretations of the language as also not fail due to an initiating AOO such as encompassing the reactor containment reactivity or loss of normal heat removal for those designs that use a traditional event.

approach to containment.

MHTGR-DC 26 (1) has been explicitly Suggested Change written to denote the helium coolant The rationale should be revised to include pressure boundary and not the reactor an explanation for the apparent scope Page 63 of 105

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changes. In addition, a change in the title, functional containment - see the such as Reactivity Control System response to Comment 74.

Shutdown Capability, would better align the ARDC and its title. Safe Shutdown, Cold Conditions All of these points need clarification. Agree. ARDC 26 (4) has been reworded to:

Safe Shutdown, Cold Conditions A system for holding the reactor Terminology shutdown under conditions which allow Suggested alternative to cold conditions for interventions such as fuel loading, for SFR DC 26. Use the definition of inspection and repair.

subcritical under cold conditions comes from the work on GIF SFR design criteria. ARDC Development References Subcritical under cold conditions is The NRC staff does not agree with this defined as the state with the comment. The use of ARDC reactivity of the reactor kept to a development references were primarily margin below criticality under a used to determine the safety function of prescribed coolant temperature the second reactivity system in GDC 26.

condition in which interventions The revised ARDC 26 (2) is consistent such as fuel reloading, periodic with the second reactivity in GDC 26 inspection and repair work in the which is SAFDL preservation under reactor can be achievable. planned, normal power changes.

Use of Design-Basis This is very similar to cold conditions for Agree. Design basis will be replaced LWRs if the prescribed temperature with AOO and postulated accident condition is < boiling at atmospheric consistent with the rest of the ARDC pressure. This might work for the language.

MHTGR; if so, it could be used in ARDC since it will work for fluid fueled MSRs as Common Cause Failures well. It would avoid the confusion of cold The NRC staff does not agree with this for these high temperature systems. comment. No change is necessary. The rationale states that secondary reactivity Suggested Change system (or mechanism) should be designed such that a common failure Page 64 of 105

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Consider using the definition of subcritical mode does not exist with the safety under cold conditions for all design related system(s) needed for ARDC 26 criteria (1) and (3).

ARDC Development References Achieving Cold shutdown The first paragraph of the rationale notes The NRC staff does not agree with this that the development of ARDC 26 was comment. No change is necessary.

informed by a number of references. Most Only one of the two reactivity systems of these references preceded the current specified by ARDC 26 needs to be version of the GDC. capable of achieving shutdown conditions that allow for interventions Suggested Change such as fuel loading, inspection and An explanation of how these older repair; but, a third reactivity system is references supported the changes from not precluded.

the current GDC would be helpful. Basis for Operational Requirement The NRC staff does not agree with this Use of Design-Basis Event comment. No change necessary as the Language requirement for a reactivity control It is not clear why the wording design- system necessary to preserve the basis event conditions is used explicitly fission product barriers has been in item (2) whereas postulated accidents included in the revised ARDC 26 (2).

is used consistently for the rest of the ARDC/SFR-DC/MHTGR-DC sets.

Suggested Change Either correct or explain inconsistency.

Common Cause Failures Suggest changing the Rationale discussion regarding diverse from different design than the safety related means to different design not subject to common cause failures.

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Suggested Change Suggest changing the Rationale discussion regarding diverse from different design than the safety related means to different design not subject to common cause failures.

Definition of Cold Shutdown Item (2) specifies safe shutdown whereas item 3 specifies reactor being subcritical under cold conditions. Safe shutdown state is defined in the rationale but a definition of cold shutdown is also needed (confusion might arise for some systems if the coolant is frozen at room temperature).

Suggested Change Suggest including a sentence in the rationale that cold conditions imply temperatures at which refueling, inspections, and repair functions can be performed.

Achieving Cold Shutdown It is not clear if item (3) calls for a third system/mechanism to render the reactor subcritical.

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A paragraph should be added in the rationale to clarify that the safety-related shutdown system is expected to achieve safe shutdown; but cold shutdown can be achieved by either a safety or non-safety shutdown system.

Basis for Operational Requirement The reference should be provided where the staff identified the requirement that the third sentence of GDC 26 is considered to be an operational requirement and not relevant as a DC.

Suggested Change The reference should be provided where the staff identified the requirement that the third sentence of GDC 26 is considered to be an operational requirement and not relevant as a DC.

The deletion of the list of postulated Positive comment, no change required.

reactivity accidents, leaving each design to determine its list of postulated reactivity 77 DOE/Lab MHTGR-DC 28 accidents, is a very good change.

Suggested Change Positive comment, no change suggested.

With the inclusion of AOOs within The NRC staff does not agree with this MHTGR GDC 20, 25, and 26, it is comment. ARDCs 20, 25, 26, and 29 78 Industry/NEI MHTGR-DC 29 recommended that this GDC is are related to the Protection and duplicative and can be deleted. Reactivity Control Systems. The protection requirements are covered by Page 67 of 105

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Suggested Change ARDC 20 and 25, and the reactivity Delete MHTGR-DC 29 controls are covered by ARDC 26. GDC 29 is related to protection against AOOs. AOOs are explicitly described in ARDC 20 and 26. The specific intent for GDC 29 is to design the protection and reactivity control systems in such a way that a safety function would be accomplished in the event of an AOO.

Similar comment as the one for SFR-DC The NRC staff does not agree with this

14. The definition of the primary coolant comment. See response to Comment boundary includes the cover gas no. 45 boundary. A cover gas leakage would lead to very limited safety consequences (no impact on the fission process, no impact or limited radiological consequences). This allows for safety valves on the cover gas system to limit abnormal pressure on the reactor vessel.

On the other hand, the failure of the 79 Industry/NEI SFR-DC 30 reactor vessel could have very severe consequences (e.g. reactivity insertion, failure of the core coolability).

Suggested Change It is therefore proposed to state that Each components that are is parts of the primary coolant boundary shall be designed, fabricated, erected and tested to the highest quality standards practical Page 68 of 105

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with high quality standards, consistent with its safety significance.

The NRC staffs addition of the last Positive comment, no change required.

sentence to this criterion is an excellent improvement.

