ML16301A307

DG-1330 Guidance for Developing Principal Design Criteria for Non-Light Water Reactors Jan 2017

ML16301A307
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Issue date:	01/31/2017
From:	Mazza J NRC/NRO/DEIA/ARPB
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Orr M
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DG-1330 DG-1330
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U.S. NUCLEAR REGULATORY COMMISSION February 2017 OFFICE OF NUCLEAR REGULATORY RESEARCH Division 1 DRAFT REGULATORY GUIDE Technical Lead Jan Mazza DRAFT REGULATORY GUIDE DG-1330 (Proposed New Regulatory Guide 1.232)

GUIDANCE FOR DEVELOPING PRINCIPAL DESIGN CRITERIA FOR NON-LIGHT WATER REACTORS A. INTRODUCTION Purpose 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 reactor designers, applicants, and licensees to develop principal design criteria (PDC) for any non-LWR designs, as required by the applicable NRC regulations. The RG also 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).

Applicability This RG applies to reactor designers, applicants, and licensees of non-LWR designs subject to 10 CFR Part 50 and 10 CFR Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants (Ref. 2).

Applicable Regulations

• 10 CFR Part 50 provides regulations for licensing production and utilization facilities.

o 10 CFR Part 50, Appendix A, contains the GDC that establish the minimum requirements for the PDC for water-cooled nuclear power plants. Appendix A also establishes that the GDC are generally applicable to other types of nuclear power units and are intended to provide guidance in determining the PDC for such other units.

This RG is being issued in draft form to involve the public in the development of regulatory guidance in this area. It has not received final staff review or approval and does not represent an NRC final staff position. Public comments are being solicited on this draft guide and its associated regulatory analysis. Comments should be accompanied by appropriate supporting data. Comments may be submitted through the Federal rulemaking Web site, http://www.regulations.gov, by searching for draft regulatory guide DG-1330. Alternatively, comments may be submitted to the Rules, Announcements, and Directives Branch, Office of Administration, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001. Comments must be submitted by the date indicated in the Federal Register (FR) notice.

Electronic copies of this draft regulatory guide, previous versions of this guide, and other recently issued guides are available through the NRCs public Web site under the Regulatory Guides document collection of the NRC Library at http://www.nrc.gov/reading-rm/doc-collections/reg-guides/. The draft regulatory guide is also available through the NRCs Agencywide Documents Access and Management System (ADAMS) at http://www.nrc.gov/reading-rm/adams.html, under Accession No. ML16301A307. The regulatory analysis may be found in ADAMS under Accession No. ML16330A179.

o 10 CFR 50.34(a)(3)(i) requires that an application for a construction permit include the PDC for a proposed facility.

• 10 CFR Part 52 governs the issuance of early site permits, standard design certifications, combined licenses, standard design approvals, and manufacturing licenses for nuclear power facilities.

o 10 CFR 52.47(a)(3)(i) requires that an application for a design certification include the PDC for a proposed facility.

o 10 CFR 52.79(a)(4)(i) requires that an application for a combined license include the PDC for a proposed facility.

o 10 CFR 52.137(a)(3)(i) requires that an application for a standard design approval include the PDC for a proposed facility.

o 10 CFR 52.157(a) requires that an application for a manufacturing license include the PDC for a proposed facility.

Related Guidance, Communications, and Policy Statements

• NUREG-1338, Draft Preapplication Safety Evaluation Report for the Modular High-Temperature Gas-Cooled Reactor (MHTGR), issued December 1995, provides the NRC staffs review and insights on the mHTGR design (Ref. 3).

• NUREG-1368, Preapplication Safety Evaluation Report for the Power Reactor Innovative Small Module (PRISM) Liquid Metal Reactor, issued February 1994, provides the NRC staffs review and insights on the design for the GE-Hitachi PRISM liquid-metal reactor (LMR) (Ref. 4).

• NUREG-0968, Safety Evaluation Report Related to the Construction of the Clinch River Breeder Reactor Plant, issued March 1983, provides the staffs evaluation of the Clinch River construction permit application (Ref. 5).

• NUREG-1369, Preapplication Safety Evaluation Report for the Sodium Advanced Fast Reactor (SAFR) Liquid-Metal Reactor, issued December 1991, provides the NRC staffs review and insights on the SAFR design (Ref. 6).

• SECY-93-092, Issues Pertaining to the Advanced Reactor (PRISM, mHTGR, and PIUS) and CANDU 3 Designs and their Relationship to Current Regulatory Requirements, dated April 8, 1993, provides staff insights on issues pertaining to advanced designs and proposes resolutions (Ref. 7).

• Staff Requirements Memorandum (SRM)-SECY-93-092, Issues Pertaining to the Advanced Reactor (PRISM, mHTGR, and PIUS) and CANDU 3 Designs and their Relationship to Current Regulatory Requirements, issued July 1993, provides the Commission position on topics discussed in SECY-93-092 (Ref. 8).

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• SECY-03-0047, Policy Issues Related to Licensing Non-Light-Water Reactor Designs, dated March 28, 2003, provides, for Commission consideration, options and recommended positions for resolving the seven policy issues associated with the design and licensing of future non-LWR designs (Ref. 9).

• SRM-SECY-03-0047, Policy Issues Related to Licensing Non-Light-Water Reactor Designs, issued June 26, 2003, provides the Commission position on the topics discussed in SECY-03-0047 (Ref. 10).

• NRC, Next Generation Nuclear 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).

• NRC, Policy Statement on the Regulation of Advanced Reactors (73 FR 60612, October 14, 2008), establishes the Commissions expectations related to advanced reactor designs to protect the environment and public health and safety and promote the common defense and security with respect to advanced reactors (Ref. 12).

Purpose of Regulatory Guides The NRC issues RGs to describe to the public methods that the staff considers acceptable for use in implementing specific parts of the agencys regulations, to explain techniques that the staff uses in evaluating specific problems or postulated events, and to provide guidance to applicants. Regulatory guides are not substitutes for regulations and compliance with them is not required. Methods and solutions that differ from those set forth in RGs will be deemed acceptable if they provide a basis for the findings required for the issuance or continuance of a permit or license by the Commission.

Paperwork Reduction Act This RG contains information collection requirements that are subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et seq.). These information collections were approved by the Office of Management and Budget (OMB), approval number 3150-0011 and 3150-0151.

Public Protection Notification The NRC may not conduct or sponsor, and a person is not required to respond to, a request for information or an information collection requirement unless the requesting document displays a currently valid OMB control number.

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CONTENTS A. Introduction ............................................................................................................................................. 1 B. Discussion ............................................................................................................................................... 5 C. Staff Regulatory Guidance .................................................................................................................... 11 D. Implementation ..................................................................................................................................... 20 Acronyms .................................................................................................................................................... 22 References ................................................................................................................................................... 23 Appendix A. Advanced Reactor Design Criteria ..................................................................................... A-1 Appendix B. Sodium-Cooled Fast Reactor Design Criteria .................................................................... B-1 Appendix C. Modular High-Temperature Gas-Cooled Reactor Design Criteria ..................................... C-1 DG-1330, Page 4

B. DISCUSSION Reason for Issuance This revision (Revision 0) provides guidance for developing PDC for non-LWRs. Applications for a construction permit, design certification, combined license, standard design approval, or manufacturing license are required by 10 CFR 50.34(a)(3), 10 CFR 52.47(a)(3)(i), 10 CFR 52.79(a)(4)(i),

10 CFR 52.137(a)(3)(i), and 10 CFR 52.157(a), respectively, to include the PDC for the facility in their applications.

Background

The NRC Regulatory Framework In accordance with its mission, the NRC protects public health and safety and the environment by regulating the design, siting, construction, and operation of commercial nuclear power facilities. The NRC conducts its reactor licensing activities through a combination of regulatory requirements and guidance. The applicable regulatory requirements are found in Chapter I of Title 10, Energy, of the Code of Federal Regulations, Parts 1 through 199. Regulatory guidance is additional detailed information on specific acceptable means to meet the requirements in regulation. Guidance is provided in several forms, such as in RGs, interim staff guidance, standard review plans, NUREGs, review standards, and Commission policy statements. These regulatory requirements and guidance represent the entirety of the regulatory framework that an applicant should consider when preparing an application for review by the NRC. A key part of the regulatory requirements is in the general design criteria (GDC) in Appendix A to 10 CFR Part 50. These high-level GDC requirements support the design of the current nuclear power plants and are addressed in 10 CFR 50.34, Contents of Applications; Technical Information. Because the current GDC are based on LWR technology, the NRC developed the non-LWR design criteria, included as appendices to this RG, to provide guidance for developing PDC for non-LWR technology.

The nuclear power plants presently operating in the United States were licensed under the process described in 10 CFR Part 50. The NRC and its predecessor, the U. S. Atomic Energy Commission (AEC),

approved construction permits for these plants between 1964 and 1978 and granted the most recent operating license under 10 CFR Part 50 in 2015. The regulations in 10 CFR Part 50 evolved over the years to address specific safety issues discovered as a result of operating experience and industry events.

Some examples include fire protection in 10 CFR 50.48, emergency plans in 10 CFR 50.47, and aircraft impact assessment in 10 CFR 50.150. The NRC applied some of these new regulations retroactively to operating reactors while applying others only to new reactors.

The NRC used its experience in licensing the current nuclear power plants to develop 10 CFR Part 52, which it issued in 1989 and has used for the most recent new reactor licensing reviews, reactor design certifications, and early site permits. The regulations in 10 CFR Part 52 apply lessons learned from licensing the operating reactors, provide an alternative to the current process described in 10 CFR Part 50, and increase the standardization of the next generation of nuclear power plants. For many years, new nuclear power plant licensing and guidance development activities have focused on the licensing processes in 10 CFR Part 52, rather than those in 10 CFR Part 50. For this reason, some Commission decisions regarding new nuclear power plant licensing issues have been incorporated into 10 CFR Part 52, without similar requirements consistently being incorporated into 10 CFR Part 50. For example, 10 CFR Part 52 includes requirements derived from the Commission Policy Statement on Severe Reactor Accidents Regarding Future Designs and Existing Plants (Ref. 13), with explicit requirements related to the Three Mile Island items in 10 CFR 50.34(f), severe accidents, probabilistic risk assessment, and other topics, whereas no similar requirements have been incorporated for new DG-1330, Page 5

10 CFR Part 50 nuclear power plant applications. In response to recent industry interest in employing the 10 CFR Part 50 process for new designs, SECY-15-0002, Proposed Updates of Licensing Policies Rules, and Guidance for Future New Reactor Applications (Ref. 14), was written to request that the Commission confirm that its policies and requirements apply to all new nuclear power plant applications, regardless of the selected licensing approach. In the SRM to SECY-15-0002 (Ref. 15), the Commission approved the staffs recommendation to revise the regulations in 10 CFR Part 50 and Part 52 for new power reactor applications to reflect lessons learned from recent new reactor licensing activities and to more closely align with each other.

Role of the General Design Criteria in the Regulatory Framework As mentioned above, the GDC contained in Appendix A to 10 CFR Part 50 are an important part of the NRCs regulatory framework. For LWRs, they provide minimum requirements for PDC, which establish the necessary design, fabrication, construction, testing, and performance requirements for structures, systems, and components (SSCs) that are important to safety; that is, as stated in Appendix A, SSCs that provide reasonable assurance that the nuclear power plant can be operated without undue risk to the health and safety of the public. 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 postulated to occur during normal operations, anticipated operational occurrences (AOOs), and postulated accidents.

NRC Policy on Advanced Reactors From the NRC staffs regulatory perspective, the characteristics of an advanced reactor have evolved over time, and this evolution is expected to continue. For example, the passive features in the AP1000 design were advanced concepts when first introduced in 2002. On October 14, 2008, the Commission issued its most recent policy statement on advanced reactors, Policy Statement on the Regulation of Advanced Reactors, which included items to be considered in their designs. The Commissions 2008 policy statement reinforced and updated the policy statements on advanced reactors previously published in 1986 and 1994. In part, the 2008 update to the policy states the following:

Regarding advanced reactors, the Commission expects, as a minimum, at least the same degree of protection of the environment and public health and safety and the common defense and security that is required for current generation light-water reactors [i.e., those licensed before 1997]. Furthermore, the Commission expects that advanced reactors will provide enhanced margins of safety and/or use simplified, inherent, passive, or other innovative means to accomplish their safety and security functions.

The Advanced Reactor Policy Statement makes clear the Commissions expectations that advanced reactor designs will address all current regulations, including those related to severe accidents, beyond-design-basis accidents, defense in depth, and probabilistic risk assessment requirements.

Depending on the design attributes of the different non-LWR technologies, the NRC regulations and policies may be addressed in a different manner than for traditional LWRs.

Role of the General Design Criteria for Non-LWRs As discussed in Section A of this RG, applications for a construction permit, design certification, combined license, standard design approval, or manufacturing license, respectively, must include the PDC for the facility. The PDC are derived from the GDC in Appendix A to 10 CFR Part 50.

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Title 10 CFR 50.34 1 states:

Appendix A to 10 CFR part 50, general design criteria (GDC), establishes minimum requirements for the principal design criteria for watercooled nuclear power plants similar in design and location to plants for which construction permits have previously been issued by the Commission and provides guidance to applicants in establishing principal design criteria for other types of nuclear power units.

Appendix A to 10 CFR part 50 states:

These General 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. 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.

Together, these requirements recognize that different requirements may be necessary for non-LWR designs. The non-LWR design criteria developed by the NRC staff and included in Appendices A to C of this regulatory guide, are intended to provide stakeholders with insight into the staffs views on how the GDC could be interpreted to address non-LWR design features; however, these are not considered to be final or binding regarding what may eventually be required from a non-LWR applicant.

It is the applicants responsibility to develop the PDC for its facility based on the specifics of its unique design, using the GDC, non-LWR design criteria, or other design criteria as the foundation. Further, the applicant is responsible for considering public safety matters and fundamental concepts, such as defense in depth, in the design of their specific facility and for identifying and satisfying necessary safety requirements.

The non-LWR design criteria are an important first step to address the unique characteristics of non-LWR technology. The NRC recognizes the benefits to risk informing the non-LWR design criteria to the extent possible, depending on the design information and data available. The NRCs draft Vision and Strategy: Safely Achieving Effective and Efficient Non-Light-Water Reactor Mission Readiness (Ref. 16) outlines mid- and long-term activities to develop, as necessary, a risk-informed, performance-based non-LWR regulatory framework. Implementing the mid- and long-term Implementation Action Plans as part of the Vision and Strategy activities will help NRC determine whether risk informed non-LWR design criteria should be included as part of a new regulatory framework.

DOE-NRC Initiative Phase 1 In July 2013, the NRC and U.S. Department of Energy (DOE) established a joint initiative to address a key element in the regulatory framework that could apply to non-LWR technologies specifically, to address the existing GDC, which may not directly apply to non-LWR power plant designs.

The purpose of the initiative is to assess the GDC to determine whether they apply to non-LWR designs and, if not, to propose the PDC that address non-LWR design features while recognizing that the underlying safety objective of each GDC still applies.

The assessment of the GDC with respect to non-LWR designs was accomplished in two phases.

Phase 1 was managed by a team including representatives of the DOE and its national laboratories, and 1 Similar language is included in 10 CFR 52.47(a)(3)(i), 10 CFR 52.79(a)(4), 10 CFR 52.137(a)(3), and 10 CFR 52.157(a).

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consisted of reviews and evaluations of applicable technical information. The DOE team reviewed information related to six different types of non-LWR technologies (i.e., sodium-cooled fast reactors (SFRs), lead fast reactors (LFRs), gas-cooled fast reactors (GCRs), modular high-temperature gas-cooled reactors (mHTGRs), fluoride high-temperature reactors (FHRs), and molten-salt reactors (MSRs)). Using this information, DOE then reviewed the existing NRC GDC to determine their applicability and whether they should be modified to reflect non-LWR designs.

The results of DOEs assessment are contained in a DOE report titled, Guidance for Developing Principal Design Criteria for Advanced (Non-Light Water) Reactors. DOE submitted this report to the NRC for consideration in December 2014 (Ref. 17). In it, DOE proposed a set of advanced reactor design criteria (ARDC), which could serve the same purpose for non-LWRs as the GDC serve for LWRs. The ARDC are intended to be technology neutral and, therefore, could apply to any type of non-LWR design.