80 DOE/Lab MHTGR-DC 30 Suggested Change Positive comment, no change suggested.

Concern Regarding Coolant Concern Regarding Coolant Chemistry Chemistry Item (2) adds and coolant chemistry Staff agrees. GE Hitachi submitted an to material property considerations. This informal public comment related to creates a degree of uncertainty. The SFR-DC 31, which stated, in part:

justification identifies unique potential coolants as a concern but chemistry GEH recommends adding infers a reactive property. Does this requirements for coolant chemistry and include secondary/tertiary reaction service degradation of properties, product interactions decedent from some creep, fatigue, and stress rupture to initial coolant chemistry? Are coolant address unique concerns of CRBRP contaminants considered in the criterion? because of the high design and 81 DOE/Lab ARDC 31 Coolant chemistry could be interpreted operating temperatures of the primary as a scope expansion and is unnecessary coolant boundary and the use of sodium given ARDC-14 requirements. as the coolant under NUREG-0968.

Based on this comment, the staff added the phrase coolant chemistry to ARDC Missing Words 31, SFR-DC 13, and MHTGR-DC 31 Proposed ARDC language seems to since each reactor type has high design accidentally drop the highlighted words in and operating temperatures, and unique item (2) The design shall reflect coolants.

consideration of service temperatures, service degradation of material The staffs addition of the phrase coolant chemistry was intended to Page 69 of 105

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properties These words properly include the current composition of the appear in SFR-DC 31 and GDC 31. coolant including fission products and contaminants.

Suggested Change None provided. Based on this formal public comment the staff proposes changing ARDC 31 to clarify the initial intent. The staff proposes revising ARDC 31 to state:

(2) the effects of irradiation and coolant composition, including contaminants and reaction products, on material properties, The staff proposes changing ARDC 31 rationale to state:

Specific examples are added to the MHTGR-DC to account for the high design and operating temperatures, coolant composition, contaminants, and reaction products.

The staff proposes changing SFR-DC 31 to state:

(2) the effects of irradiation and coolant composition, including contaminants and reaction products, on material properties The staff proposes changing SFR-DC 31 rationale to state:

Specific examples are added to the SFR-DC to account for the high design and operating temperatures, coolant Page 70 of 105

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composition, contaminants, and reaction products.

Missing Words The staff agrees with the second part of the comment and added as ARDC 31 contains the words consideration of so that ARDC 31 will state, The design shall reflect consideration of service temperatures, service degradation of material properties which is consistent with GDC 31 and SFR-DC 31.

Coolant Chemistry Agree. See response to Comment no.

The staff has added coolant chemistry 81 to item (2) in the criterion, and the second paragraph of the rationale refers to unique potential coolants. The working fluid in the modular HTGR is helium, which is chemically inert. Concerns regarding coolant chemistry in HTGRs pertain to the effects of contaminants on 82 DOE/Lab MHTGR-DC 31 material properties.

Suggested Change Item (2) in the criterion should be changed to, (2) the effects of irradiation and helium contaminants on material properties, The last three words of the rationale should be replaced with, potential helium contaminants.

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Addition of the word Functional Addition of the Word Functional For the replacement of testing with Functional testing is testing that functional testing; information should be assesses component and system added to the rationale to explain the intent operational readiness such as required behind the addition of the word in the ASME OM Code as incorporated ARDC 32, SFR- functional. The word is not included in by reference in 10 CFR 50.55a and in 83 DOE/Lab DC 32 GDC 32. What kind of functional testing is Plant Technical Specifications.

intended? What is the rationale for the Functional testing was added to ARDCs addition of this word? 32, 37, 40, 43, 46; SFR-DCs 32, 37, 40, 43, 46, 77; and MHTGR-DCs 32, 37, 46 Suggested Change None provided.

Addition of the Word Functional Addition of the Word Functional Replacement of testing with functional The NRC staff does not agree with this testing; information should be added to comment. Functional testing is testing the rationale to explain the intent behind that assesses component and system the addition of the word functional. The operational readiness such as required word is not included in GDC 32. in the ASME OM Code as incorporated by reference in 10 CFR 50.55a and in Suggested Change Plant Technical Specifications.

The rationale for the criterion (and for the ARDC and SFR criteria) does not address Leaktight vs. Allowable Leakage ARDC, SFR-DC, 84 DOE/Lab this change in wording and does not The NRC staff does not agree. GDC MHTGR-DC 32 explain what is intended by functional 32, and the other related GDCs, have a testing. Either an explanation should be requirement of inspecting and testing provided in the three rationales or, the RCS to ensure structural and preferably, the word functional should leaktight integrity. There is no current be deleted. industry definition of structural integrity of a piping system.

Leaktight vs. Allowable Leakage The inclusion of the words and leaktight Leak tightness is a critical part of the in the criterion is not necessary when philosophy of defense-in-depth. For an structural integrity is sufficient to LWR plant, operational leakage within Page 72 of 105

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describe the requirement. The allowable the Plant Technical Specifications is leak rate for a given design should be one permitted with understanding that the of the acceptance criteria for the test for system will be returned to leak-tight structural integrity. conditions during the next outage.

Suggested Change For the MHTGR specifically, the helium The words and leaktight should be pressure boundary may still be part of deleted here and in the ARDC and the the functional containment and SFR versions of this criterion. therefore credited in limiting radionuclide release. Therefore, the quality and leak tightness of the reactor helium pressure boundary may still serve a safety function. If the primary system is treated as a barrier for radionuclide retention than the baseline of the system should be leak-tightness.

A specific vendor can provide justification on the safety function of the helium pressure boundary that would impact the design and inspection requirements For advanced reactors, a leak may represent a significant safety hazard in a way different than LWRs. Because none of the design criteria specify requirements on the heat transfer medium for the systems, a leak in these system may contaminate the reactor coolant or containment environment. If a SFR or MHTGR containment had a significant amount of water due to a Page 73 of 105

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leak, an accident event may result in sodium fires or failure of the TRISO fuel beyond that which is expected.