In addition to the technology-neutral ARDC, DOE proposed two sets of technology-specific, non-LWR design criteria. These criteria are intended to apply to SFRs and mHTGRs and are referred to as the SFR design criteria (SFR-DC) and the mHTGR design criteria (mHTGR-DC), respectively. The DOE developed the technology specific design criteria to demonstrate how the GDC could be adapted to specific technologies in which there was some level of maturity and documented design information available. DOE determined that the safety objectives for some of the current GDC did not address design features specific to SFR and mHTGR technologies (e.g., sodium or helium coolant, passive heat removal systems, etc.). Additional design criteria were developed to address unique features of those designs.

DOE-NRC Initiative Phase 2 After DOE issued its report in December 2014, an NRC multidisciplinary team was assembled to review the report, other pertinent references, and NRC documents, such as NUREGs, reports, and white papers. The NRC held a public meeting on January 21, 2015, to discuss the report with DOE and to describe NRCs plans to develop regulatory guidance for non-LWR reactor design criteria (Ref. 18).

During its review, the NRC staff formulated questions and clarifications necessary to obtain a full understanding of the design aspects of the non-LWR technologies and the reasoning that DOE employed in developing its proposal for the ARDC, SFR-DC, and mHTGR-DC. The following documents contain the NRC questions and DOE responses:

• NRC Staff Questions on the DOE Report, Guidance for Developing Principal Design Criteria for Advanced Non-Light Water Reactors, dated June 5, 2015, and Response to NRC Staff Questions on the U.S. Department of Energy Report, Guidance for Developing Principal Design Criteria for Advanced Non-Light Water Reactors, dated July 15, 2015 (Ref. 19 for both), and

• Questions on the U.S. Department of Energy Report, Guidance for Developing Principal Design Criteria for Advanced Non-Light Water Reactors, dated August 17, 2015, and Response to NRC Staff Questions on the U.S. Department of Energy Report, Guidance for Developing Principal Design Criteria for Advanced Non-Light Water Reactors, dated September 15, 2015 (Ref. 20 for both).

After consideration of the DOE report, DOE responses to NRC staff questions, and other applicable information relevant to the NRC regulatory philosophy and current understanding of non-LWR designs, the NRC developed its own version of the ARDC, SFR-DC, and mHTGR-DC. While reviewing the DOE report, NRC staff considered whether to develop one generic set of non-LWR design criteria or to follow the DOE model and develop the technology specific design criteria as well. After considering the diversity of the design features for the two mature technologies, the NRC staff chose to develop the SFR-DC and mHTGR-DC in addition to the ARDC.

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The NRC issued a draft version of design criteria for informal public comment titled Invitation and Instructions for Public Comment, on April 7, 2016 (Ref. 21). The NRC staff noted in the introductory material of this invitation that comments received would not be responded to individually but would be considered by the NRC staff when developing the draft RG. By June 8, 2016 the NRC received over 350 public comments from over 20 stakeholder organizations (Ref. 22). The NRC held a public meeting to discuss the public comments on October 11, 2016 (Ref. 23). The tables in Appendices A, B, and C of this RG represent the staffs second draft version of the design criteria that incorporates many of the informal public comments.

Key Assumptions and Clarifications Regarding the non-LWR Design Criteria The NRC staff applied the following key assumptions when developing the non-LWR design criteria:

• The underlying safety objectives of the GDC still apply.

• ARDC, SFR DC, and mHTGR DC apply to normal operations, anticipated operational occurrences, and postulated accidents (design basis).

• The NRC has regulations and orders on severe accidents and beyond-design-basis events (BDBEs) for LWRs. Similar regulations for non-LWRs were not defined as part of this initiative.

The current regulations may or may not be applicable to non-LWRs. It is the responsibility of the applicant to demonstrate compliance with applicable severe accident and BDBE regulations and orders, demonstrate why any that are not applicable do not apply, and demonstrate why other design specific severe accidents or BDBE that can occur will be mitigated.

• While developing the non-LWR design criteria, the staff assumed that a core disruptive accident will be demonstrated to be a severe accident or a BDBE by the applicant. A core disruptive accident would result in a loss of a coolable geometry such that multiple non-LWR design criteria would be violated.

• Safety design objectives for non-LWRs can differ substantially from those associated with LWRs.

• Proposed GDC adaptations were focused on those needed for improved regulatory certainty and clarity.

• 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 or technology than the ARDC.

• The SFR-DC and mHTGR-DC are intended to apply to all designs of these technologies.

Additional sets of technology-specific design criteria (e.g., MSRs, LFRs) may be developed in the future as more information about the designs becomes available.

• Non-LWR designs should provide enhanced margins of safety and/or use simplified, inherent, passive, or other innovative means to accomplish their safety and security functions.

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Harmonization with International Standards The International Atomic Energy Agency (IAEA), in collaboration with the International Project on Innovative Nuclear Reactors and Fuel Cycles and the Generation IV International Forum, established the Sodium-Cooled Fast Reactor Task Force. The SFR Task Force is collaborating with international designers, government organizations, and regulators to develop safety design criteria and safety design guidelines for SFRs. The NRC will continue to monitor and collaborate on these documents and consider using them to the extent practical in developing SFR design criteria.

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C. STAFF REGULATORY GUIDANCE This section contains information on the intended use of the RG. It also contains NRC staffs determination of the applicability of each GDC to the non-LWR design criteria. This is illustrated in the table titled, Table 1: Non-Light-Water Reactor Crosswalk. The actual ARDC, SFR-DC, and mHTGR-DC and NRC staff technology-specific rationale for adaptions to the GDCs to develop the PDC are contained in Appendices A-C to this RG.

Intended Use of This Regulatory Guide This RG provides guidance to reactor designers, applicants, and licensees of non-LWR designs for developing PDC. Non-LWR applicants would not need to request an exemption from the GDC in 10 CFR Part 50 when proposing PDC for a specific design.

Applicants may use this RG to develop all or part of the PDC and are free to choose among the ARDC, SFR-DC, or mHTGR-DC to develop each PDC. For example, FHRs are liquid-metal reactors that use tristructural isotropic (TRISO) fuel, which is the same fuel used for mHTGR technologies. An FHR designer could use the mHTGR-DC where appropriate for the design. Another example is the MSRs that use molten fuel. An MSR designer may need to develop new PDC for molten fuel and systems to support this design.

In each case, it is the designers/applicants responsibility to provide not only the PDC for the design but also supporting information that justifies to the NRC how the design meets the PDC submitted.

In instances where a GDC or non-LWR design criterion (ARDC, SFR-DFC, mHTGR-DC) is not proposed, the designer/applicant must provide a basis and justify the omission from a safety perspective.

As noted earlier in this RG under the subheading, Role of the General Design Criteria for Non-LWRs, the current GDC are regulations and therefore use the words shall and must that are appropriate for regulatory requirements. The proposed ARDC, SFR-DC, and mHTGR-DC presented in Appendices A, B, and C to this RG also use the words shall, and must for consistency with the GDC, and so that non-LWR applicants can use them in the same manner as GDC when developing PDC.

However, this wording does not make them regulatory requirements, as they are contained in a guidance document.

Finally, the non-LWR design criteria as developed by the NRC staff are intended to provide stakeholders with insights into the staffs 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 may eventually be required from a non-LWR applicant.

Non-LWR Crosswalk Table The following table (Table 1) provides a summary and crosswalk between the LWR GDC contained in Appendix A and the NRC staffs determination of their applicability to the ARDC, SFR-DC, and mHTGR-DC. For each design criterion, the table denotes the status (same as GDC, same as ARDC, modified for ARDC, modified for SFR-DC, or modified for mHTGR-DC). Table 1 also uses redline-strikeout to identify the design criteria titles that have been modified for non-LWRs. Words removed from the title are in red with a strikethrough and words that have been added are in blue and underlined.

The actual ARDC, SFR-DC, and mHTGR-DC and NRC staff technology-specific rationale for adaptions to the GDCs are contained in Appendices A-C to this RG.

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The table consists of five columns:

Column 1Criterion Number Column 2Current GDC Title (from 10 CFR Part 50, Appendix A)

Column 3ARDC Title/Status (showing conformity/deviation from 10 CFR Part 50, Appendix A)

Column 4SFR-DC Title/Status (showing conformity/deviation from 10 CFR Part 50, Appendix A)

Column 5mHTGR-DC Title/Status (showing conformity/deviation from 10 CFR Part 50, Appendix A)

The table is divided into seven sections similar to those in 10 CFR Part 50, Appendix A:

Section IOverall Requirements (Criteria 1-5)

Section IIMultiple Barriers (Criteria 10-19)

Section IIIReactivity Control (Criteria 20-29)

Section IVFluid Systems (Criteria 30-46)

Section VReactor Containment (Criteria 50-57)

Section VIFuel and Radioactivity Control (Criteria 60-64)

Section VIIAdditional SFR-DC (Criteria 70-77) and Additional mHTGR-DC (Criteria 70-72)

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TABLE 1: NON-LIGHT-WATER-REACTOR CROSSWALK I. Overall Requirements Criterion Current GDC Title ARDC Title/Status SFR-DC Title/Status mHTGR-DC Title/Status 1 Quality standards and records. Same as GDC Same as GDC Same as GDC 2 Design bases for protection Same as GDC Same as GDC Same as GDC against natural phenomena.

3 Fire protection. Fire protection. Same as ARDC Same as ARDC Modified for ARDC 4 Environmental and dynamic Environmental and dynamic Environmental and dynamic Environmental and dynamic effects design bases. effects design bases. effects design bases. effects design bases.

Modified for ARDC Modified for SFR-DC Modified for mHTGR-DC 5 Sharing of structures, systems, Same as GDC Same as GDC Same as GDC and components.

II. Multiple Barriers Criterion Current GDC Title ARDC Title/Status SFR-DC Title/Status mHTGR-DC Title/Status 10 Reactor design. Same as GDC Same as GDC Reactor design.

Modified for mHTGR-DC 11 Reactor inherent protection. Reactor inherent protection. Same as ARDC Same as ARDC Modified for ARDC 12 Suppression of reactor Suppression of reactor power Same as ARDC Suppression of reactor power power oscillations. oscillations. oscillations.

Modified for ARDC Modified for mHTGR-DC 13 Instrumentation and control. Instrumentation and control. Instrumentation and control. Instrumentation and control.

Modified for ARDC Modified for SFR-DC Modified for mHTGR-DC DG-1330, Page 13

TABLE 1: NON-LIGHT-WATER-REACTOR CROSSWALK II. Multiple Barriers Criterion Current GDC Title ARDC Title/Status SFR-DC Title/Status mHTGR-DC Title/Status 14 Reactor coolant pressure Reactor coolant pressure Primary coolant pressure Reactor helium coolant boundary. boundary. boundary. pressure boundary.

Modified for ARDC Modified for SFR-DC Modified for mHTGR-DC 15 Reactor coolant system Reactor coolant system design. Primary Reactor coolant Reactor helium pressure design. Modified for ARDC system design. boundary coolant system Modified for SFR-DC design.

Modified for mHTGR-DC 16 Containment design. Same as GDC Containment design. Containment design.

Modified for SFR-DC Modified for mHTGR-DC 17 Electric power systems. Electric power systems. Electric power systems. Electric power systems.

Modified for ARDC Modified for SFR-DC Modified for mHTGR-DC 18 Inspection and testing of Inspection and testing of electric Same as ARDC Same as GDC electric power systems. power systems.

Modified for ARDC 19 Control room. Control room. Control room. Same as ARDC Modified for ARDC Modified for SFR-DC III. Reactivity Control Criterion Current GDC Title ARDC Title/Status SFR-DC Title/Status mHTGR-DC Title/Status 20 Protection system functions. Same as GDC Same as GDC Protection system functions.

Modified for mHTGR-DC 21 Protection system reliability Same as GDC Same as GDC Same as GDC and testability.

22 Protection system Same as GDC Same as GDC Same as GDC independence.

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TABLE 1: NON-LIGHT-WATER-REACTOR CROSSWALK III. Reactivity Control Criterion Current GDC Title ARDC Title/Status SFR-DC Title/Status mHTGR-DC Title/Status 23 Protection system failure Same as GDC Protection system failure Same as GDC modes. modes.

Modified for SFR-DC 24 Separation of protection and Same as GDC Same as GDC Same as GDC control systems.

25 Protection system Protection system requirements for Same as ARDC Protection system requirements for reactivity reactivity control malfunctions. requirements for reactivity control malfunctions. Modified for ARDC control malfunctions.

Modified for mHTGR-DC 26 Reactivity control system Reactivity control systems Same as ARDC Same as ARDC redundancy and capability. redundancy and capacity Modified for ARDC 27 Combined reactivity control Combined reactivity control Same as ARDC Same as ARDC systems capability systems capability DELETED and incorporated into ARDC 26 28 Reactivity limits. Reactivity limits. Reactivity limits. Reactivity limits.

Modified for ARDC Modified for SFR-DC Modified for mHTGR-DC 29 Protection against Same as GDC Same as GDC Same as GDC anticipated operational occurrences.

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TABLE 1: NON-LIGHT-WATER-REACTOR CROSSWALK IV. Fluid Systems Criterion Current GDC Title ARDC Title/Status SFR-DC Title/Status mHTGR-DC Title/Status 30 Quality of reactor coolant Quality of reactor coolant Quality of reactor primary Quality of reactor helium coolant pressure boundary. pressure boundary. coolant pressure boundary. pressure boundary.

Modified for ARDC Modified for SFR-DC Modified for mHTGR-DC 31 Fracture prevention of reactor Fracture prevention of reactor Fracture prevention of reactor Fracture prevention of reactor coolant pressure boundary. coolant pressure boundary. primary coolant pressure helium coolant pressure Modified for ARDC boundary. boundary.

Modified for SFR-DC Modified for mHTGR-DC 32 Inspection of reactor coolant Inspection of reactor coolant Inspection of reactor primary Inspection of reactor helium pressure boundary. pressure boundary. coolant pressure boundary. coolant pressure boundary.

Modified for ARDC Modified for SFR-DC Modified for mHTGR-DC 33 Reactor coolant makeup. Reactor coolant inventory Reactor Primary coolant Not applicable to mHTGR.

maintenance .makeup inventory maintenance Modified for ARDC makeup.

Modified for SFR-DC 34 Residual heat removal. Residual heat removal. Residual heat removal. Passive residual heat removal.

Modified for ARDC Modified for SFR-DC Modified for mHTGR-DC Emergency core cooling. Same as ARDC Not applicable to mHTGR.

35 Emergency core cooling.

Modified for ARDC Inspection of passive emergency Inspection of emergency core Inspection of emergency core core cooling residual heat 36 cooling system. Same as ARDC cooling system. removal system.

Modified for ARDC Modified for mHTGR-DC Testing of passive residual heat Testing of emergency core Testing of emergency core removal emergency core cooling 37 cooling system. Same as ARDC cooling system. system.

Modified for ARDC Modified for mHTGR-DC Containment heat removal.

38 Containment heat removal. Same as ARDC Not applicable to mHTGR.

Modified for ARDC DG-1330, Page 16

TABLE 1: NON-LIGHT-WATER-REACTOR CROSSWALK IV. Fluid Systems Criterion Current GDC Title ARDC Title/Status SFR-DC Title/Status mHTGR-DC Title/Status Inspection of containment heat Inspection of containment heat 39 removal system. Same as ARDC Not applicable to mHTGR.

removal system.

Modified for ARDC Testing of containment heat Testing of containment heat 40 removal system. Same as ARDC Not applicable to mHTGR.

removal system.

Modified for ARDC Containment atmosphere Containment atmosphere 41 cleanup. Same as ARDC Not applicable to mHTGR.

cleanup.

Modified for ARDC Inspection of containment Same as GDC Same as GDC 42 Not applicable to mHTGR.

atmosphere cleanup systems.

Testing of containment Testing of containment 43 atmosphere cleanup systems. Same as ARDC Not applicable to mHTGR.

atmosphere cleanup systems.

Modified for ARDC Structural and equipment Structural and equipment 44 Cooling water. cooling. Cooling water Same as ARDC cooling. Cooling water Modified for ARDC Modified for mHTGR-DC Inspection of structural and Inspection of cooling water equipment cooling water 45 Same as ARDC Same as ARDC system. systems.