The goal of GDC 33 is that the cooling The NRC staff does not agree with this function of the primary heat removal comment. The NRC staff believes that system shall not be impacted during ensuring the cooling functions of the normal operation by primary coolant primary and residual heat removal inventory loss due to leakage from the systems serve the same purpose as primary coolant boundary and rupture of ensuring that SADFL are not exceeded.

small piping or other small components The proper measure for adequacy of which are part of the boundary. For SFRs the cooling function is meeting the specifically, the primary concern is SAFDL.

ensuring primary coolant inventory is sufficient to maintain the cooling function for the primary heat removal system. This 85 Industry/NEI SFR-DC 33 ensures specified acceptable fuel design limits are not exceeded.

Suggested Change Replace the phrase specified acceptable fuel design limits are not exceeded with the phrase the cooling functions of the primary heat removal system and the residual heat removal system are not impacted. To eliminate redundancy, delete the phrase for protection against small breaks in the primary coolant boundary.

SFR-DC 34 deleted reference to The NRC staff does not agree with this postulated accidents (e.g. DBAs) without comment. No change is necessary.

86 Industry/NEI SFR-DC 34 an explanation in the rationale section. SFR-DC 34 addresses SAFDL protection consistent with the current Page 74 of 105

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Suggested Change GDC 34. SFR-DC 35 addresses Explain the reasoning for SFR-DC 34 residual heat removal under postulated being for normal operations and AOOs, accident conditions. The staff similar to the explanation provided for acknowledges that a single residual SFR-DC 35. heat removal system can be used to satisfy both SFR-DC 34 and DC 35.

This comment also applies to ARDC 34 and 35.

Passive vs. Active Residual Heat Passive vs. Active Residual Heat Removal Removal To ensure that the first line of the criterion The NRC staff does not agree with this is not interpreted as requiring that the comment. The word passive was residual heat removal system operate added, based on the definition of an passively during normal operations and MHTGR. The system may operate AOOs, the first paragraph of the rationale actively during normal operations but should note that the system may operate must operate passively during AOOs actively for heat removal during normal and postulated accidents. In definitions operations/AOOs, but that it shall operate Section 3.1 of the DOE report titled passively during postulated accidents. Guidance for Developing Principal Design Criteria for Advanced (Non-87 DOE/Lab MHTGR-DC 34 Suggested Change Light-Water) Reactors (Ref. 17), it is Note in the first paragraph of the rationale noted that the MHTGR design has a low that the system may operate actively for power density and hence residual heat heat removal during normal is removed by a passive system.

operations/AOOs, but that it shall operate passively during postulated accidents. Effective Core Cooling Agree. The second paragraph has been Effective Core Cooling modified to read:

In the second paragraph of this criterion, During postulated accidents, the system NRC staff has changed the words safety function shall provide effective effective cooling submitted by DOE/INL cooling.

to effective core cooling. DOE/INL used the words effective cooling because it is Page 75 of 105

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not just the core that needs to be Additionally Rationale will be modified effectively cooled during postulated to read:

accidents, but also structural components such as the core barrel and the reactor Effective cooling under postulated vessel. Effective cooling for these accident conditions is defined as components is needed to ensure that a maintaining fuel temperature limits passively coolable geometry is below design values to help ensure the maintained. siting regulatory dose limits criteria at the exclusion area boundary (EAB) and Suggested Change low-population zone (LPZ) are not Remove the word core from effective exceeded and the integrity of the core, core cooling. the core structural components, and the To explain the basis for changing reactor vessel is maintained under effective core cooling to effective postulated accident conditions, thereby cooling, the following paragraph should ensuring a geometry required for be added to the rationale: passive heat removal.

The modular HTGR residual heat removal system protects the Rationale for Ultimate Heat Sink integrity of the core, the core Staff does not agree with this comment.

structural components, and the Although MHTGRs do not have a reactor vessel when needed under cooling water system, the function that postulated accident conditions, the system provides, structural and thereby helping to ensure that the equipment cooling, may still be needed.

geometry required for passive heat MHTGR-DC 44-46 were included in removal is maintained. Therefore, DG-1330 as a reminder for reviewers to effective core cooling was verify that this function is accomplished replaced with effective cooling to in the design. Also, the MHTGR-DC reflect the broader range of states that, systems to transfer heat necessary cooling provided by the from structures systems and system during postulated components to an ultimate heat sink accidents. shall be provided, as necessary Rationale for Ultimate Heat Sink Definition of Effective Core Cooling Page 76 of 105

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The second paragraph of the rationale, Agree. See Effective Core Cooling which explains the basis for adding the above.

words ultimate heat sink to the criterion, is taken from the rationale for ARDC 34 that was provided in the original DOE/INL submittal. As it is written here, the second paragraph is tied to the possible need for a system like that addressed in GDC 44.

In the case of the modular HTGR version of the criterion, ultimate heat sink was added to the criterion by DOE/INL only for consistency with the ARDC and completeness, and the second paragraph was intentionally not included by DOE/INL in the modular HTGR DC 34 rationale.

The paragraph was not included because modular HTGRs, unlike LWRs, SFRs, and possibly other advanced non-LWRs, do not have or need a system that corresponds to the Cooling Water System that is required by GDC 44. The staff seems to have incorrectly assumed that the paragraph was omitted in error by DOE/INL and that the paragraph needs to be added to tie into a system like that addressed in GDC 44.

Suggested Change Delete the second paragraph from the modular HTGR rationale, and Criterion 44 and its associated criterion for inspection, etc. should be listed as Not Applicable to the modular HTGR.

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Definition of Effective Core Cooling The next to last paragraph of the rationale provides a definition of effective core cooling under postulated accident conditions. It is not clear why the staff has added this paragraph here but not done so in the ARDC or in the SFR DC.

For the modular HTGR, effective cooling is not just a matter of fuel temperature, but also of time at temperature. As it is written, this paragraph could be interpreted by future regulators as requiring a specific temperature limit, or a design value, under accident conditions.

Such a requirement would not be an accurate reflection of the effects of fuel temperature on coated particle fuel performance.

Suggested Change Delete the second to the last paragraph of the rationale should be deleted (preferred), or define effective cooling in the ARDC and SFR DC versions of Criterion 34.