Modified for ARDC Testing of structural and Testing of cooling water equipment cooling water 46 Same as ARDC Same as ARDC system. systems.

Modified for ARDC DG-1330, Page 17

TABLE 1: NON-LIGHT-WATER-REACTOR CROSSWALK V. Reactor Containment Criterion Current GDC Title ARDC Title/Status SFR-DC Title/Status mHTGR-DC Title/Status Containment design basis.

50 Containment design basis. Same as ARDC Not applicable to mHTGR.

Modified for ARDC Fracture prevention of Fracture prevention of 51 containment pressure boundary. Same as ARDC Not applicable to mHTGR.

containment pressure boundary.

Modified for ARDC Capability for containment Capability for containment 52 leakage rate testing. Same as ARDC Not applicable to mHTGR.

leakage rate testing.

Modified for ARDC Provisions for containment Provisions for containment 53 testing and inspection. Same as ARDC Not applicable to mHTGR.

testing and inspection.

Modified for ARDC Piping systems penetrating Piping systems penetrating Piping systems penetrating 54 containment. containment. Not applicable to mHTGR.

containment.

Modified for ARDC Modified for SFR-DC Reactor coolant pressure Reactor Primary coolant Reactor coolant pressure boundary penetrating pressure boundary penetrating 55 boundary penetrating Not applicable to mHTGR.

containment. containment.

containment.

Modified for ARDC Modified for SFR-DC Primary Containment isolation.

56 Primary containment isolation. Same as ARDC Not applicable to mHTGR.

Modified for ARDC Closed system isolation valves. Closed system isolation valves.

57 Closed system isolation valves. Not applicable to mHTGR.

Modified for ARDC Modified for SFR-DC DG-1330, Page 18

TABLE 1: NON-LIGHT-WATER-REACTOR CROSSWALK VI. Fuel and Radioactivity Control Criterion Current GDC Title ARDC Title/Status SFR-DC Title/Status mHTGR-DC Title/Status Control of releases of Same as GDC Same as GDC Same as GDC 60 radioactive materials to the environment.

Fuel storage and handling and Same as ARDC Same as ARDC Fuel storage and handling and 61 radioactivity control.

radioactivity control.

Modified for ARDC Prevention of criticality in fuel Same as GDC Same as GDC Same as GDC 62 storage and handling.

Monitoring fuel and waste Same as GDC Same as GDC Same as GDC 63 storage.

Monitoring radioactivity Monitoring radioactivity Monitoring radioactivity Monitoring radioactivity 64 releases. releases. releases.

releases.

Modified for ARDC Modified for SFR-DC Modified for mHTGR-DC DG-1330, Page 19

TABLE 1: NON-LIGHT-WATER-REACTOR CROSSWALK VII. Additional Technology-Specific Design Criteria Criterion Current GDC Title ARDC Title/Status SFR-DC Title/Status mHTGR-DC Title/Status Reactor vessel and reactor 70 N/A N/A Intermediate coolant system.

system structural design basis.

Primary coolant and cover gas 71 N/A N/A Reactor building design basis.

purity control.

Provisions for periodic reactor 72 N/A N/A Sodium heating systems.

building inspection.

Sodium leakage detection and 73 N/A N/A reaction prevention and N/A mitigation.

Sodium/water reaction 74 N/A N/A N/A prevention/mitigation.

Quality of the intermediate 75 N/A N/A N/A coolant boundary.

Fracture prevention of the 76 N/A N/A N/A intermediate coolant boundary.

Inspection of the intermediate 77 N/A N/A N/A coolant boundary.

Primary coolant system 78 N/A N/A N/A interfaces.

Cover gas inventory 79 N/A N/A N/A maintenance.

DG-1330, Page 20

D. IMPLEMENTATION The purpose of this section is to provide information on how applicants and licensees 2 may use this guide and information regarding the NRCs plans for using this RG. In addition, it describes how the NRC staff complies with 10 CFR 50.109, Backfitting, and any applicable finality provisions in 10 CFR Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants.

Use by Applicants and Licensees Applicants and licensees may voluntarily 3 use the guidance in this document to demonstrate compliance with the underlying NRC regulations. Methods or solutions that differ from those described in this RG may be deemed acceptable if the applicant or licensee provides sufficient basis and information for the NRC staff to verify that the proposed alternative demonstrates compliance with the appropriate NRC regulations.

Licensees may use the information in this RG for actions which do not require NRC review and approval such as changes to a facility design under 10 CFR 50.59, Changes, Tests, and Experiments.

Licensees may use the information in this RG or applicable parts to resolve regulatory or inspection issues.

Use by NRC Staff The NRC staff does not intend or approve any imposition or backfitting of the guidance in this RG. The NRC staff does not expect any existing licensee to use or commit to using the guidance in this RG, unless the licensee makes a change to its licensing basis. The NRC staff does not expect or plan to request licensees to voluntarily adopt this RG to resolve a generic regulatory issue. The NRC staff does not expect or plan to initiate NRC regulatory action which would require the use of this RG without further backfit consideration. Examples of such unplanned NRC regulatory actions include: issuance of an order requiring the use of the RG, requests for information under 10 CFR 50.54(f) as to whether a licensee intends to commit to use of this RG, or generic communication, or promulgation of a rule requiring the use of this RG.

During regulatory discussions on plant-specific operational issues, the staff may discuss with licensees various actions consistent with staff positions in this RG, as one acceptable means of meeting the underlying NRC regulatory requirement. Such discussions would not ordinarily be considered backfitting. And, unless this RG is part of the licensing basis for a facility, the staff may not represent to the licensee that the licensees failure to comply with the positions in this RG constitutes a violation.

If an existing licensee voluntarily seeks a license amendment or change and (1) the NRC staffs consideration of the request involves a regulatory issue directly relevant to this new RG and (2) the specific subject matter of this RG is an essential consideration in the staffs determination of the acceptability of the licensees request, then the staff may request that the licensee either follow the guidance in this RG or provide an equivalent alternative process that demonstrates compliance with the 2 In this section, licensees refers to licensees of nuclear power plants under 10 CFR Parts 50 and 52; the term applicants, refers to applicants for licenses and permits for (or relating to) nuclear power plants under 10 CFR Parts 50 and 52, and applicants for standard design approvals and standard design certifications under 10 CFR Part 52.

3 In this section, voluntary and voluntarily mean that the licensee is seeking the action of its own accord, without the force of a legally binding requirement or an NRC representation of further licensing or enforcement action.

DG-1330, Page 20

underlying NRC regulatory requirements. This is not considered backfitting as defined in 10 CFR 50.109(a)(1) or a violation of any of the issue finality provisions in 10 CFR Part 52.

Additionally, an existing applicant may be required to comply with new rules, orders, or guidance if 10 CFR 50.109(a)(3) applies.

If a licensee believes that the NRC is either using this RG or requesting or requiring the licensee to implement the methods or processes in this RG in a manner inconsistent with the discussion in this Implementation section, then the licensee may file a backfit appeal with the NRC in accordance with the guidance in NUREG-1409, Backfitting Guidelines (Ref. 24), and the NRC Management Directive 8.4, Management of Facility-Specific Backfitting and Information Collection (Ref. 25).

DG-1330, Page 21

ACRONYMS ANS American Nuclear Society ANSI American National Standards Institute AOO anticipated operational occurrence ARDC advanced reactor design criteria ASME American Society of Mechanical Engineers BDBE beyond-design-basis event CFR Code of Federal Regulations DOE U.S. Department of Energy DRACS Direct Reactor Auxiliary Cooling System EAB exclusion area boundary FAUNA Forschungsanlage zur Untersuchung nuklearer Aerosole FHR fluoride high-temperature reactors GCR gas-cooled fast reactors GDC general design criterion/criteria HTGR high-temperature gas-cooled reactor IAEA International Atomic Energy Agency IHTS intermediate heat transport system LFR lead fast reactor LMR liquid-metal reactor LPZ low-population zone LWR light-water reactor mHTGR modular high-temperature gas-cooled reactor mHTGR-DC mHTGR design criteria MSR molten salt reactors NaK sodium-potassium NGNP Next Generation Nuclear Plant NRC U.S. Nuclear Regulatory Commission PDC principal design criteria PRISM Power Reactor Innovative Small Module RCCS reactor cavity cooling system RG regulatory guide SARRDL specified acceptable system radionuclide release design limit SFR sodium-cooled fast reactors SFR-DC SFR design criteria SRM staff requirements memorandum/memoranda SSC structure, system, and component TRISO tristructural isotropic fuel DG-1330, Page 22

REFERENCES 4

1. U.S. Code of Federal Regulations (CFR), Domestic Licensing of Production and Utilization Facilities, Part 50, Chapter 1, Title 10, Energy. (10 CFR Part 50)

2. CFR, Licenses, Certifications, and Approvals for Nuclear Power Plants, Part 52, Title 10, Energy. (10 CFR Part 52)

3. U.S. Nuclear Regulatory Commission (NRC), NUREG-1338, Draft Preapplication Safety Evaluation Report for the Modular High-Temperature Gas-Cooled Reactor (MHTGR),

December 1995. (Agencywide Documents Access and Management System (ADAMS)

Accession No. ML052780497).

4. NRC, NUREG-1368, Preapplication Safety Evaluation Report for the Power Reactor Innovative Small Module (PRISM) Liquid-Metal Reactor, February 1994. (ADAMS Accession No. ML063410561).

5. NRC, NUREG-0968, Safety Evaluation Report Related to the Construction of the Clinch River Breeder Reactor Plant, March 1983. (ADAMS Accession No. ML082381008).

6. NRC, NUREG-1369, Preapplication Safety Evaluation Report for the Sodium Advanced Fast Reactor (SAFR) Liquid-Metal Reactor, December 1991. (ADAMS Accession No. ML063410547).

7. NRC, SECY-93-092, Issues Pertaining to the Advanced Reactor (PRISM, MHTGR, and PIUS) and CANDU 3 Designs and their Relationship to Current Regulatory Requirements, April 1993.

(ADAMS Accession No. ML040210725).

8. NRC, SRM-SECY-93-092, Issues Pertaining to the Advanced Reactor (PRISM, MHTGR, and PIUS) and CANDU 3 Designs and their Relationship to Current Regulatory Requirements, July 1993. (ADAMS Accession No. ML003760774).

9. NRC, SECY-03-0047, Policy Issues Related to Licensing Non-Light-Water Reactor Designs, March 2003. (ADAMS Accession No. ML030160002).

10. NRC, SRM-SECY-03-0047, Policy Issues Related to Licensing Non-Light-Water Reactor Designs, June 2003. (ADAMS Accession No. ML031770124).

11. NRC, Next Generation Nuclear Plant - Assessment of Key Licensing Issues, July 17, 2014.

(ADAMS Accession Nos. ML14174A734, ML14174A774 (Enclosure 1), and ML14174A845 (Enclosure 2)).

12. NRC, Policy Statement on the Regulation of Advanced Reactors (73 FR 60612),

October 14, 2008. (ADAMS Accession No. ML082750370).

4 Publicly available NRC published documents are available electronically through the NRC Library on the NRCs public Web site at http://www.nrc.gov/reading-rm/doc-collections/ and through the NRCs Agencywide Documents Access and Management System (ADAMS) at http://www.nrc.gov/reading-rm/adams.html. The documents can also be viewed online or printed for a fee in the NRCs Public Document Room (PDR) at 11555 Rockville Pike, Rockville, MD. For problems with ADAMS, contact the PDR staff at 301-415-4737 or (800) 397-4209; fax (301) 415-3548; or e-mail pdr.resource@nrc.gov.

DG-1330, Page 23

13. NRC, Policy Statement on Severe Reactor Accidents Regarding Future Designs and Existing Plants, August 1985. (ADAMS Accession No. ML003711521).

14. NRC, SECY-15-0002, Proposed Updates of Licensing Policies Rules, and Guidance for Future New Reactor Applications, January 2015. (ADAMS Accession Nos. ML13281A382, ML13277A647 (Enclosure 1), ML13277A652 (Enclosure 2)).

15. NRC, SRM-SECY-15-002, Staff Requirements - SECY-15-0002 - Proposed Updates of Licensing Policies, Rules, and Guidance for Future New Reactor Applications September 22, 2015, (ADAMS Accession No. ML15266A023).

16. NRC, NRC Vision and Strategy: Safely Achieving Effective and Efficient Non-Light Water Reactor Mission Readiness, May 2016. (ADAMS Accession No. ML16139A812).

17. U.S. Department of Energy (DOE) report, Guidance for Developing Principal Design Criteria for Advanced (Non-Light Water) Reactors, December 2014. (ADAMS Accession Nos. ML14353A246 (cover-p. 84), ML14353A248 (pp.85-144)) 5.

18. NRC, Summary of January 21, 2015, Meeting to Discuss the Department of Energy Report, Guidance for Developing Principal Design Criteria for Advanced (non-Light Water) Reactors, February 24, 2015. (ADAMS Accession No. ML15044A081).

19. NRC, NRC Staff Questions on the DOE Report, Guidance for Developing Principal Design Criteria for Advanced Non-Light Water Reactors, June 5, 2015, and Response to NRC Staff Questions on the U.S. Department of Energy Report, Guidance for Developing Principal Design Criteria for Advanced Non-Light Water Reactors, July 15, 2015. (ADAMS Accession Nos. ML15154B575 (NRC letter) and ML15204A579 (DOE response), respectively).

20. NRC, NRC Staff Questions on the DOE Report, Guidance for Developing Principal Design Criteria for Advanced Non-Light Water Reactors, August 17, 2015, and Response to NRC Staff Questions on the U.S. Department of Energy Report, Guidance for Developing Principal Design Criteria for Advanced Non-Light Water Reactors, September 15, 2015. (ADAMS Accession Nos. ML15223B331 (NRC letter) and ML15272A096 (DOE responses), respectively).

21. NRC, Public Comment Sought - Advanced Non-Light Water Reactor Design Criteria, April 2016. (ADAMS Accession No. ML16096A420).

22. NRC, Non-LWR Design Criteria Public Comments, June 2016 (ML17011A116)

23. NRC, Summary of October 11, 2016 Public Meeting Regarding Non-Light Water Reactor Design Criteria. (ADAMS Accession No. ML16314B333).

24. NRC, NUREG-1409, Backfitting Guidelines, July 1990. (ADAMS Accession No. ML032230247).

25. NRC, Management Directive 8.4, Management of Facility-Specific Backfitting and Information Collection, October 2013. (ADAMS Accession No. ML12059A460).

5 Copies of U.S. Department of Energy (DOE) documents may be obtained from DOE at 1000 Independence Avenue, SW, Washington DC, 20585 or electronically from their web site: www.doe.gov.

DG-1330, Page 24

26. NRC, Response to Gap Analysis Summary Report for Reactor System Issues, September 2016.

(ADAMS Accession No. ML16116A083).

27. NRC, Response to NuScale Gap Analysis Summary Report for Reactivity Control Systems, Addressing Gap 11, General Design Criteria 26, December 2016. (ADAMS Accession No. ML16292A589).

28. U.S. Atomic Energy Commission (AEC), Proposed General Design Criteria of 1965, AEC-R 2/49, November 5, 1967. (32 FR 10216).

29. NRC, SECY-94-084, Policy and Technical Issues Associated with the Regulatory Treatment of Non-Safety Systems in Passive Plant Designs, March 1994. (ADAMS Accession No. ML003708068).

30. Nuclear Energy Agency (NEA), Experimental Facilities for Sodium Fast Reactor Safety Studies, Task Group on Advanced Reactors Experimental Facilities (TAREF), NEA, 2011, pp. 22 and 54. Available on-line: https://www.oecd-nea.org/globalsearch/download.php?doc=77089 6.

31. International Atomic Energy Agency (IAEA), Division of Nuclear Power, Nuclear Power Technology Development Section and INPRO Group, Vienna (Austria); Generation IV International Forum, Issy-les-Moulineaux (France); vp; 2013; 1 p; 3. Joint GIF-IAEA Workshop on Safety Design Criteria for Sodium-Cooled Fast Reactors; Vienna (Austria);

February 26-27, 2013, Safety Design Criteria for GEN IV Sodium-Cooled Fast Reactor Systems, SDC-TF/2013/01 May 1, 2013, p. 57. Available on-line:

http://www.iaea.org/NuclearPower/Downloadable/Meetings/2013/2013-02-26-02-27-TM-SFR/Sagayama_Opening_SFR_WS_26Feb2013.pdf 7.

32. DOE, Tanju Sofu, Argonne National Laboratory, Sodium-cooled Fast reactor (SFR) Technology Overview, IAEA Education and Training Seminar on Fast Reactor Science and Technology, ITESM Campus, Santa Fe, Mexico City, June 29-July 3, 2015. Available on-line:

https://www.iaea.org/NuclearPower/Downloadable/Meetings/2015/2015-06-29-07-03-NPTDS-mexico/2-3-_IAEAseminarMexicoCity_TSofu_SFRTechnologyOverview.pdf 8.

33. S. Savaranan, et al., NAFCON-SF: A sodium spray fire code for evaluating thermal consequences in SFR containment, Annals of Nuclear Energy, Vol. 90, April 2016, pp. 389-409.

Available on-line: http://www.sciencedirect.com/science/article/pii/S0306454915005800 9.

6 Copies of Nuclear Energy Agency (NEA) documents may be obtained through their Web site: WWW.OECD-NEA.org/

or by writing the Nuclear Energy Agency 46, quai Alphonse Le Gallo 92100 Boulogne-Billancourt, France.

7 Copies of International Atomic Energy Agency (IAEA) documents may be obtained through their Web site:

WWW.IAEA.Org/ or by writing the International Atomic Energy Agency P.O. Box 100 Wagramer Strasse 5, A-1400 Vienna, Austria. Telephone (+431) 2600-0, Fax (+431) 2600-7, or E-Mail at Official.Mail@IAEA.Org.

8 Copies of International Atomic Energy Agency (IAEA) documents may be obtained through their Web site:

WWW.IAEA.Org/ or by writing the International Atomic Energy Agency P.O. Box 100 Wagramer Strasse 5, A-1400 Vienna, Austria. Telephone (+431) 2600-0, Fax (+431) 2600-7, or E-Mail at Official.Mail@IAEA.Org.

9 Copies of Annals of Nuclear Energy Articles may be obtained through the Science Direct Web site:

WWW.ScienceDirect.com.

DG-1330, Page 25

34. DOE, Idaho National Laboratories (INL), Mechanistic Source Terms White Paper, INL/EXT-10-17997, Rev.0, July 2010, (ADAMS Accession No. ML102040260).

35. DOE. INL, Modular HTGR Safety Basis and Approach, Idaho National Laboratory, INL/EXT-11-22708, Rev.0, August 2011, (ADAMS Accession No. ML11251A169).

DG-1330, Page 2

APPENDIX A ADVANCED REACTOR DESIGN CRITERIA The table below contains the advanced reactor design criteria (ARDC). These criteria are generally applicable to six different types of non-light-water reactor (LWR) technologies (i.e., sodium-cooled fast reactors (SFRs), lead-cooled fast reactors, gas-cooled fast reactors, modular high-temperature gas-cooled reactors (mHTGRs), fluoride high-temperature reactors, and molten salt reactors).

Applicants/designers may use the ARDC in this appendix to develop all or part of the principal design criteria (PDC) and may choose among the ARDC, SFR-DC (Appendix B), or mHTGR-DC (Appendix C) to develop each PDC. Applicants/designers may also develop entirely new PDC as needed to address unique design features in their respective designs.

To develop these ARDC, the staff of the U.S. Nuclear Regulatory Commission (NRC) reviewed each general design criterion (GDC) in Appendix A, General Design Criteria for Nuclear Power Plants, to Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Domestic Licensing of Production and Utilization Facilities, to determine its applicability to non-LWR designs. The NRC staff then determined what if any adaptation was appropriate for non-LWRs. The results are included in column 2 of the table below. The table also includes the NRC staffs rationale for the adaptations. In many cases, the rationale refers to changes made to the language of the GDC. To fully understand the context of the rationale, the user of this RG should refer to the appropriate GDC. Where the NRC staff determined that the current GDC were applicable to the ARDC, the table denotes Same as GDC.

The table consists of three columns:

Column 1Criterion Number Column 2ARDC Title and Content Column 3NRC Rationale for Adaptations to GDC The table is further divided into six sections similar to 10 CFR Part 50, Appendix A:

Section IOverall Requirements (Criterion 1 - 15)

Section II - Multiple Barriers (Criterion 10 - 20)

Section III - Reactivity Control (Criterion 21 - 29)

Section IV - Fluid Systems (Criterion 30 - 46)

Section V - Reactor Containment (Criterion 50 - 57)

Section VI - Fuel and Radioactivity Control (Criterion 60 - 64)

Appendix A to DG-1330, Page A-1

APPENDIX A. ADVANCED REACTOR DESIGN CRITERIA I. Overall Requirements Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC 1 Quality standards and records.

Same as GDC 2 Design bases for protection against natural phenomena.

Same as GDC 3 Fire protection. The phrase containing examples where noncombustible and heat-Structures, systems, and components important to safety shall resistant materials must be used has been broadened to apply to all be designed and located to minimize, consistent with other advanced reactor designs.

safety requirements, the probability and effect of fires and explosions. Noncombustible and fire-resistant materials shall be Instead of and, the phrase locations with structures, systems, and used wherever practical throughout the unit, particularly in components (SSCs) important to safety uses or, which is locations with structures, systems, or components important to logically correct in this case.

safety. Fire detection and fighting systems of appropriate capacity and capability shall be provided and designed to minimize the adverse effects of fires on structures, systems, and components important to safety. Firefighting systems shall be designed to ensure that their rupture or inadvertent operation does not significantly impair the safety capability of these structures, systems, and components.

4 Environmental and dynamic effects design bases. This change removes the light-water reactor (LWR) emphasis on Structures, systems, and components important to safety shall loss-of-coolant accidents (LOCAs) that may not apply to every be designed to accommodate the effects of and to be compatible design. For example, helium is not needed in a mHTGR to remove with the environmental conditions associated with normal heat from the core during postulated accidents and does not have the operation, maintenance, testing, and postulated accidents. These same importance as water does to LWR designs to ensure that fuel structures, systems, and components, shall be appropriately integrity is maintained. Therefore, a specific reference to LOCAs is protected against dynamic effects, including the effects of not applicable to all designs. LOCAs may still require analysis in missiles, pipe whipping, and discharging fluids, that may result conjunction with postulated accidents if relevant to the design.

from equipment failures and from events and conditions outside the nuclear power unit. However, dynamic effects associated Reference to pipe whip may not be applicable to designs that with postulated pipe ruptures in nuclear power units may be operate at low pressure.

excluded from the design basis when analyses reviewed and approved by the Commission demonstrate that the probability of Appendix A to DG-1330, Page A-2

APPENDIX A. ADVANCED REACTOR DESIGN CRITERIA I. Overall Requirements Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC fluid system piping rupture is extremely low under conditions consistent with the design basis for the piping.

5 Sharing of structures, systems, and components.

Same as GDC II. Multiple Barriers Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC 10 Reactor design.

Same as GDC 11 Reactor inherent protection. The wording has been changed to broaden the applicability from The reactor core and associated systems that contribute to coolant systems to additional factors (including structures or other reactivity feedback shall be designed so that, in the power fluids) that may contribute to reactivity feedback. These systems are operating range, the net effect of the prompt inherent nuclear to be designed to compensate for rapid reactivity increase.

feedback characteristics tends to compensate for a rapid increase in reactivity.

12 Suppression of reactor power oscillations. The word structures was added because items such as reflectors, The reactor core; associated structures; and associated coolant, which could be considered either outside or not part of the reactor control, and protection systems shall be designed to ensure that core, may affect susceptibility of the core to power oscillations.

power oscillations that can result in conditions exceeding specified acceptable fuel design limits are not possible or can be reliably and readily detected and suppressed.

13 Instrumentation and control. Reactor coolant pressure boundary has been relabeled as reactor Instrumentation shall be provided to monitor variables and coolant boundary to create a more broadly applicable non-LWR systems over their anticipated ranges for normal operation, for term that defines the boundary without giving any implication of anticipated operational occurrences, and for accident conditions, system operating pressure. As such, the term "reactor coolant as appropriate to ensure adequate safety, including those boundary" is applicable to non-LWRs that operate at either low or variables and systems that can affect the fission process, the high pressure.

integrity of the reactor core, the reactor coolant boundary, and Appendix A to DG-1330, Page A-3

APPENDIX A. ADVANCED REACTOR DESIGN CRITERIA II. Multiple Barriers Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC the containment and its associated systems. Appropriate controls shall be provided to maintain these variables and systems within prescribed operating ranges.

14 Reactor coolant boundary. Reactor coolant pressure boundary has been relabeled as reactor The reactor coolant boundary shall be designed, fabricated, coolant boundary to create a more broadly applicable non-LWR erected, and tested so as to have an extremely low probability of term that defines the boundary without giving any implication of abnormal leakage, of rapidly propagating failure, and of gross system operating pressure. As such, the term reactor coolant rupture. boundary is applicable to non-LWRs that operate at either low or high pressure.

15 Reactor coolant system design. Reactor coolant pressure boundary has been relabeled as reactor The reactor coolant system and associated auxiliary, control, coolant boundary to create a more broadly applicable non-LWR and protection systems shall be designed with sufficient margin term that defines the boundary without giving any implication of to ensure that the design conditions of the reactor coolant system operating pressure. As such, the term "reactor coolant boundary are not exceeded during any condition of normal boundary" is applicable to non-LWRs that operate at either low or operation, including anticipated operational occurrences. high pressure.

16 Containment design. For non-LWR technologies other than SFRs and mHTGRs, Same as GDC designers may use the current GDC to develop applicable principal design criteria. However, it is also recognized that characteristics of the coolants, fuels, and containments to be used in non-LWR designs could share common features with SFRs and mHTGRs.

Hence designers may propose using the SFR-DC-16 or mHTGR-DC 16 as appropriate. Use of the mHTGR-DC 16 will be subject to a policy decision by the Commission. See rationale for mHTGR-DC 16 for further information on the policy decision.

17 Electric power systems. A reliable power system is required for SSCs during postulated Electric power systems shall be provided to permit functioning accident conditions. Power systems shall be sufficient in capacity, of structures, systems, and components important to safety. The capability, and reliability to ensure vital safety functions are safety function for the systems shall be to provide sufficient maintained. The emphasis is placed on requiring reliability of power capacity, capability, and reliability to ensure that (1) specified sources rather than prescribing how such reliability can be attained.

acceptable fuel design limits and design conditions of the Reference to onsite vs. offsite electric power systems was deleted to reactor coolant boundary are not exceeded as a result of Appendix A to DG-1330, Page A-4

APPENDIX A. ADVANCED REACTOR DESIGN CRITERIA II. Multiple Barriers Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC anticipated operational occurrences and (2) vital functions that provide for those reactor designs that do not depend on offsite rely on electric power are maintained in the event of postulated power for the functioning of SSCs important to safety.

accidents.

Text related to supplies, including batteries, and the onsite The onsite electric power systems shall have sufficient distribution system, was deleted to allow increased flexibility in independence, redundancy, and testability to perform their the design of offsite power systems for advanced reactor designs.

safety functions, assuming a single failure. However, it is still expected that such onsite systems must remain capable of performing assigned safety functions during accidents as a condition of requisite reliability.

The existing single switchyard allowance remains available under ARDC 17. If a particular advanced design requires the use of GDC single switchyard allowance wording, the designer should look to GDC 17 for guidance when developing PDC.

If electrical power is not required to permit functioning of SSCs important to 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.

Reactor coolant pressure boundary has been relabeled as reactor coolant boundary to create a more broadly applicable non-LWR term that defines the boundary without giving any implication of system operating pressure. As such, the term reactor coolant boundary is applicable to non-LWRs that operate at either low or high pressure.

18 Inspection and testing of electric power systems. GDC 18 is a design-independent companion criterion to GDC 17.

Electric power systems important to safety shall be designed to permit appropriate periodic inspection and testing of important Wording pertaining to additional system examples has been deleted areas and features, such as wiring, insulation, connections, and to allow increased flexibility associated with various designs.

switchboards, to assess the continuity of the systems and the condition of their components. The systems shall be designed Appendix A to DG-1330, Page A-5

APPENDIX A. ADVANCED REACTOR DESIGN CRITERIA II. Multiple Barriers Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC with a capability to test periodically (1) the operability and The text related to the nuclear power unit, offsite power system, and functional performance of the components of the systems, such onsite power system was deleted to be consistent with ARDC 17.

as onsite power sources, relays, switches, and buses, and (2) the operability of the systems as a whole and, under conditions as close to design as practical, the full operation sequence that brings the systems into operation, including operation of applicable portions of the protection system, and the transfer of power among systems.

19 Control room. The criterion was updated to remove specific emphasis on LOCAs, A control room shall be provided from which actions can be which may be not appropriate.

taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident Reference to whole body, or its equivalent to any part of the body conditions. Adequate radiation protection shall be provided to has been updated to the current total effective dose equivalent permit access and occupancy of the control room under accident standard as defined in § 50.2.

conditions without personnel receiving radiation exposures in excess of 5 rem total effective dose equivalent as defined in A control room habitability requirement beyond that associated with

§ 50.2 for the duration of the accident. radiation protection has been added to address the concern that nonradionuclide accidents may also affect control room access and Adequate habitability measures shall be provided to permit occupancy.

access and occupancy of the control room during normal operations and under accident conditions. Equipment at The last paragraph of the GDC has been eliminated for the ARDC appropriate locations outside the control room shall be provided because it is not applicable to future applicants.

(1) with a design capability for prompt hot shutdown of the reactor, including necessary instrumentation and controls to maintain the unit in a safe condition during hot shutdown, and (2) with a potential capability for subsequent cold shutdown of the reactor through the use of suitable procedures.

Appendix A to DG-1330, Page A-6

APPENDIX A. ADVANCED REACTOR DESIGN CRITERIA III. Reactivity Control Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC 20 Protection system functions.

Same as GDC 21 Protection system reliability and testability.

Same as GDC 22 Protection system independence.

Same as GDC 23 Protection system failure modes.

Same as GDC 24 Separation of protection and control systems.

Same as GDC 25 Protection system requirements for reactivity control Text has been added to clarify that the protection system is designed malfunctions. to protect the specified acceptable fuel design limits for anticipated The protection system shall be designed to ensure that specified operational occurrences (AOOs) in combination with a single acceptable fuel design limits are not exceeded during any failure; the protection system does not have to protect the specified anticipated operational occurrence accounting for a single acceptable fuel design limits during a postulated accident in malfunction of the reactivity control systems. combination with a single failure. The example was deleted to make the ARDC technology neutral.

26 Reactivity control systems. Recent licensing activity associated with the application of GDC 26 Reactivity control systems shall include the following and GDC 27 to new reactor designs Response to Gap Analysis capabilities: Summary Report for Reactor System Issues, (Ref. 26) and Response to NuScale Gap Analysis Summary Report for (1) A means of shutting down the reactor shall be provided to Reactivity Control Systems, Addressing Gap 11, General Design ensure that, under conditions of normal operation, including Criteria 26, (Ref. 27), revealed that additional clarity could be anticipated operational occurrences, and with appropriate provided in the area of reactivity control requirements. ARDC 26 margin for malfunctions, design limits for fission product combines the scope of GDC 26 and GDC 27. The development of barriers are not exceeded. ARDC 26 is informed by the proposed General Design Criteria of 1965, AEC-R 2/49 and November 5, 1967 (32 FR 10216) (Ref. 28);

(2) A means of shutting down the reactor and maintaining a the current GDC 26 and 27; the definition of safety-related SSC in safe shutdown under design-basis event conditions, with 10 CFR 50.2; and SECY-94-084, Policy and Technical Issues appropriate margin for malfunctions, shall be provided. A Associated with the Regulatory Treatment of Non-Safety Systems Appendix A to DG-1330, Page A-7

APPENDIX A. ADVANCED REACTOR DESIGN CRITERIA III. Reactivity Control Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC second means of reactivity control shall be provided that is in Passive Plant Designs (Ref. 29); and the prior application of independent, diverse, and capable of achieving and reactivity control requirements.

maintaining safe shutdown under design-basis event conditions. Current GDC 26, first sentence, states that two reactivity control systems of different design principles shall be provided. In addition, (3) A system for holding the reactor subcritical under cold the NRC has not licensed a power reactor that did not provide two conditions shall be provided. independent means of shutting down the reactor.