ARDC 35 states A system to provide The NRC staff agrees with this sufficient emergency core cooling shall be comment. See resolution to number 90.

provided. The system safety function shall 88 Industry/NEI ARDC 35 be to transfer heat from the reactor core such that effective core cooling is maintained and fuel damage is limited.

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Regarding the addition of the words and fuel damage is limited to the first paragraph of the criterion, the rationale does not provide guidance for how these new words (which reflect an expansion relative to GDC 35) should be interpreted or why they have been added.

The added words are ambiguous when considering (1) to what level should fuel damage be limited? (2) What are the appropriate measures of fuel damage?

(3) How would fuel damage be interpreted for a molten salt reactor or for a modular HTGR?

Suggested Change It appears that the cited ARDC 35 text expands the scope of the existing GDC, and is therefore outside of the scope of this ARDC effort. Absent further information regarding the intent of these words, it is recommended that they be deleted from the criterion.

Reference to Fuel Damage The NRC staff agrees with this Regarding the addition of the words and comment. See resolution to number 90.

fuel damage is limited to the first ARDC 35, SFR- paragraph of the criterion, the rationale 89 DOE/Lab DC 35 does not provide guidance for how these new words (which reflect an expansion in scope relative to GDC 35) should be interpreted or why they have been added.

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The added words are ambiguous when considering (1) to what level should fuel damage be limited? (2) What are the appropriate measures of fuel damage?

(3) How would fuel damage be interpreted for a molten salt reactor or for a modular HTGR? It appears that the cited ARDC 35 text expands the scope of the existing GDC, and is therefore outside of the scope of this ARDC effort. Absent further information regarding the intent of these words, it is recommended that they be deleted from the criterion.

ARDC Missing Words Proposed ARDC language seems to accidentally drop the following highlighted words: The system safety function shall be to transfer heat from the reactor core at a rate such that effective core cooling is maintained.

Suggested Change None provided.

For SFRs, the residual heat removal NRC staff agrees with this comment.

system may be all that is required to ARDC 35 will be modified to read:

provide adequate heat removal during 90 Industry/NEI postulated accidents. A system to assure sufficient core SFR-DC 35 cooling during postulated accidents and SFR-DC 34 is specified as being to remove residual heat following applicable for normal and AOO postulated accidents shall be provided.

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conditions. However, residual heat The system safety function shall be to removal will also be necessary for transfer heat from the reactor core postulated accident conditions and should during and following postulated be addressed in SFR-DC 35. accidents such that fuel and clad damage that could interfere with The draft SFR-DC 35 added and fuel continued effective core cooling is damage is limited. Other than prevented.

maintaining effective core cooling, the meaning of this statement is not clear - The statement, and the design what is being prevented by limiting the conditions of the primary system fuel damage? Suggest using wording boundary are not exceeded was not similar to that used in GDC 35; that is use included as SFR-DC 35 and ARDC 35

. such that fuel and clad damage that allow for coolant makeup (injection) if could interfere with continued effective the primary system boundary fails core cooling is prevented. instead of similar to the current LWR LOCA.

. such that effective core cooling is maintained and fuel damage is limited SFR-DC 35 does not address protection of the primary coolant system boundary.

Add and the design conditions of the primary system boundary are not exceeded.

Suggested Change Replace the first paragraph of SFR-DC 35 with the following paragraph:

A system to assure sufficient core cooling during postulated accidents and to remove residual heat following postulated accidents shall be provided.

The system safety function shall be to Page 81 of 105

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transfer heat from the reactor core during and following postulated accidents such that fuel and clad damage that could interfere with continued effective core cooling is prevented and the design conditions of the primary system boundary are not exceeded.

Suggested Rationale Wording Change The NRC staff agrees with this The decision to classify Criterion 35 as comment. Change was incorporated.

not applicable to the modular HTGR is correct. However, the rationale cites the reactor power density and the core length-to-diameter ratio as the reasons that maintaining helium inventory is not needed. The power density and core geometry are only two of the reasons that might be listed. Others include, but are not limited to, high graphite heat capacity and the high temperature capability of the 91 DOE/Lab MHTGR-DC 35 fuel and the graphite.

Suggested Change Rather than trying to list all of the factors that apply, it would be better to revise the first sentence of the rationale as follows:

In the MHTGR design maintaining the helium inventory is not necessary to maintain effective cooling. Note that this suggested wording also deletes the word core, consistent with the comment on the rationale for modular HTGR DC 34.

Editorial Comment The NRC staff agrees with this 92 DOE Lab MHTGR-DC 36 Suggested Change comment. Change was incorporated.

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In the first line of the criterion, the word system should be inserted between the words removal and shall.

Add the word "system" after residual heat The NRC staff agrees with this removal. comment. Change was incorporated.

93 Industry/NEI MHTGR-DC 36 Suggested Change Add the word "system" after residual heat removal.

The title of these SFR-DC refers to the The NRC staff agrees with this residual heat removal system. The text comment. Change was incorporated.

that follows refers to the emergency core cooling system. While a single system may be provided to perform both residual heat removal and emergency core cooling functions, it would be logical for the title Industry/NEI and the text to use the same 94 SFR-DC 36 & 37 DOE/Lab nomenclature to describe the system.

Suggested Change Revise title of SFR-DC 36 to Inspection of emergency core cooling system.

Revise title of SFR-DC 37 to Inspection of emergency core cooling system.

Use of the Word Leaktight Use of the Word Leaktight Leaktight standards may not be The NRC staff does not agree with this necessary for certain advanced reactor comment. Same response as SSCs, but keeping this word in the Comment 84.

95 DOE/Lab ARDC 37 criterion infers expectation of leaktight capability. Determination of the degree to Title Change which a system is leaktight should be Agree. This change was made to subject to acceptance criteria that are ARDC, SFR-DC, and MHTGR-DC 37, appropriate for each reactor technology. and ARDC, SFR-DC, and MHTGR-DC Page 83 of 105

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36 related to inspection of emergency Suggested Change core cooling system.

The words and leaktight should be deleted. Connection Between Defense in Depth and System Leakage Title Change The NRC staff does not agree with this Title should read Testing of residual heat comment. Same response as removal emergency core cooling system. Comment 84.

Suggested Change As noted.

Connection Between Defense in Depth and System Leakage Additional clarification is needed in the rationale to explain the criterion that a non-leaktight system may be acceptable if defense in depth is not impacted by system leakage. This clarification applies to other criteria (e.g., ARDC 40, 43, and

46) that address defense in depth.

Suggested Change None provided.

Leaktight vs. Allowable Leakage Leaktight vs. Allowable Leakage As in MHTGR-DC 32, the inclusion of the The NRC staff does not agree with this word leaktight in the criterion is not comment. Same response as DOE/Lab necessary when structural integrity is Comment 84.

96 MHTGR-DC 37 sufficient to describe the requirement.

The allowable leak rate for a given design Air-Cooled vs. Water-Cooled RCCS should be one of the acceptance criteria The staff agrees with this comment. The for the test for structural integrity. In staff will add if applicable and delete Page 84 of 105

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particular, for the air-cooled variant of the the words in MHTGR-DC 37 that follow:

RCCS, the system is open and not relied upon during postulated leaktight at all. accidents Suggested Change The staff modified MHTGR-DC 37 as The words and leaktight should be follows:

deleted here and in the ARDC and the SFR versions of this criterion. (3) the operability of the system as a whole and, under conditions as close to Air-Cooled vs. Water-Cooled RCCS design as practical, the performance of Item (3) of the criterion addresses the full the full operational sequence that brings operational sequence that brings the the system into operation, including RCCS into operation, which is intended to associated systems, and, if applicable, include the transition from the normal any system(s) necessary to transition active operating mode to the passive from active normal operation to passive operating mode. The DOE/INL suggested mode for AOO or postulated accident text for this criterion included the words if decay heat removal to the ultimate heat applicable with this part of the criterion, sink.

but those words were omitted by the NRC staff. The words were proposed because Removal of Text from Rationale there are two possible designs of the Agree. The staff modified the 4th RCCS. The air-cooled design operates paragraph in the rationale of MHTGR-passively both during normal operating DC 37 to convey that including conditions and during postulated accident operation of associated systems conditions. There is no transition such as means testing any auxiliary or that intended to be described under Item secondary systems needed to perform (3) of the criterion. The water-cooled the passive residual heat removal design variant, on the other hand, function.

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described under Item (3) of the criterion is applicable.

Suggested Change Edit the beginning of the criterion item (3) to read as follows: the operability of the system as a whole and, if applicable, under conditions as close to design as practical, the performance of the full operational sequence It appears from the words at the end of the third paragraph of the rationale for this criterion that the NRC staff intended to include the words if applicable in the criterion, but they were inadvertently omitted.

Removal of Text from Rationale Also, at the end of Item (3), the NRC staff has added wording at the end of the item, relative to the DOE/INL proposed language, regarding operation of applicable portions of the protection system and the operation of the associated structural and equipment cooling water system. These words are not included in either the ARDC or SFR versions of Criterion 37, so the reasons for adding them only to the modular HTGR version of the criterion are not clear. The protection system does not play a role in operation of the RCCS.

Furthermore, as noted in comments above on modular HTGR DC 34, modular Page 86 of 105

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HTGRs, unlike LWRs, SFRs, and possibly other advanced non-LWRs, do not have or need a system that corresponds to the Cooling Water System that is required by GDC 44.

Suggested Change All words at the end of the criterion that follow relied upon during postulated accidents should be deleted.

It appears from the fourth paragraph of the rationale for this criterion that at one time there was also reference to power transfers, which are also not applicable to operation of the RCCS, which does not rely on electric power for its operation.

The fourth paragraph of the rationale should also be deleted.

The conclusion of the NRC staff that Positive comment, no change required.

these criteria are not applicable to the modular HTGR is appropriate. This comment also applies to MHTGR-DC 39 97 DOE/Lab MHTGR-DC 38 through MHTGR-DC 43.

Suggested Change Positive comment, no change suggested.

Use of the Word Leaktight Use of the Word Leaktight ARDC, SFR-DC Leaktight standards may not be The NRC staff does not agree with this 98 DOE/Lab 40, 43, 46 necessary for certain advanced reactor comment. Same response as SSCs but keeping it in the criterion infers Comment 84 Page 87 of 105

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expectation of leaktight capability.

Leaktight should be interpreted as a structural integrity element and subject to functional testing in that capacity.

Determination of the degree to which a system is leaktight should be subject to acceptance criteria that are appropriate for each reactor technology.

Suggested Change The words and leaktight should be deleted Additional Wording The NRC staff agrees with this First paragraph should end as to comment. Change was incorporated.

ensure that containment integrity and other safety functions are maintained. If the intent is to exempt SFR-DC 41 from the requirement for other safety ARDC 41, SFR-99 DOE/Lab functions, then Same as ARDC phrase DC 41 should be removed.

Suggested Change Add and other safety functions are maintained, to the end of the first paragraph The opening sentence is confusing. The NRC staff agrees with this comment. Sentence was changed to, Suggested Change A system to transfer heat from The opening sentence needs to be structures, systems, and components 100 Industry/NEI SFR-DC 44 revised to make its meaning clearer. important to safety to an ultimate heat sink shall be provided, as necessary, to transfer the combined heat load of these structures, systems, and Page 88 of 105

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components under normal operating and accident conditions.

Clarify important refers to important to The NRC staff does not agree with this safety comment. The term safety related is not used in the original GDCs or in the Suggested Change non-LWR design criteria. The language Change to The structural and equipment from the original GDC, important 101 Industry/NEI ARDC 45 cooling systems shall be designed to components, will be retained.

permit appropriate periodic inspection of safety related components, such as heat exchangers and piping, to ensure the integrity and capability of the systems.

Clarify applicability to SSCs with a safety The NRC staff does not agree with this function comment. The term safety related is not used in the original GDCs or in the 102 Industry/NEI ARDC 45 Suggested Change non-LWR design criteria. The language Change to The structural Safety Related from the original GDC, structural and structural and equipment cooling systems equipment cooling systems, will be shall be designed retained.