(1) Current GDC 26, second sentence, states that one of the reactivity control systems shall use control rods and shall be capable of reliably controlling reactivity changes to ensure that, under conditions of normal operation, including AOOs, and with appropriate margin for malfunctions such as stuck rods, specified acceptable fuel design limits are not exceeded. The staff recognizes that specifying control rods may not be suitable for advanced reactors. Additionally, reliably controlling reactivity, as required by GDC 26, has been interpreted as ensuring the control rods are capable of rapidly (i.e., within a few seconds) shutting down the reactor (Ref. 27).

The staff changed control rods to means in recognition that advanced reactor designs may not rely on control rods to rapidly shut down the reactor (e.g., alternative system designs or inherent feedback mechanisms may be relied upon to perform this function).

Additionally, specified acceptable fuel design limits is replaced with design limits for fission product barriers to be consistent with the AOO acceptance criteria. ARDC 10 and ARDC 15 provide the appropriate design limits for the fuel and reactor coolant boundary, respectively. A non-LWR may not necessarily shut down rapidly (within seconds) but the shutdown should occur in a time frame such that the fission product barrier design limits are not exceeded. In regards to safety class, the capability to shut down the Appendix A to DG-1330, Page A-8

APPENDIX A. ADVANCED REACTOR DESIGN CRITERIA III. Reactivity Control Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC reactor is identified as a function performed by safety-related SSCs in the 10 CFR 50.2 definition of safety-related SSCs.

(2) Current GDC 27 states that the reactivity control systems shall be designed to have a combined capability of reliably controlling reactivity changes to assure that, under postulated accident conditions and with appropriate margin for stuck rods, the capability to cool the core is maintained. Reliably controlling reactivity, as required by GDC 27, requires that the reactor achieve and maintain safe, stable conditions, including subcriticality, using only safety related equipment with margin for stuck rods (Ref. 26).

The first sentence of ARDC 26 (2) refers to the safety-related means (systems and/or mechanisms) to achieve and maintain safe shutdown. Maintain safe shutdown indicates subcriticality in the long term or an equilibrium condition naturally achieved by the design.

The staff changed reactivity control systems to means in recognition that advanced reactor designs may rely on a system, inherent feedback mechanism, or some combination thereof to shut down the reactor and maintain a safe shutdown under design-basis event conditions. SECY-94-084, Policy and Technical Issues Associated with the Regulatory Treatment of Non-Safety Systems in Passive Plant Designs (Ref. 29), describes the characteristics of a safe shutdown condition as reactor subcriticality, decay heat removal, and radioactive materials containment. The staff replaced postulated accident conditions with design-basis event conditions, to emphasize that plants are required to maintain a safe shutdown following AOOs as well as postulated accidents.

The second sentence of ARDC 26(2) refers to a means of achieving and maintaining shutdown that is important to safety but not Appendix A to DG-1330, Page A-9

APPENDIX A. ADVANCED REACTOR DESIGN CRITERIA III. Reactivity Control Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC necessarily safety related. The second means of reactivity control serves as a backup to the safety-related 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). Independent indicates no shared systems or components with the safety-related means and diverse indicates a different design than the safety-related means. The purpose of an independent and diverse means of controlling reactivity is to preclude a potential common cause failure affecting both means of reactivity control, which would lead to the inability to shut down the reactor. The second means of reactivity control does not have to demonstrate that design limits for fission product barriers are met.

Additionally, the current GDC 26, third sentence, states that the second reactivity control system shall be capable of reliably controlling the rate of changes resulting from planned, normal power changes (including xenon burnout) to assure acceptable fuel design limits are not exceeded. Staff has identified this as an operational requirement that is not necessary to ensure reactor safety provided a design complies with ARDC 26(1). Therefore, this sentence is not retained in ARDC 26.

27 Combined reactivity control systems capability.

DELETEDInformation incorporated into ARDC 26 28 Reactivity limits. Reactor coolant pressure boundary has been relabeled as reactor The reactivity control systems shall be designed with coolant boundary to create a more broadly applicable non-LWR appropriate limits on the potential amount and rate of reactivity term that defines the boundary without giving any implication of increase to ensure that the effects of postulated reactivity system operating pressure. As such, the term reactor coolant accidents can neither (1) result in damage to the reactor coolant boundary is applicable to non-LWRs that operate at either low or boundary greater than limited local yielding nor (2) sufficiently high pressure.

Appendix A to DG-1330, Page A-10

APPENDIX A. ADVANCED REACTOR DESIGN CRITERIA III. Reactivity Control Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC disturb the core, its support structures, or other reactor vessel internals to impair significantly the capability to cool the core. The list of postulated reactivity accidents has been deleted to make the ARDC technology neutral.

29 Protection against anticipated operational occurrences.

Same as GDC IV. Fluid Systems Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC 30 Quality of reactor coolant boundary. Reactor coolant pressure boundary has been relabeled as reactor Components that are part of the reactor coolant boundary shall coolant boundary to create a more broadly applicable non-LWR be designed, fabricated, erected, and tested to the highest quality term that defines the boundary without giving any implication of standards practical. Means shall be provided for detecting and, system operating pressure. As such, the term "reactor coolant to the extent practical, identifying the location of the source of boundary" is applicable to non-LWRs that operate at either low or reactor coolant leakage. high pressure.

31 Fracture prevention of reactor coolant boundary. Reactor coolant pressure boundary has been relabeled as reactor The reactor coolant boundary shall be designed with sufficient coolant boundary to create a more broadly applicable non-LWR margin to ensure that when stressed under operating, term that defines the boundary without giving any implication of maintenance, testing, and postulated accident conditions, (1) the system operating pressure. As such, the term "reactor coolant boundary behaves in a nonbrittle manner and (2) the probability boundary" is applicable to non-LWRs that operate at either low or of rapidly propagating fracture is minimized. The design shall high pressure.

reflect service temperatures, service degradation of material properties, creep, fatigue, stress rupture, and other conditions of Specific examples are added to the ARDC to account for the high the boundary material under operating, maintenance, testing, design and operating temperatures and unique potential coolants.

and postulated accident conditions and the uncertainties in determining (1) material properties, (2) the effects of irradiation and coolant chemistry on material properties, (3) residual, steady-state, and transient stresses, and (4) size of flaws.

Appendix A to DG-1330, Page A-11

APPENDIX A. ADVANCED REACTOR DESIGN CRITERIA IV. Fluid Systems Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC 32 Inspection of reactor coolant boundary. Reactor coolant pressure boundary has been relabeled as reactor Components that are part of the reactor coolant boundary shall coolant boundary to create a more broadly applicable non-LWR be designed to permit (1) periodic inspection and functional term that defines the boundary without giving any implication of testing of important areas and features to assess their structural system operating pressure. As such, the term "reactor coolant and leaktight integrity, and (2) an appropriate material boundary" is applicable to non-LWRs that operate at either low or surveillance program for the reactor vessel. high pressure.

The staff modified the LWR GDC by replacing the term reactor pressure vessel with reactor vessel, which the staff believes is a more generically applicable term.

33 Reactor coolant inventory maintenance. ARDC 33 was relabeled as inventory maintenance to provide A system to maintain reactor coolant inventory for protection more flexibility for advanced reactor designs. The first sentence is against small breaks in the reactor coolant boundary shall be modified so that it ends with ...shall be provided as necessary provided as necessary to ensure that specified acceptable fuel and is combined with the second sentence as necessary to design limits are not exceeded as a result of reactor coolant ensure (without the opening phrase The system safety function inventory loss due to leakage from the reactor coolant boundary shall be) to recognize that the inventory control system may be and rupture of small piping or other small components that are unnecessary for some designs to maintain safety functions that part of the boundary. The system shall be designed to ensure ensure fuel design limits are not exceeded.

that the system safety function can be accomplished using the piping, pumps, and valves used to maintain reactor coolant Reactor coolant pressure boundary has been relabeled as reactor inventory during normal reactor operation. coolant boundary to create a more broadly applicable non-LWR term that defines the boundary without giving any implication of system operating pressure. As such, the term "reactor coolant boundary" is applicable to non-LWRs that operate at either low or high pressure.

The staff maintained the words system safety function of GDC 33 because reactor coolant inventory maintenance may be necessary in some designs to support residual heat removal, which is a safety function. If not required for maintaining residual heat removal capability, the qualifier as necessary in the first sentence would apply. For example, if all small breaks or leaks would result Appendix A to DG-1330, Page A-12

APPENDIX A. ADVANCED REACTOR DESIGN CRITERIA IV. Fluid Systems Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC in reactor coolant inventory levels such that the residual heat removal function would still be performed, and the fuel design limits met, no safety function would be associated with the inventory maintenance system.

The GDC reference to electric power was removed. Refer to ARDC 17 concerning those systems that require electric power.

34 Residual heat removal. In most advanced reactor designs, a single system (i.e the residual A system to remove residual heat shall be provided. For normal heat removal system) is provided to perform both the residual heat operations and anticipated operational occurrences, the system removal and emergency core cooling functions. In this case, the safety function shall be to transfer fission product decay heat single system would be designed to meet the requirements of and other residual heat from the reactor core at a rate such that ARDC 34 and ARDC 35. (for more discussion see NUREG-0968 specified acceptable fuel design limits and the design conditions (Ref. 5) and NUREG-1368 (Ref.4)) However, the staff of the reactor coolant boundary are not exceeded. acknowledges that this may not be the case for every advanced reactor design. Therefore, to allow current and future non-LWR Suitable redundancy in components and features and suitable designers the flexibility to provide a single system or multiple interconnections, leak detection, and isolation capabilities shall systems to perform residual heat removal and emergency core be provided to ensure that the system safety function can be cooling, the staff decided to keep the ARDC 34 and ARDC 35 accomplished, assuming a single failure. separate in lieu of combining them into a single criterion. The staffs approach to provide two separate criteria is consistent with the approach taken in the LWR GDCs.

Reactor coolant pressure boundary has been relabeled as reactor coolant boundary to create a more broadly applicable non-LWR term that defines the boundary without giving any implication of system operating pressure. As such, the term reactor coolant boundary is applicable to non-LWRs that operate at either low or high pressure.

The second paragraph addresses residual heat removal system redundancy.

Appendix A to DG-1330, Page A-13

APPENDIX A. ADVANCED REACTOR DESIGN CRITERIA IV. Fluid Systems Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC The GDC reference to electric power was removed. Refer to ARDC 17 concerning those systems that require electric power.

35 Emergency core cooling. In most advanced reactor designs, a single system (i.e the residual A system to provide sufficient emergency core cooling shall be heat removal system) is provided to perform both the residual heat provided. The system safety function shall be to transfer heat removal and emergency core cooling functions. In this case, the from the reactor core such that effective core cooling is single system would be designed to meet the requirements of maintained and fuel damage is limited. ARDC 34 and ARDC 35. (for more discussion see NUREG-0968 (Ref. 5) and NUREG-1368 (Ref. 4)) However, the staff Suitable redundancy in components and features and suitable acknowledges that this may not be the case for every advanced interconnections, leak detection, isolation, and containment reactor design. Therefore, to allow current and future non-LWR capabilities shall be provided to ensure that the system safety designers the flexibility to provide a single system or multiple function can be accomplished, assuming a single failure. systems to perform residual heat removal and emergency core cooling, the staff decided to keep the ARDC 34 and ARDC 35 separate in lieu of combining them into a single criterion. Effective core cooling may include maintaining the primary coolant boundary in a condition necessary for adequate postulated accident heat removal. The staffs approach to provide two separate criteria is consistent with the approach taken in the LWR GDCs.

This change removes the light-water reactor emphasis on loss of coolant accidents that may not apply to every design. Loss of coolant accidents may still require analysis in conjunction with postulated accidents if they are relevant to the design.

The GDC reference to electric power was removed. Refer to ARDC 17 concerning those systems that require electric power.

36 Inspection of emergency core cooling system. In most advanced reactor designs, a single system (i.e the residual A system that provides emergency core cooling shall be heat removal system) is provided to perform both the residual heat designed to permit appropriate periodic inspection of important removal and emergency core cooling functions. In this case, the components to ensure the integrity and capability of the system. single system would be designed to meet the requirements of ARDC 34 and ARDC 35. (for more discussion see NUREG-0968 Appendix A to DG-1330, Page A-14

APPENDIX A. ADVANCED REACTOR DESIGN CRITERIA IV. Fluid Systems Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC (Ref. 5) and NUREG-1368 (Ref. 4)) However, the staff acknowledges that this may not be the case for every advanced reactor design. Therefore, to allow current and future non-LWR designers the flexibility to provide a single system or multiple systems to perform residual heat removal and emergency core cooling, the staff decided to keep the ARDC 34 and ARDC 35 separate in lieu of combining them into a single criterion. The staffs approach to provide two separate criteria is consistent with the approach taken in the LWR GDCs.

The ARDC has slightly different wording than the GDC to clarify the scope of the criterion. Any system, or portions of a system, credited with an emergency core cooling function during postulated accidents (for example, a system that performs both the residual heat removal function and the emergency core cooling function) would need to meet ARDC 36.

The list of examples has been deleted because it applies to LWR designs, and each specific design will have different important components associated with residual heat removal. This revision allows for a technology-neutral ARDC.

Review of the proposed DOE SFR and mHTGR DC found that only the SFR provided specific examples of important components but were generic in nature and did not include any significant additional guidance.

37 Testing of residual heat removal system. In most advanced reactor designs, a single system (i.e the residual A system that provides emergency core cooling shall be heat removal system) is provided to perform both the residual heat designed to permit appropriate periodic functional testing to removal and emergency core cooling functions. In this case, the ensure (1) the structural and leaktight integrity of its single system would be designed to meet the requirements of components, (2) the operability and performance of the system ARDC 34 and ARDC 35. (for more discussion see NUREG-0968 components, and (3) the operability of the system as a whole (Ref. 5) and NUREG-1368 (Ref. 4)) However, the staff Appendix A to DG-1330, Page A-15

APPENDIX A. ADVANCED REACTOR DESIGN CRITERIA IV. Fluid Systems Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC and, under conditions as close to design as practical, the acknowledges that this may not be the case for every advanced performance of the full operational sequence that brings the reactor design. Therefore, to allow current and future non-LWR system into operation, including operation of any associated designers the flexibility to provide a single system or multiple systems and interfaces necessary to transfer decay heat to the systems to perform residual heat removal and emergency core ultimate heat sink. cooling, the staff decided to keep the ARDC 34 and ARDC 35 separate in lieu of combining them into a single criterion. The staffs approach to provide two separate criteria is consistent with the approach taken in the LWR GDCs.

The ARDC has slightly different wording than the GDC to clarify the scope of the criterion. Any system, or portions of a system, credited with an emergency core cooling function during postulated accidents (for example, a system that performs both the residual heat removal function and the emergency core cooling function) would need to meet ARDC 37.

Specific mention of pressure testing has been removed yet remains a potential requirement should it be necessary as a component of appropriate periodic functional testing... of cooling systems.

A non-leaktight system may be acceptable for some designs provided that (1) the system leakage does not impact safety functions under all conditions, and (2) defense in depth is not impacted by system leakage.

Active has been deleted in item (2) as appropriate operability and performance system component testing are required, regardless of an active or passive nature.

Reference to the operation of applicable portions of the protection system, cooling water system, and power transfers is considered part of the more general associated systems. Together with the Appendix A to DG-1330, Page A-16

APPENDIX A. ADVANCED REACTOR DESIGN CRITERIA IV. Fluid Systems Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC ultimate heat sink, they are part of the operability testing of the system as a whole.

The GDC reference to electric power was removed. Refer to ARDC 17 concerning those systems that require electric power.

38 Containment heat removal. as necessary is meant to condition an ARDC 38 application A system to remove heat from the reactor containment shall be to designs requiring heat removal for conventional containments provided as necessary to maintain the containment pressure and that are found to require heat removal measures.

temperature within acceptable limits following postulated accidents. The LOCA reference has been removed to provide for any postulated accident that might affect the containment structure.

Suitable redundancy in components and features and suitable interconnections, leak detection, isolation, and containment Containment structure safety system redundancy is addressed in capabilities shall be provided to ensure that the system safety the second paragraph.

function can be accomplished, assuming a single failure.

39 Inspection of containment heat removal system. Examples were deleted to make the ARDC technology neutral.

The containment heat removal system shall be designed to permit appropriate periodic inspection of important components to ensure the integrity and capability of the system.