Cooling Water Systems The NRC staff does not agree with this As noted in comments on modular HTGR comment. Although MHTGRs do not DC 34 and 37, modular HTGRs (unlike have a cooling water system, the LWRs), SFRs, and possibly other function that the system provides, advanced non-LWRs, do not have or structural and equipment cooling, may need a system that corresponds to the still be needed. MHTGR-DC 44-46 MHTGR-DC 44, Cooling Water System that is required by were included in DG-1330 as a 103 DOE/Lab 45, 46 GDC 44. The DOE/INL comment in this reminder for reviewers to verify that this regard on MHTGR-DC 34 offers a function is accomplished in the design.

possible explanation of why NRC staff Also, the MHTGR-DC states that, seems incorrectly to believe otherwise. systems to transfer heat from The addition of the words as necessary structures systems and components to to the criterion is helpful, but relative to an ultimate heat sink shall be provided, the language in the rationale for this as necessary Page 89 of 105

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criterion, every design that is consistent with the definition of the modular HTGR contained in the DOE/INL submittal is designed such that the RCCS provides indefinite core cooling capability.

Suggested Change Criteria 44, 45, and 46 should be marked as Not Applicable to the modular HTGR.

Editorial: The example at the end of The NRC staff agrees with this subpart 1 of the ARDC GDC 50 is LWR comment. The last sentence in rationale specific is modified accordingly.

104 Industry/NEI ARDC 50 Suggested Change As indicated.

SFR structures are sensitive to pressure The NRC staff agrees with this and it may be chosen to avoid high comment. Change was incorporated.

pressure elevation in the containment Same as ARDC is removed since the design during leakage rate testing, in SFR-DC 52 now deviates from ARDC-order to preserve the facility and prevent 52.

undesirable over or under pressurization risks during those tests. It may be chosen 105 Industry/NEI SFR-DC 52 to perform those tests at a pressure below the containment design pressure, in order to extrapolate them at the containment design pressure (in this case the relevance of the extrapolation will of course have to be justified).

Suggested Change We propose to state that:

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The reactor containment structure and other equipment that may be subjected to containment test conditions shall be designed so that periodic integrated leakage rate testing can be conducted to demonstrate resistance at containment design pressure.

As indicated in criterion 57, an isolation of The NRC staff does not agree with this lines penetrating the reactor containment comment. SFR-DC-54 is written to structure may not be required in some provide the designer the opportunity to cases. This could for example could apply present the safety case without to the intermediate heat transport system containment isolation valves and the penetrating the reactor containment associated need for testing.

(provided adequate justification is given).

106 Industry/NEI SFR-DC 54 Suggested Change To ensure coherency of the text, this could be reflected in the Criterion 54:

Piping systems penetrating the reactor containment structure shall be provided with leak detection, isolation if necessary and containment capabilities ()

Why is Isolation valves outside The NRC staff agrees with this containment deleted? Its not deleted comment. It appears that this paragraph in 55. It appears from the wording that the was deleted accidentally. It was added intent was that this phrase NOT be back in.

107 Industry/NEI SFR-DC 56 deleted from SFR-DC 56. Deletion may SFR-DC 56 applies to instrumentation have been unintentional. lines etc. Therefore, its primary purpose is to assure the containment integrity.

Suggested Change The paragraph is intended to provide Add the wording to SFR-DC 56. that assurance.

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In several cases, the word reactor is The NRC staff agrees with this removed from reactor containment in comment. The word reactor has been recognition that containment is a barrier removed from ARDC 50-57.

between the fission products and the environment, yet reactor containment is ARDC 50-57 108 Industry/NEI retained in several other cases. (As an SFR-DC 50-57 example, ARDC 57 and SFR-DC differ in this regard) reactor (LWR) containment.

Suggested Change Consider removing reactor for consistency or explain the distinction.

The conclusion of the NRC staff that Positive comment, no change required.

these criteria are not applicable to the modular HTGR is appropriate. This MHTGR-DC 50- comment also applies to MHTGR-DC 51 109 DOE/Lab 57 through MHTGR-DC 57.

Suggested Change Positive comment, no change suggested.

Missing Wording The NRC staff agrees with this Following passage seems accidentally comment. Change was incorporated.

dropped from the end: confinement, and filtering systems, (4) with a residual heat removal capability having reliability and testability that reflects the importance 110 DOE/Lab SFR-DC 61 to safety of decay heat and other residual heat removal, and (5) to prevent significant reduction in fuel storage cooling under accident conditions.

Suggested Change Add missing wording.

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The first sentence, If an intermediate The NRC staff agrees with this coolant system is provided, then the comment. The first sentence of SFR-DC system shall be designed to transport 70 was deleted.

heat from the primary coolant system to the energy conversion system as required, is not required.

111 Industry/NEI SFR-DC 70 Suggested Change Rewrite the DC to state If an intermediate cooling system is provided, then the system shall be designed with sufficient margin Sodium freezing may not impact the The NRC staff agrees with this safety function of all systems. comment. Sentence was modified as shown in comment #113 Suggested Change 112 Industry/NEI SFR-DC 72 Add phrase ...if necessary to ensure that the safety function of the system is accomplished to the beginning of the first sentence.

Heating systems shall be provided for The NRC staff agrees with this systems and components important to comment. Change was incorporated.

safety, which contain or could be required to contain sodium, could be inferred to mean that all systems and components important to safety contain or could be 113 Industry/NEI SFR-DC 72 required to contain sodium.

Suggested Change To minimize confusion, restate as:

Heating systems shall be provided for systems and components that are Page 93 of 105

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important to safety, which and that contain or could be required to contain sodium.

Is the intent of the last sentence to ensure The NRC staff agrees with this that all sodium systems be in inerted comment. Sentence was modified as enclosures or guard vessels? Not all plant shown in comment #115.

systems containing sodium need to be in 114 Industry/NEI SFR-DC 73 inerted spaces.

Suggested Change Recommend deleting the last sentence.

Special features, such as inerted The NRC staff agrees with this enclosures or guard vessels, shall be comment. Change was incorporated.

provided for systems containing sodium.

implies a significant hazard exists for any system containing sodium.

115 Industry/NEI SFR-DC 73 Suggested Change Replace this sentence in its entirety with:

Systems from which sodium leakage constitutes a significant safety hazard shall include measures for protection, such as inerted enclosures or guard vessels.

Fire protection and mitigation due to NRC does not agree with this comment.

sodium water interaction is covered by The SFR-DC-73 addresses sodium SFR-DC 3 and SFR-DC 73. leakage and detection. SFR-DC 3 addresses consideration against fire in 116 Industry/NEI SFR-DC 74 Suggested Change general. SFR-DC-74 focuses on Delete phase , including mitigation of sodium-water reaction as a specific the effects of any resulting fire involving safety concern.

sodium.

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SFR-DC 70 states The intermediate NRC does not agree with this comment.

coolant system to be designed with See resolution to comments 122-127 sufficient margin to assure that (1) the design conditions of its boundary are not exceeded during normal operations and anticipated operational occurrences, and (2) the integrity of the primary coolant boundary is maintained during intermediate coolant system accidents.

SFR-DC 75, 76, and 77 are superfluous when evaluated in combination with the cited text from SFR-DC 70. SFR-DC 75, 76, and 77 appear to be applicable when the role of the intermediate coolant system is commensurate with a safety Industry/NEI 117 SFR-DC 75-77 function. However, other than the case DOE/Lab when it could serve as a path for decay heat removal, the intermediate coolant system does not have any safety function.

If the intermediate cooling system provides a safety-related heat removal capability, then SFR-DC 34-37 and SFR-DC 78 specify its requirements. The quality and fracture prevention requirements specified in SFR-DC 75 and 76 are supplementary requirements that are not consistent with the requirements for the decay heat removal and emergency core cooling systems specified in SFR-DC 34 and 35. Likewise, the inspection and testing requirements Page 95 of 105

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specified in SFR-DC 77 for the intermediate cooling system are contained in SFR-DC 36 and 37.

Therefore, for the case where the intermediate cooling system provides safety-related heat removal capability, SFR-DC 75, 76, and 77 are redundant and unnecessary.

If the intermediate cooling system does not provide safety-related heat removal capability, then only the requirements of SFR-DC 70 are necessary to specify the system design with appropriate margin to assure the design conditions of its boundary and the integrity of the primary coolant boundary. Therefore, for the case where the intermediate cooling system does not provide safety-related heat removal capability, SFR-DC 75, 76, and 77 are also redundant and unnecessary.

Suggested Change Recommend deletion of SFR-DC 75, 76, and 77. If SFR-DC 76 is not deleted, it should include wording such as commensurate with their importance to safety.

It is possible that there either be such a NRC does not agree with this comment.

configuration or that there be not be See resolution to comments 122-124 118 Industry/NEI SFR-DC 78 enough liquid metal to cause a severe and 126 consequence or even a significant consequence due to reactions with either Page 96 of 105

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air or water or both, both in terms of the reaction itself as well as consequence to the reactor and safety system functions.

Instead of being prescriptive, there needs to be a mechanistic method to determine whether multiple boundaries are necessary. Ultimately, the prescriptive condition for two boundaries is redundant; for both fluids and coolants which are compatible or incompatible, the required conditions should be the same, which are the conditions (1) and (2). So long as there is no failure of the intended safety functions of structures, systems or components important to safety or result in exceeding the fuel design limits, then the size of the reaction is small enough to justify not needing redundant boundaries.

Suggested Change Move the first sentence to the end with added wording described below:

After compatible in the second sentence, add or incompatible.

Add wording to the end to read: If the primary coolant system interfaces with a structure, system, or component containing fluid that is chemically incompatible with the primary coolant, and cannot meet Page 97 of 105

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condition (1) and condition (2), the interface location shall be designed to ensure that the primary coolant is separated from the chemically incompatible fluid by two redundant, passive barriers.

The requirement to ensure that primary The NRC staff does not agree with this coolant sodium limits are not exceeded comment. SFR-DC 71 is related to as a result of cover gas leakage are maintaining the purity of the primary already addressed in SFR-DC 71, item coolant and cover gas to ensure that (4). the primary coolant sodium limits are 119 Industry/NEI SFR-DC 79 not exceeded.

Suggested Change Delete SFR-DC 79 SFR-DC 79 is related to the makeup system for the cover gas similar to the makeup system for primary coolant that is described in SFR-DC 33.

The word passive implies that only a The NRC staff does not agree with this passive system is to be provided. comment. The word passive was Maintaining geometry is needed for both used based on the specific definition of active and passive means of heat an MHTGR. It is not intended to be MHTGR-DC 34 removal. applicable to a broader class of HTGR 120 Industry/NEI MHTGR-DC 71 Note that proposed new MHTGR-DC 72 designs that may utilize active or MHTGR-DC 72 does not mention passive (while the passive systems.

rationale does).

Suggested Change Remove the word passive The wording adopted by the staff for Positive comment, no change required.

MHTGR-DC 70-121 DOE/Lab these criteria is correct and consistent 72 with the modular HTGR approach to Page 98 of 105

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safety design. This comment also applies to MHTGR-DC 71 and MHTGR-DC 72.

Suggested Change Positive comment, no change suggested.

122 DOE/Lab SFR-DC-75 implies that the ICS may The NRC staff does not agree with this perform a safety function. What safety comment. 10 CFR Part 50, 10 CFR functions are envisioned other than the Part 52, Regulatory Guides, and potential decay heat removal path that is NUREGs use the terms Important to covered in SFR-DC 34-37? Safety, Safety Related, Safety Functions, etc. to discuss specific Suggested Change regulatory requirements and regulatory Delete SFR-DC 75 scope. An applicant may choose to use the ICS system as a heat removal system, but may not choose to credit the ICS system as a RHR or ECCS system.

This may be done to add defense-in-depth or to lower the risk of the plant.