40 Testing of containment heat removal system. Specific mention of pressure testing has been removed yet The containment heat removal system shall be designed to remains a potential requirement should it be necessary as a permit appropriate periodic functional testing to ensure (1) the component of appropriate periodic functional testing... of structural and leaktight integrity of its components, (2) the containment heat removal.

operability and performance of the system components, and (3) the operability of the system as a whole, and under A non-leaktight system may be acceptable for some designs conditions as close to the design as practical, the performance of provided that (1) the system leakage does not impact safety the full operational sequence that brings the system into functions under all conditions, and (2) defense in depth is not operation, including the operation of associated systems. impacted by system leakage.

Reference to the operation of applicable portions of the protection system, structural and equipment cooling , and power transfers is Appendix A to DG-1330, Page A-17

APPENDIX A. ADVANCED REACTOR DESIGN CRITERIA IV. Fluid Systems Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC considered part of the more general associated systems for operability testing of the system as a whole.

The GDC reference to electric power was removed. Refer to ARDC 17 concerning those systems that require electric power.

41 Containment atmosphere cleanup. Advanced reactors offer potential for reaction product generation Systems to control fission products and other substances that that is different from that associated with clad metal-water may be released into the reactor containment shall be provided interactions. Therefore, the terms hydrogen and oxygen are as necessary to reduce, consistent with the functioning of other removed while other substances is retained to allow for associated systems, the concentration and quality of fission exceptions.

products released to the environment following postulated accidents and to control the concentration of other substances in Considering that a passive containment cooling system may be the containment atmosphere following postulated accidents to used or that the containment may have an additional safety ensure that containment integrity and other safety functions are function other than radionuclide retention, additional wording for maintained. maintaining safety-functions is added.

Each system shall have suitable redundancy in components and The GDC reference to electric power was removed. Refer to features and suitable interconnections, leak detection, isolation, ARDC 17 concerning those systems that require electric power.

and containment capabilities to ensure that its safety function can be accomplished, assuming a single failure.

42 Inspection of containment atmosphere cleanup systems.

Same as GDC 43 Testing of containment atmosphere cleanup systems. Active has been deleted in item (2), as appropriate operability The containment atmosphere cleanup systems shall be designed and performance testing of system components is required to permit appropriate periodic functional testing to ensure regardless of an active or passive nature, as are cited examples of (1) the structural and leaktight integrity of its components, active system components.

(2) the operability and performance of the system components, and (3) the operability of the systems as a whole and, under Examples of active systems under item (2) have been deleted, both conditions as close to design as practical, the performance of the to conform to similar wording in ARDC 37 and 40 and ensure that full operational sequence that brings the systems into operation, passive as well as active system components are considered.

including the operation of associated systems.

Appendix A to DG-1330, Page A-18

APPENDIX A. ADVANCED REACTOR DESIGN CRITERIA IV. Fluid Systems Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC Specific mention of pressure testing has been removed yet remains a potential requirement should it be necessary as a component of appropriate periodic functional testing... of cooling systems. A non-leaktight system may be acceptable for some designs provided that (1) the system leakage does not impact safety functions under all conditions, and (2) defense in depth is not impacted by system leakage.

The GDC reference to electric power was removed. Refer to ARDC 17 concerning those systems that require electric power.

44 Structural and equipment cooling. This renamed ARDC accounts for advanced reactor design system A system to transfer heat from structures, systems, and differences to include cooling requirements for SSCs, if applicable; components important to safety to an ultimate heat sink shall be this ARDC does not address the residual heat removal system provided, as necessary, to transfer the combined heat load of required under ARDC 34, and ECCS system under ARDC 35 these structures, systems, and components under normal operating and accident conditions. The GDC reference to electric power was removed. Refer to ARDC 17 concerning those systems that require electric power.

Suitable redundancy in components and features and suitable interconnections, leak detection, and isolation capabilities shall be provided to ensure that the system safety function can be accomplished, assuming a single failure.

45 Inspection of structural and equipment cooling systems. This renamed ARDC accounts for advanced reactor system design The structural and equipment cooling systems shall be designed differences to include possible cooling requirements for SSCs to permit appropriate periodic inspection of important important to safety.

components, such as heat exchangers and piping, to ensure the integrity and capability of the systems.

46 Testing of structural and equipment cooling systems. This renamed ARDC accounts for advanced reactor system design The structural and equipment cooling systems shall be designed differences to include possible cooling requirements for SSCs to permit appropriate periodic functional testing to ensure important to safety. Specific mention of pressure testing has (1) the structural and leaktight integrity of their components, been removed yet remains a potential requirement should it be (2) the operability and performance of the system components, necessary as a component of appropriate periodic functional Appendix A to DG-1330, Page A-19

APPENDIX A. ADVANCED REACTOR DESIGN CRITERIA IV. Fluid Systems Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC and (3) the operability of the systems as a whole and, under testing... of cooling systems. A non-leaktight system may be conditions as close to design as practical, the performance of the acceptable for some designs provided that (1) the system leakage full operational sequences that bring the systems into operation does not impact safety functions under all conditions, and (2) for reactor shutdown and postulated accidents, including the defense in depth is not impacted by system leakage.

operation of associated systems.

Active has been deleted in item (2) because appropriate operability and performance tests of system components are required regardless of their active or passive nature. The LOCA reference has been removed to provide for any postulated accident that might affect subject SSCs.

The GDC reference to electric power was removed. Refer to ARDC 17 concerning those systems that require electric power.

V. Reactor Containment Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC 50 Containment design basis. ARDC 50 specifically addresses a containment structure in the The reactor containment structure, including access openings, opening sentence and ARDC 51-57 support the containment penetrations, and the containment heat removal system shall be structures design basis. Therefore, ARDC 51-57 are modified by designed so that the containment structure and its internal adding the word structure to highlight the containment structure-compartments can accommodate, without exceeding the design specific criteria.

leakage rate and with sufficient margin, the calculated pressure and temperature conditions resulting from postulated accidents. The phrase loss-of-coolant accident is LWR specific because this This margin shall reflect consideration of (1) the effects of is understood to be the limiting containment structure accident for potential energy sources that have not been included in the an LWR design. It is replaced by the phrase postulated accident to determination of the peak conditions, (2) the limited experience allow for consideration of the design-specific containment structure and experimental data available for defining accident limiting accident for non-LWR designs.

phenomena and containment responses, and (3) the conservatism of the calculational model and input parameters. The example at the end of subpart 1 of the ARDC is LWR specific and therefore deleted.

Appendix A to DG-1330, Page A-20

APPENDIX A. ADVANCED REACTOR DESIGN CRITERIA V. Reactor Containment Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC 51 Fracture prevention of containment pressure boundary. ARDC 51-57 support ARDC 50, which specifically applies to non-The boundary of the reactor containment structure shall be LWR designs that use a fixed containment structure. Therefore, the designed with sufficient margin to ensure that, under operating, word structure is added to each of these ARDC to clearly convey maintenance, testing, and postulated accident conditions, (1) its the understanding that this criterion applies to designs employing materials behave in a nonbrittle manner and (2) the probability containment structures.

of rapidly propagating fracture is minimized. The design shall reflect consideration of service temperatures and other The term ferritic was removed to avoid limiting the scope of the conditions of the containment boundary materials during criterion to ferritic materials. With this revision, the staff believes operation, maintenance, testing, and postulated accident that this criterion is more broadly applicable to all non-LWR conditions, and the uncertainties in determining (1) material designs.

properties, (2) residual, steady-state, and transient stresses, and (3) size of flaws. The word pressure was left in the title to reflect that, while a design might not have a high-pressure containment like a traditional LWR, the containment still serves a pressure-retaining function.

52 Capability for containment leakage rate testing. ARDC 51-57 support ARDC 50, which specifically applies to non-The reactor containment structure and other equipment that may LWR designs that use a fixed containment structure. Therefore, the be subjected to containment test conditions shall be designed so word structure is added to each of these ARDC to clearly convey that periodic integrated leakage rate testing can be conducted at the understanding that this criterion applies to designs employing containment design pressure. containment structures.

53 Provisions for containment testing and inspection. ARDC 51-57 support ARDC 50, which specifically applies to non-The reactor containment structure shall be designed to permit LWR designs that use a fixed containment structure. Therefore, the (1) appropriate periodic inspection of all important areas, such word structure is added to each of these ARDC to clearly convey as penetrations, (2) an appropriate surveillance program, and the understanding that this criterion only applies to designs (3) periodic testing at containment design pressure of the leak- employing containment structures.

tightness of penetrations that have resilient seals and expansion bellows.

54 Piping systems penetrating containment. ARDC 51-57 support ARDC 50, which specifically applies to non-Piping systems penetrating the containment structure shall be LWR designs that use a fixed containment structure. Therefore, the provided with leak detection, isolation, and containment word structure is added to each of these ARDC to clearly convey capabilities having redundancy, reliability, and performance the understanding that this ARDC only applies to designs capabilities that reflect the importance to safety of isolating employing containment structures. The word reactor was Appendix A to DG-1330, Page A-21

APPENDIX A. ADVANCED REACTOR DESIGN CRITERIA V. Reactor Containment Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC these piping systems. Such piping systems shall be designed removed because the containment is a barrier between the fission with the capability to verify, by testing, the operational products and the environment. There are diverse advanced reactor readiness of any isolation valves and associated apparatus designs and, hence, there is no single containment concept. In all periodically and to confirm that valve leakage is within cases, the rules for containment penetrations to fulfill containment acceptable limits. isolation would apply. How this is accomplished should be left to the designer of the particular advanced reactor design, without being too prescriptive as to whether it is a primary or secondary or reactor containment. There may be a need for a containment structure outside the reactor region. For example, in the MSR design, some of the molten fuel salt is drawn off to a processing system to clean it up and remove fission products before returning it to the reactor. The molten fuel salt is highly radioactive and would need a containment around the entire system. Alternatively, in an SFR, the guard vessel would be the primary containment and, in the case of the PRISM design, a dome-shaped structure above it that would be the secondary containment. The secondary containment also has penetrations and needs containment isolation requirements to be fulfilled.

The adjustment to the last sentence enhances the clarity of the sentence with respect to the latest terminology used for periodic valve verification and operational readiness.

The American Society of Mechanical Engineers (ASME) Operation and Maintenance of Nuclear Power Plants, Division 1: OM Code:

Section IST (ASME OM Code) defines operational readiness as the ability of a component to perform its specified functions. The ASME OM Code is incorporated by reference in the NRC regulations in 10 CFR 50.55a, including the definition of operational readiness for pumps, valves, and dynamic restraints.

55 Reactor coolant boundary penetrating containment. ARDC 51-57 support ARDC 50, which specifically applies to non-Each line that is part of the reactor coolant boundary and that LWR designs that use a fixed containment structure. Therefore, the penetrates the containment structure shall be provided with word structure is added to each of these ARDC to clearly convey Appendix A to DG-1330, Page A-22

APPENDIX A. ADVANCED REACTOR DESIGN CRITERIA V. Reactor Containment Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC containment isolation valves, as follows, unless it can be the understanding that this ARDC only applies to designs demonstrated that the containment isolation provisions for a employing containment structures. The word reactor was specific class of lines, such as instrument lines, are acceptable removed because the containment is a barrier between the fission on some other defined basis: products and the environment. There are diverse advanced reactor designs and, hence, there is no single containment concept. In all (1) One locked closed isolation valve inside and one locked cases, the rules for containment penetrations to fulfill containment closed isolation valve outside containment; or isolation would apply. How this is accomplished should be left to (2) One automatic isolation valve inside and one locked closed the designer of the particular advanced reactor design, without isolation valve outside containment; or being too prescriptive as to whether it is a primary or secondary or (3) One locked closed isolation valve inside and one automatic reactor containment. There may be a need for a containment isolation valve outside containment. A simple check valve may structure outside the reactor region. For example, in the MSR not be used as the automatic isolation valve outside design, some of the molten fuel salt is drawn off to a processing containment; or system to clean it up and remove fission products before returning it (4) One automatic isolation valve inside and one automatic to the reactor. The molten fuel salt is highly radioactive and would isolation valve outside containment. A simple check valve may need a containment around the entire system. Alternatively, in an not be used as the automatic isolation valve outside SFR, the guard vessel would be the primary containment and, in the containment. case of the PRISM design, a dome-shaped structure above it that would be the secondary containment. The secondary containment Isolation valves outside containment shall be located as close to also has penetrations and needs containment isolation requirements containment as practical and upon loss of actuating power, to be fulfilled.

automatic isolation valves shall be designed to take the position that provides greater safety. Reactor coolant pressure boundary has been relabeled as reactor coolant boundary to create a more broadly applicable non-LWR Other appropriate requirements to minimize the probability or term that defines the boundary without giving any implication of consequences of an accidental rupture of these lines or of lines system operating pressure. As such, the term reactor coolant connected to them shall be provided as necessary to ensure boundary is applicable to non-LWRs that operate at either low or adequate safety. Determination of the appropriateness of these high pressure.

requirements, such as higher quality in design, fabrication, and testing; additional provisions for inservice inspection; protection against more severe natural phenomena; and additional isolation valves and containment, shall include consideration of the population density, use characteristics, and physical characteristics of the site environs.

Appendix A to DG-1330, Page A-23

APPENDIX A. ADVANCED REACTOR DESIGN CRITERIA V. Reactor Containment Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC 56 Containment isolation. ARDC 51-57 support ARDC 50, which specifically applies to non-Each line that connects directly to the containment atmosphere LWR designs that use a fixed containment structure. Therefore, the and penetrates the containment structure shall be provided with word structure is added to each of these ARDC to clearly convey containment isolation valves as follows, unless it can be the understanding that this criterion only applies to designs demonstrated that the containment isolation provisions for a employing containment structures. The word primary in the title specific class of lines, such as instrument lines, are acceptable and the text was removed, and the word reactor was also removed on some other defined basis: because the containment is a barrier between the fission products (1) One locked closed isolation valve inside and one locked and the environment. There are diverse advanced reactor designs closed isolation valve outside containment; or and, hence, there is no single containment concept. In all cases, the (2) One automatic isolation valve inside and one locked closed rules for containment penetrations to fulfill containment isolation isolation valve outside containment; or would apply. How this is accomplished should be left to the (3) One locked closed isolation valve inside and one automatic designer of the particular advanced reactor design, without being isolation valve outside containment. A simple check valve may too prescriptive as to whether it is a primary or secondary or reactor not be used as the automatic isolation valve outside containment. There may be a need for a containment structure containment; or outside the reactor region. For example, in the MSR design, some of (4) One automatic isolation valve inside and one automatic the molten fuel salt is drawn off to a processing system to clean it isolation valve outside containment. A simple check valve may up and remove fission products before returning it to the reactor.

not be used as the automatic isolation valve outside The molten fuel salt is highly radioactive and would need a containment. containment around the entire system. Alternatively, in an SFR, the guard vessel would be the primary containment and, in the case of Isolation valves outside containment shall be located as close to the PRISM design, a dome-shaped structure above it that would be the containment as practical and upon loss of actuating power, the secondary containment. The secondary containment also has automatic isolation valves shall be designed to take the position penetrations and needs containment isolation requirements to be that provides greater safety. fulfilled.

57 Closed system isolation valves. ARDC 51-57 support ARDC 50, which specifically applies to non-Each line that penetrates the containment structure and is neither LWR designs that use a fixed containment structure. Therefore, the part of the reactor coolant boundary nor connected directly to word structure is added to each of these ARDC to clearly convey the containment atmosphere shall have at least one containment the understanding that this criterion only applies to designs isolation valve, unless it can be demonstrated that the employing containment structures. The word reactor was containment safety function can be met without an isolation removed because the containment is a barrier between the fission valve and assuming failure of a single active component. The products and the environment. There are diverse advanced reactor isolation valve, if required, shall be either automatic, or locked designs and, hence, there is no single containment concept. In all Appendix A to DG-1330, Page A-24

APPENDIX A. ADVANCED REACTOR DESIGN CRITERIA V. Reactor Containment Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC closed, or capable of remote manual operation. This valve shall cases, the rules for containment penetrations to fulfill containment be outside containment and located as close to the containment isolation would apply. How this is accomplished should be left to as practical. A simple check valve may not be used as the the designer of the particular advanced reactor design, without automatic isolation valve. being too prescriptive as to whether it is a primary or secondary or reactor containment. There may be a need for a containment structure outside the reactor region. For example, in the MSR design, some of the molten fuel salt is drawn off to a processing system to clean it up and remove fission products before returning it to the reactor. The molten fuel salt is highly radioactive and would need a containment around the entire system. Alternatively, in an SFR, the guard vessel would be the primary containment and, in the case of the PRISM design, a dome-shaped structure above it that would be the secondary containment. The secondary containment also has penetrations and needs containment isolation requirements to be fulfilled.

Reactor coolant pressure boundary is relabeled as reactor coolant boundary to create a more broadly applicable non-LWR term that defines the boundary without giving any implication of system operating pressure. As such, the term reactor coolant boundary is applicable to non-LWRs that operate at either low or high pressure.

Appendix A to DG-1330, Page A-25

APPENDIX A. ADVANCED REACTOR DESIGN CRITERIA VI. Fuel and Radioactivity Control Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC 60 Control of releases of radioactive materials to the environment.

Same as GDC 61 Fuel storage and handling and radioactivity control. The underlying concept of establishing functional requirements for The fuel storage and handling, radioactive waste, and other radioactivity control in fuel storage and fuel handling systems is systems that may contain radioactivity shall be designed to independent of the design of non-LWR advanced reactors.

ensure adequate safety under normal and postulated accident However, some advanced designs may use dry fuel storage that conditions. These systems shall be designed (1) with a incorporates cooling jackets that can be liquid cooled or air cooled capability to permit appropriate periodic inspection and testing to remove heat. This modification to this GDC allows for both of components important to safety, (2) with suitable shielding liquid and air cooling of the dry fuel storage containers.

for radiation protection, (3) with appropriate containment, confinement, and filtering systems, (4) with a residual heat removal capability having reliability and testability that reflects the importance to safety of decay heat and other residual heat removal, and (5) to prevent significant reduction in fuel storage cooling under accident conditions.

62 Prevention of criticality in fuel storage and handling.

Same as GDC 63 Monitoring fuel and waste storage.

Same as GDC 64 Monitoring radioactivity releases. The phrase spaces containing components for recirculation of loss-Means shall be provided for monitoring the reactor containment of-coolant accident fluids was removed to allow for plant designs atmosphere, effluent discharge paths, and plant environs for that do not have LOCA fluids but may have other similar equipment radioactivity that may be released from normal operations, in spaces where radioactivity should be monitored.

including anticipated operational occurrences, and from postulated accidents.

Appendix A to DG-1330, Page A-26

APPENDIX B SODIUM-COOLED FAST REACTOR DESIGN CRITERIA The table below contains the sodium-cooled reactor design criteria (SFR-DC). These criteria are applicable to SFRs of both pool- and loop-type designs. Applicants/designers may use the SFR-DC in this appendix to develop all or part of the principal design criteria (PDC) and may choose among the advanced reactor design criteria (ARDC) (Appendix A), SFR-DC (Appendix B), or modular high-temperature gas-cooled reactor design criteria (mHTGR)-DC (Appendix C) to develop each PDC.

Applicants/designers may also develop entirely new PDC as needed to address unique design features in their respective designs.

To develop the SFR-DC, the staff of the U.S. Nuclear Regulatory Commission (NRC) reviewed each general design criterion (GDC) in Appendix A, General Design Criteria for Nuclear Power Plants, to Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Domestic Licensing of Production and Utilization Facilities, to determine its applicability to SFR designs. The NRC staff then determined what if any adaptation was appropriate for SFRs. The results are included in column 2 of the table below. The table also includes the NRC staffs rationale for the adaptations. In many cases, the rationale refers to changes made to the language of the GDC. To fully understand the context of the rationale, the user of this RG should refer to the appropriate GDC. Where the NRC staff determined that the current GDC or the ARDC were applicable to the SFR-DC, the table denotes Same as GDC or Same as ARDC, respectively.

The table consists of three columns:

Column 1Criterion Number Column 2SFR-DC Title and Content Column 3Staff Rationale for Adaptations to GDC The table is further divided into seven sections similar to those in 10 CFR Part 50, Appendix A:

Section IOverall Requirements (Criteria 1-5)

Section IIMultiple Barriers (Criteria 10-19)

Section IIIReactivity Control (Criteria 20-29)

Section IVFluid Systems (Criteria 30-46)

Section VReactor Containment (Criteria 50-57)

Section VIFuel and Radioactivity Control (Criteria 60-64)

Section VIIAdditional SFR-DC (Criteria 70-77)

Appendix B to DG-1330, Page B-1

APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA I. Overall Requirements Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC 1 Quality standards and records.

Same as GDC 2 Design bases for protection against natural phenomena.

Same as GDC 3 Fire protection. The phrase containing examples where noncombustible and fire-Same as ARDC resistant materials must be used has been broadened to apply to all Structures, systems, and components important to safety shall advanced reactor designs.

be designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and Instead of and, the phrase locations with structures, systems, and explosions. Noncombustible and fire-resistant materials shall be components (SSCs) important to safety uses or, which is used wherever practical throughout the unit, particularly in logically correct in this case.

locations with structures, systems, or components important to safety. Fire detection and fighting systems of appropriate capacity and capability shall be provided and designed to minimize the adverse effects of fires on structures, systems, and components important to safety. Firefighting systems shall be designed to ensure that their rupture or inadvertent operation does not significantly impair the safety capability of these structures, systems, and components.

4 Environmental and dynamic effects design bases. This change removes the light-water reactor (LWR) emphasis on Structures, systems, and components important to safety shall loss-of-coolant accidents (LOCAs) that may not apply to every be designed to accommodate the effects of, and to be design. For example, helium is not needed in a mHTGR to remove compatible with, the environmental conditions associated with heat from the core during postulated accidents and does not have the normal operation, maintenance, testing, anticipated operational same importance as water does for LWR designs to ensure that fuel occurrences, and postulated accidents, including the effects of integrity is maintained. Therefore, a specific reference to LOCAs is liquid sodium and its aerosols and oxidation products. These not applicable to all designs. LOCAs may still require analysis in structures, systems, and components shall be appropriately conjunction with postulated accidents if relevant to the design.

protected against dynamic effects, including the effects of missiles, pipe whipping, and discharging fluids that may result The phrase the environmental conditions associated with from equipment failures and from events and conditions outside anticipated operational occurrences has been added to ensure that the nuclear power unit. However, dynamic effects associated Appendix B to DG-1330, Page B-2

APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA I. Overall Requirements Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC with postulated pipe ruptures in nuclear power units may be the criterion would apply to all SFR design-basis events, as excluded from the design basis when analyses reviewed and suggested in NUREG-1368.

approved by the Commission demonstrate that the probability of fluid system piping rupture is extremely low under conditions A new sentence is added to ensure the designer considers the effects consistent with the design basis for the piping. of sodium leakage and associated chemical reactions with SSCs important to safety, which must be protected.

Chemical consequences of accidents, such as sodium leakage, shall be appropriately considered for the design of structures, systems, and components important to safety, which must be protected.

5 Sharing of structures, systems, and components.

Same as GDC II. Multiple Barriers Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC 10 Reactor design.

Same as GDC 11 Reactor inherent protection. The wording has been changed to broaden the applicability from Same as ARDC coolant systems to additional factors (including structures or other The reactor core and associated systems that contribute to fluids) that may contribute to reactivity feedback. These systems are reactivity feedback shall be designed so that, in the power to be designed to compensate for rapid reactivity increase.

operating range, the net effect of the prompt inherent nuclear feedback characteristics tends to compensate for a rapid increase in reactivity.

12 Suppression of reactor power oscillations. The word structures was added because items such as reflectors, Same as ARDC which could be considered either outside or not part of the reactor The reactor core; associated structures; and associated coolant, core, may affect susceptibility of the core to power oscillations.

control, and protection systems shall be designed to ensure that Appendix B to DG-1330, Page B-3

APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA II. Multiple Barriers Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC power oscillations that can result in conditions exceeding specified acceptable fuel design limits are not possible or can be reliably and readily detected and suppressed.

13 Instrumentation and control. Reactor coolant pressure boundary has been relabeled as primary Instrumentation shall be provided to monitor variables and coolant boundary to conform to standard terms used in the liquid-systems over their anticipated ranges for normal operation, for metal reactor (LMR) industry.

anticipated operational occurrences, and for accident conditions, as appropriate to ensure adequate safety, including those The use of the term primary indicates that the SFR-DC are variables and systems that can affect the fission process, the applicable to the primary cooling system, not the intermediate integrity of the reactor core, the primary coolant boundary, and cooling system.

the containment and its associated systems. Appropriate controls shall be provided to maintain these variables and systems within prescribed operating ranges.

14 Primary coolant boundary. Reactor coolant pressure boundary (RCPB) has been relabeled as The primary coolant boundary shall be designed, fabricated, primary coolant boundary to conform to standard terms used in erected, and tested so as to have an extremely low probability of the LMR industry.

abnormal leakage, of rapidly propagating failure, and of gross rupture. The use of the term primary indicates that the SFR-DC are applicable only to the primary cooling system, not the intermediate cooling system.

The cover gas boundary is included as part of the primary coolant boundary (referred to as RCPB by PRISM) per NUREG-1368 (page 3-38).

15 Primary coolant system design. Reactor coolant pressure boundary has been relabeled as primary The primary coolant system and associated auxiliary, control, coolant boundary to conform to standard terms used in the LMR and protection systems shall be designed with sufficient margin industry.

to ensure that the design conditions of the primary coolant boundary are not exceeded during any condition of normal The use of the term primary indicates that the SFR-DC are operation, including anticipated operational occurrences. applicable only to the primary cooling system, not the intermediate cooling system.

Appendix B to DG-1330, Page B-4

APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA II. Multiple Barriers Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC The cover gas boundary is included as part of the primary coolant boundary (referred to as RCPB by PRISM) per NUREG-1368 (page 3-38).

16 Containment design. The Commission approved the staffs recommendation to restrict A reactor containment consisting of a high-strength, low- the leakage of the containment to be less than that needed to meet leakage, pressure-retaining structure surrounding the reactor and the acceptable onsite and offsite dose consequence limits in its primary cooling system shall be provided to control the SECY-93-092 (Ref. 7). Therefore, the Commission agreed that the release of radioactivity to the environment and to ensure that the containment leakage for advanced reactors, similar to and including reactor containment design conditions important to safety are PRISM, NUREG-1368 (Ref. 4) should not be required to meet the not exceeded for as long as postulated accident conditions essentially leaktight statement in GDC 16.

require.

Furthermore, all past, current, and planned SFR designs use a high-The containment leakage shall be restricted to be less than that strength, low-leakage, pressure-retaining containment concept, needed to meet the acceptable onsite and offsite dose which aims to provide a barrier to contain the fission products and consequence limits, as specified in 10 CFR 50.34 for postulated other substances and to control the release of radioactivity to the accidents. environment.

Reactions of sodium with air or water, sodium fires, and hypothetical reactivity accidents caused by sodium voiding or boiling could release significant energy inside the reactor containment structure. Therefore, a high-strength, low-leakage, pressure-retaining structure surrounding the reactor and its primary cooling system is required. Note that a design could have a low design pressure for the containment.

Several technical reports and presentations support the need for a pressure-retaining structure surrounding SFRs.

The report, Experimental Facilities for Sodium Fast Reactor Safety Studies, Task Group on Advanced Reactors Experimental Facilities (TAREF)(Ref. 30), indicates that it is necessary for structures to Appendix B to DG-1330, Page B-5

APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA II. Multiple Barriers Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC withstand the thermo-mechanical load caused by sodium fire to avoid fire propagation and dispersion of aerosols.

The report, Safety Design Criteria for GEN IV Sodium-Cooled Fast Reactor Systems, (Ref. 31) notes that the design basis for containment shall consider pressure increase and thermal loads due to sodium fire.

During the presentation, SFR Technology Overview, IAEA Education and Training Seminar on Fast Reactor Science and Technology (Ref. 32), the technical expert noted that low design pressure for the containment basis is the heat produced by a potential sodium fire.

In the Annals of Nuclear Energy, the article, NAFCON-SF: A sodium spray fire code for evaluating thermal consequences in SFR containment, (Ref. 33) notes that Beschreibung der Forschungsanlage zur Untersuchung nuklearer Aerosole (FAUNA) spray fire experiments show peak pressures in containment over 3.5 bars within the first 5 seconds, gradually tapering downwards to less than 3.5 bars at 25 seconds.

17 Electric power systems. A reliable power system is required for SSCs during postulated Electric power systems shall be provided to permit functioning accident conditions. Power systems shall be sufficient in capacity, of structures, systems, and components important to safety. The capability, and reliability to ensure vital safety functions are safety function for the systems shall be to provide sufficient maintained. The emphasis is placed on requiring reliability of power capacity, capability, and reliability to ensure that (1) specified sources rather than prescribing how such reliability can be attained.

acceptable fuel design limits and design conditions of the Reference to onsite vs. offsite electric power systems was deleted to primary coolant boundary are not exceeded as a result of provide for those reactor designs that do not depend on offsite anticipated operational occurrences and (2) vital functions that power for the functioning of SSCs important to safety.

rely on electric power are maintained in the event of postulated The text related to supplies, including batteries, and the onsite accidents. distribution system, was deleted to allow increased flexibility in the design of onsite power systems for advanced reactor designs.

Appendix B to DG-1330, Page B-6

APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA II. Multiple Barriers Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC The onsite electric power systems shall have sufficient However, such onsite systems are expected to remain capable of independence, redundancy, and testability to perform their performing assigned safety functions during accidents as a condition safety functions, assuming a single failure. of requisite reliability. Reactor coolant pressure boundary has been relabeled as primary coolant boundary to conform to standard terms used in the LMR industry.

The existing single switchyard allowance remains available under ARDC 17. If a particular advanced design requires use of GDC single switchyard allowance wording, the designer should look to GDC 17 for guidance when developing PDC.

If electrical power is not required to permit the functioning of SSCs important to safety, the requirements in the SFR-DC 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.

18 Inspection and testing of electric power systems. GDC 18 is a design-independent companion criterion to GDC 17.

Same as ARDC.

Electric power systems important to safety shall be designed to Wording pertaining to additional system examples has been deleted permit appropriate periodic inspection and testing of important to allow increased flexibility associated with various designs.

areas and features, such as wiring, insulation, connections, and switchboards, to assess the continuity of the systems and the The text related to the nuclear power unit, offsite power system, and condition of their components. The systems shall be designed onsite power system was deleted to be consistent with ARDC 17.

with a capability to test periodically (1) the operability and functional performance of the components of the systems, such as onsite power sources, relays, switches, and buses, and (2) the operability of the systems as a whole and, under conditions as close to design as practical, the full operation sequence that brings the systems into operation, including operation of applicable portions of the protection system, and the transfer of power among systems.

Appendix B to DG-1330, Page B-7

APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA II. Multiple Barriers Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC 19 Control room. The criterion was updated to remove specific emphasis on LOCAs, A control room shall be provided from which actions can be which may be not appropriate for advanced designs such as the taken to operate the nuclear power unit safely under normal SFR.

conditions and to maintain it in a safe condition under accident conditions. Adequate radiation protection shall be provided to Reference to whole body, or its equivalent to any part of the body permit access and occupancy of the control room under accident has been updated to the current total effective dose equivalent conditions without personnel receiving radiation exposures in standard as defined in § 50.2.

excess of 5 rem total effective dose equivalent, as defined in

§ 50.2 for the duration of the accident. A control room habitability requirement beyond that associated with radiation protection has been added to address the concern that Adequate habitability measures shall be provided to permit nonradionuclide accidents, including accidental sodium leakage and access and occupancy of the control room during normal sodium fire, which could release sodium aerosols, may also affect operations and under accident conditions. control room access and occupancy.

Adequate protection against sodium aerosols shall be provided The last paragraph of the GDC has been eliminated for the SFR-DC to permit access and occupancy of the control room under because it is not applicable to future applicants.

accident conditions.

Equipment at appropriate locations outside the control room shall be provided (1) with a design capability for prompt hot shutdown of the reactor, including necessary instrumentation and controls to maintain the unit in a safe condition during hot shutdown, and (2) with a potential capability for subsequent cold shutdown of the reactor through the use of suitable procedures.

Appendix B to DG-1330, Page B-8

APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA III. Reactivity Control Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC 20 Protection system functions.