SFR-DC 75 The term important to safety describes a larger scope of systems than just safety-related components. As stated in the November 20, 1981 memo Standard Definitions for Commonly-Used Safety Classification Terms (Memo is contained in ADAMS Accession No. ML031150515):

Important to Safety is defined as, Those structures, systems, and components that provide reasonable assurance that the facility can be operated without Page 99 of 105

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undue risk to the health and safety of the public.

Additionally, Important to Safety Encompasses the broad class of plant features, covered (not necessarily explicitly) in the General Design Criteria that contribute in important way to safe operation and protection of the public in all phases and aspects of facility operation (i.e.

normal operation and transient control as well as accident mitigation).

Finally, Important to Safety Includes safety-related as a subset As described in SECY-04-109, deterministically evaluated Important to Safety components some share characteristics with risk-informed RTNSS systems. A vendor may decide to credit cooling functions to the ICS as a RTNSS system. In this case, the ICS would be expected to be designed, fabricated, erected, and tested to a quality standard commensurate with the risk-significance of the safety function performed. The staff supplemented the rationale for SFR-DC 75 to further clarify Page 100 of 105

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commensurate with the systems importance to safety.

123 DOE/Lab SFR-DC-76 does not include the phrase The NRC staff does not agree with this commensurate with the importance of comment. The staff intentionally did not the safety functions as SFR-DC-75 and include the words commensurate with 77 do. Doesnt this exclusion imply that the importance of the safety functions ICS has be a safety grade system even in this criterion.

though it may not serve any safety function? If the concern is sodium fires, The verbiage is not intended to imply isn't it already addressed in SFR-DC-73 that the ICS will serve a safety function.

which requires "sodium leakage detection The SFR-DC 76 wording implies that and reaction prevention and mitigation"? the sodium within the system has an inherent relation to the safety of the Suggested Change plant. A sudden rupture of the ICS Delete SFR-DC 76 would result in massive sodium-air, -

water, or -concrete reactions and would constitute a risk to the safe operation of SFR-DC 76 the plant and challenge the integrated safety of the plant.

The staff asserts that requiring the intermediate system (that is filled with sodium or NaK) to not fail in a brittle manner or rapidly propagate a fracture is necessary engineering practice. The staff modified the rationale for SFR-DC 76 to clarify this.

Considering the low pressure and high temperatures of the ICS, typical materials (stainless steels) should have sufficient ductility to meet SFR-DC 76 Page 101 of 105

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without modification to ASME/ASTM specifications.

The staff does not believe that SFR-DC 73 and SFR-DC 76 overlap. SFR-DC 73 provides criteria for detecting and mitigating leakage of systems. SFR-DC 76 ensures that gross failure of the ICS is prevented as part of the design of the plant.

124 DOE/Lab SFR-DC-77 imposes inspection, testing, The NRC staff does not agree with this and leak detection requirements for the comment. SFR-DC-73 does not have a ICS. In what way are these specific ICS requirement for periodic inspections or requirements differ from the sodium testing of sodium systems. The staff leakage detection and reaction prevention added, and identify sodium leakage as and mitigation requirements already practical, to SFR-DC 73 and removed it SFR-DC 77 covered under SFR-DC-73? from SFR-DC 77.

Suggested Change SFR-DC 77 will be retained and Delete SFR-DC 77 additional discussion added to the rationale to clarify why inspection, testing and material surveillance is needed.

125 DOE/Lab The NRC rationale in slide 54 of the The distinction is stated in the August 24, 2017 Public Meeting Slides November 20, 1981 memo Standard (ML17233A213 first bullet) seems to Definitions for Commonly-Used Safety suggest a distinction between safety Classification Terms (Memo is SFR-DC 77 related and important to safety. Can contained in ADAMS Accession you clarify this distinction? ML031150515).

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126 DOE/Lab The NRC rationale in slide 54 of the This statement was made to clarify that August 24, 2017 Public Meeting Slides the ICS system may have other (ML17233A213 second bullet) seems to regulatory functions other than those suggest that ICS leakage can have discussed as part of the advanced impact on other aspects (post-accident reactor design criteria. An ICS that is recovery). Isnt this concern already not credited for residual heat removal or addressed in SFR-DC-73 for sodium emergency core cooling (which are leakage detection and reaction prevention systems associated with normal, SFR-DC 70 and mitigation? abnormal, and accident conditions) may be credited to reduce risk or increase Suggested Change defense-in-depth. This logic is None provided consistent with the introductory paragraphs of 10 CFR Part 50 Appendix A. SFR-DC 70 only describes functions that the ICS needs to meet during normal, abnormal, and accident conditions.

127 DOE/Lab In general, an ICS leakage (without The NRC staff does not agree with this activated coolant) is an operational comment. The staff would need more concern (unless the ICS performs a information on a SFR design to decay heat removal function). Do SFR- determine the scope of design basis DC-75-77 imply that any ICS leakage will accidents.

be treated as a postulated accident?

If ICS leakage would result in a credible SFR-DC 75-77 Suggested Change challenge to the public health and safety Delete SFR-DC 75-77 or represent a significant hazard to the environment, then the designer should consider ICS leakage as a postulated accident. However, the staff does not believe that ICS leakage would rise to this level.

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The safety significance of an individual system does not necessarily determine whether the failure of that system should or should not be considered a postulated accident. LWR designs have postulated accidents which are caused by non-safety systems (flooding, fires, main steam or main feedline breaks, etc.). For a LWR pressurized water reactor, the reactor coolant pump seals have a significant impact on plant safety even though the reactor coolant pumps have limited safety significance.

The selection of licensing basis events is outside the scope of this regulatory guide. The staff suggests that interested members of the public should monitor the progress on the following two projects:

The NRC is supporting activities related to the Licensing Modernization Project (LMP) being led by Southern Company, coordinated by NEI, and cost-shared by DOE. The NRC is currently reviewing LMP white paper Modernization of Technical Requirements for Licensing of Advanced Non-Light Water Reactors - Selection of Page 104 of 105

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Licensing Basis Events, Draft Report.

NRC Non-LWR Vision and Strategy and Implementation Action Plans, Near-Term Strategy 3 includes Contributing Activity No. 3.2: Determine and document appropriate non-LWR licensing bases and accident sets for highly prioritized non-LWR technologies.

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