Same as GDC 21 Protection system reliability and testability.

Same as GDC 22 Protection system independence.

Same as GDC 23 Protection system failure modes. In NUREG-1368, Table 3.3 (page 3-21) (Ref. 4), the NRC staff The protection system shall be designed to fail into a safe state recommended adding the phrase sodium and sodium reaction or into a state demonstrated to be acceptable on some other products to the list of postulated adverse environments in the GDC.

defined basis, if conditions such as disconnection of the system, Therefore, sodium and sodium reaction products are added to the loss of energy (e.g., electric power, instrument air), or second list of examples in parentheses in SFR-DC 23.

postulated adverse environments (e.g., extreme heat or cold, fire, sodium and sodium reaction products, pressure, steam, water, and radiation) are experienced.

24 Separation of protection and control systems.

Same as GDC 25 Protection system requirements for reactivity control Text has been added to clarify that the protection system is designed malfunctions. to protect the specified acceptable fuel design limits for AOOs in Same as ARDC combination with a single failure; the protection system does not The protection system shall be designed to ensure that specified have to protect the specified acceptable fuel design limits during a acceptable fuel design limits are not exceeded during any postulated accident in combination with a single failure. The anticipated operational occurrence accounting for a single example was deleted to make the SFR technology neutral.

malfunction of the reactivity control systems.

26 Reactivity control systems. Recent licensing activity associated with the application of GDC 26 Same as ARDC and GDC 27 to new reactor designs Response to Gap Analysis Reactivity control systems shall include the following Summary Report for Reactor System Issues, (Ref. 26) and capabilities: Response to NuScale Gap Analysis Summary Report for Reactivity Control Systems, Addressing Gap 11, General Design (1) A means of shutting down the reactor shall be provided to Criteria 26, (Ref. 27), revealed that additional clarity could be ensure that, under conditions of normal operation, including provided in the area of reactivity control requirements. ARDC 26 Appendix B to DG-1330, Page B-9

APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA III. Reactivity Control Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC anticipated operational occurrences, and with appropriate combines the scope of GDC 26 and GDC 27. The development of margin for malfunctions, design limits for fission product ARDC 26 is informed by the proposed General Design Criteria of barriers are not exceeded. 1965, AEC-R 2/49 and November 5, 1967 (32 FR 10216) (Ref. 28);

the current GDC 26 and 27; the definition of safety-related SSC in (2) A means of shutting down the reactor and maintaining a 10 CFR 50.2; and SECY-94-084, Policy and Technical Issues safe shutdown under design-basis event conditions, with Associated with the Regulatory Treatment of Non-Safety Systems appropriate margin for malfunctions, shall be provided. A in Passive Plant Designs (Ref. 29); and the prior application of second means of reactivity control shall be provided that is reactivity control requirements.

independent, diverse, and capable of achieving and maintaining safe shutdown under design-basis event Current GDC 26, first sentence, states that two reactivity control conditions. systems of different design principles shall be provided. In addition, the NRC has not licensed a power reactor that did not provide two (3) A system for holding the reactor subcritical under cold independent means of shutting down the reactor.

conditions shall be provided.

(1) Current GDC 26, second sentence, states that one of the reactivity control systems shall use control rods and shall be capable of reliably controlling reactivity changes to ensure that, under conditions of normal operation, including AOOs, and with appropriate margin for malfunctions such as stuck rods, specified acceptable fuel design limits are not exceeded. The staff recognizes that specifying control rods may not be suitable for advanced reactors. Additionally, reliably controlling reactivity, as required by GDC 26, has been interpreted as ensuring the control rods are capable of rapidly (i.e., within a few seconds) shutting down the reactor (Ref. 27).

The staff changed control rods to means in recognition that advanced reactor designs may not rely on control rods to rapidly shut down the reactor (e.g., alternative system designs or inherent feedback mechanisms may be relied upon to perform this function).

Additionally, specified acceptable fuel design limits is replaced with design limits for fission product barriers to be consistent with the AOO acceptance criteria. ARDC 10 and ARDC 15 provide Appendix B to DG-1330, Page B-10

APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA III. Reactivity Control Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC the appropriate design limits for the fuel and reactor coolant boundary, respectively. A non-LWR may not necessarily shut down rapidly (within seconds) but the shutdown should occur in a time frame such that the fission product barrier design limits are not exceeded. In regards to safety class, the capability to shut down the reactor is identified as a function performed by safety-related SSCs in the 10 CFR 50.2 definition of safety-related SSCs.

(2) Current GDC 27 states that the reactivity control systems shall be designed to have a combined capability of reliably controlling reactivity changes to assure that, under postulated accident conditions and with appropriate margin for stuck rods, the capability to cool the core is maintained. Reliably controlling reactivity, as required by GDC 27, requires that the reactor achieve and maintain safe, stable conditions, including subcriticality, using only safety related equipment with margin for stuck rods (Ref. 26).

The first sentence of ARDC 26 (2) refers to the safety-related means (systems and/or mechanisms) to achieve and maintain safe shutdown. Maintain safe shutdown indicates subcriticality in the long term or an equilibrium condition naturally achieved by the design.

The staff changed reactivity control systems to means in recognition that advanced reactor designs may rely on a system, inherent feedback mechanism, or some combination thereof to shut down the reactor and maintain a safe shutdown under design-basis event conditions. SECY-94-084, Policy and Technical Issues Associated with the Regulatory Treatment of Non-Safety Systems in Passive Plant Designs (Ref. 29), describes the characteristics of a safe shutdown condition as reactor subcriticality, decay heat removal, and radioactive materials containment. The staff replaced postulated accident conditions with design-basis event Appendix B to DG-1330, Page B-11

APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA III. Reactivity Control Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC conditions, to emphasize that plants are required to maintain a safe shutdown following AOOs as well as postulated accidents.

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 serves as a backup to the safety-related 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). Independent indicates no shared systems or components with the safety-related means and diverse indicates a different design than the safety-related means. The purpose of an independent and diverse means of controlling reactivity is to preclude a potential common cause failure affecting both means of reactivity control, which would lead to the inability to shut down the reactor. The second means of reactivity control does not have to demonstrate that design limits for fission product barriers are met.

Additionally, the current GDC 26, third sentence, states that the second reactivity control system shall be capable of reliably controlling the rate of changes resulting from planned, normal power changes (including xenon burnout) to assure acceptable fuel design limits are not exceeded. Staff has identified this as an operational requirement that is not necessary to ensure reactor safety provided a design complies with ARDC 26(1). Therefore, this sentence is not retained in ARDC 26.

27 Combined reactivity control systems capability.

Same as ARDC DELETEDInformation incorporated into SFR 26 Appendix B to DG-1330, Page B-12

APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA III. Reactivity Control Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC 28 Reactivity limits. Reactor coolant pressure boundary has been relabeled as primary The reactivity control systems shall be designed with coolant boundary to conform to standard terms used in the LMR appropriate limits on the potential amount and rate of reactivity industry. The use of the term primary indicates that the SFR-DC increase to ensure that the effects of postulated reactivity are applicable to the primary cooling system, not the intermediate accidents can neither (1) result in damage to the primary coolant cooling system.

boundary greater than limited local yielding nor (2) sufficiently disturb the core, its support structures or other reactor vessel The list of postulated reactivity accidents has been deleted.

internals to impair significantly the capability to cool the core.

29 Protection against anticipated operational occurrences.

Same as GDC IV. Fluid Systems Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC 30 Quality of primary coolant boundary. Reactor coolant pressure boundary has been relabeled as Components that are part of the primary coolant boundary shall primary coolant boundary to conform to standard terms used in be designed, fabricated, erected, and tested to the highest the LMR industry.

quality standards practical. Means shall be provided for detecting and, to the extent practical, identifying the location of The use of the term primary indicates that the SFR-DC are the source of primary coolant leakage. applicable only to the primary cooling system, not the intermediate cooling system.

The cover gas boundary is included as part of the reactor primary coolant boundary per NUREG-1368 (page 3-38).

31 Fracture prevention of primary coolant boundary. Reactor coolant pressure boundary has been relabeled as The primary coolant boundary shall be designed with sufficient primary coolant boundary to conform to standard terms used in margin to ensure that, when stressed under operating, the LMR industry.

maintenance, testing, and postulated accident conditions, (1) the boundary behaves in a nonbrittle manner and (2) the probability of rapidly propagating fracture is minimized. The Appendix B to DG-1330, Page B-13

APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA IV. Fluid Systems Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC design shall reflect consideration of service temperatures, The use of the term primary indicates that the SFR-DC are service degradation of material properties, creep, fatigue, stress applicable only to the primary cooling system, not the intermediate rupture, and other conditions of the boundary material under cooling system.

operating, maintenance, testing, and postulated accident conditions and the uncertainties in determining (1) material Specific examples are added to the SFR-DC to account for the high properties, (2) the effects of irradiation and coolant chemistry design and operating temperatures and sodium coolant.

on material properties, (3) residual, steady-state, and transient stresses, and (4) size of flaws. The cover gas boundary is included as part of the reactor primary coolant boundary (referred to as RCPB by PRISM) per NUREG-1368 (page 3-38) (Ref. 4).

32 Inspection of primary coolant boundary. Reactor coolant pressure boundary has been relabeled as Components that are part of the primary coolant boundary shall primary coolant boundary to conform to standard terms used in be designed to permit (1) periodic inspection and functional the LMR industry.

testing of important areas and features to assess their structural and leaktight integrity, and (2) an appropriate material The use of the term primary indicates that the SFR-DC are surveillance program for the reactor vessel. applicable only to the primary cooling system, not the intermediate cooling system.

The cover gas boundary is included as part of the reactor primary coolant boundary (referred to as RCPB by PRISM) per NUREG-1368 (page 3-38) (Ref.4).

The staff modified the LWR GDC by replacing the term reactor pressure vessel with reactor vessel, which the staff believes is a more generically applicable term.

33 Primary coolant inventory maintenance. This SFR-DC was retitled as inventory maintenance to provide A system to maintain primary coolant inventory for protection more flexibility for advanced reactor designs.

against small breaks in the primary coolant boundary shall be provided as necessary to ensure that specified acceptable fuel The first sentence is modified so that it ends with ...shall be design limits are not exceeded as a result of primary coolant provided as necessary and is combined with the second sentence inventory loss due to leakage from the primary coolant as necessary to ensure (without the opening phrase, The boundary and rupture of small piping or other small system safety function shall be) to recognize that the inventory Appendix B to DG-1330, Page B-14

APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA IV. Fluid Systems Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC components that are part of the boundary. The system shall be control system may be unnecessary for some designs to maintain designed to ensure that the system safety function can be safety functions that ensure fuel design limits are not exceeded.

accomplished using the piping, pumps, and valves used to maintain primary coolant inventory during normal reactor Reactor coolant pressure boundary has been relabeled as operation. primary coolant boundary to reflect that the SFR primary system operates at low pressure and to conform to standard terms used in the LMR industry.

The SFR primary coolant boundary design requirements differ from the traditional LWR requirements. The effects of low-pressure design are acknowledged in NUREG-1368 (page 3-28) (Ref. 4), in the discussion of GDC 4, and on (page 3-30), under GDC 14. The use of the term primary implies the GDC is applicable to the primary cooling system, not the intermediate cooling system.

Both pool- and loop-type SFR designs limit loss of primary coolant so that an inventory adequate to perform the safety function of the residual heat removal system is maintained under operating, maintenance, testing, and postulated accident conditions.

The GDC reference to electric power was removed. Refer to SFR-DC 17 concerning those systems that require electric power.

34 Residual heat removal. In most advanced reactor designs the residual heat removal system A system to remove residual heat shall be provided. For normal is designed to meet the requirements of SFR-DC 34 and SFR-DC operations and anticipated operational occurrences, the system 35 (for more discussion see NUREG-0968 (Ref. 5) and NUREG-safety function shall be to transfer fission product decay heat 1368 (Ref. 4)).

and other residual heat from the reactor core at a rate such that specified acceptable fuel design limits and the design It is anticipated that the RHR system for non-LWRs will have the conditions of the primary coolant boundary are not exceeded. same regulatory treatment as the current LWR fleet.

Suitable redundancy in components and features and suitable Reactor coolant pressure boundary has been relabeled as interconnections leak detection, and isolation capabilities, shall primary coolant boundary to reflect that the SFR primary system operates at low-pressure and to conform to standard terms used in Appendix B to DG-1330, Page B-15

APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA IV. Fluid Systems Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC be provided to ensure that the system safety function can be the LMR industry. The use of the term primary indicates that the accomplished, assuming a single failure. SFR-DC are applicable to the primary cooling system, not the intermediate cooling system.

The second paragraph addresses residual heat removal system redundancy.

The discussion related to sodium leakage and required barriers was moved to a new SFR-DC 78.

The GDC reference to electric power was removed. Refer to SFR-DC17 concerning those systems that require electric power.

35 Emergency core cooling. In most advanced reactor designs, a single system (i.e the residual Same as ARDC heat removal system) is provided to perform both the residual heat A system to provide sufficient emergency core cooling shall be removal and emergency core cooling functions. In this case, the provided. The system safety function shall be to transfer heat single system would be designed to meet the requirements of SFR-from the reactor core such that effective core cooling is DC 34 and SFR-DC 35. (for more discussion see NUREG-0968 maintained and fuel damage is limited. (Ref. 5) and NUREG-1368 (Ref. 4)) However, the staff acknowledges that this may not be the case for every advanced Suitable redundancy in components and features and suitable reactor design. Therefore, to allow current and future non-LWR interconnections, leak detection, isolation, and containment designers the flexibility to provide a single system or multiple capabilities shall be provided to ensure that the system safety systems to perform residual heat removal and emergency core function can be accomplished, assuming a single failure. cooling, the staff decided to keep the SFR-DC 34 and SFR-DC 35 separate in lieu of combining them into a single criterion. Effective core cooling may include maintaining the primary coolant boundary in a condition necessary for adequate postulated accident heat removal. The staffs approach to provide two separate criteria is consistent with the approach taken in the LWR GDCs.

This change removes the light-water reactor emphasis on loss of coolant accidents that may not apply to every design. Loss of Appendix B to DG-1330, Page B-16

APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA IV. Fluid Systems Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC coolant accidents may still require analysis in conjunction with postulated accidents if they are relevant to the design.

The discussion related to sodium leakage and required barriers was moved to a new SFR-DC 78.

The GDC reference to electric power was removed. Refer to SFR-DC17 concerning those systems that require electric power.

36 Inspection of residual heat removal system. In most advanced reactor designs, a single system (i.e the residual Same as ARDC heat removal system) is provided to perform both the residual heat A system that provides emergency core cooling shall be removal and emergency core cooling functions. In this case, the designed to permit appropriate periodic inspection of important single system would be designed to meet the requirements of SFR-components to ensure the integrity and capability of the system. DC 34 and SFR-DC 35. (for more discussion see NUREG-0968 (Ref. 5) and NUREG-1368 (Ref. 4)) However, the staff acknowledges that this may not be the case for every advanced reactor design. Therefore, to allow current and future non-LWR designers the flexibility to provide a single system or multiple systems to perform residual heat removal and emergency core cooling, the staff decided to keep the SFR-DC 34 and SFR-DC 35 separate in lieu of combining them into a single criterion. The staffs approach to provide two separate criteria is consistent with the approach taken in the LWR GDCs.

The SFR-DC has slightly different wording than the GDC to clarify the scope of the criteria. Any system, or portions of a system, credited with an emergency core cooling function during postulated accidents (for example, a system that performs both the residual heat removal function and the emergency core cooling function) would need to meet SFR-DC 36.

The list of examples has been deleted because it applies to LWR designs, and each specific design will have different important Appendix B to DG-1330, Page B-17

APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA IV. Fluid Systems Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC components associated with residual heat removal. This revision allows for a technology-neutral SFR-DC.

Review of the proposed DOE SFR and HTGR DC found that only SFR provided specific examples of important components but were generic in nature and did not include any significant additional guidance.

37 Testing of residual heat removal system. In most advanced reactor designs, a single system (i.e the residual Same as ARDC heat removal system) is provided to perform both the residual heat A system that provides emergency core cooling shall be removal and emergency core cooling functions. In this case, the designed to permit appropriate periodic functional testing to single system would be designed to meet the requirements of SFR-ensure (1) the structural and leaktight integrity of its DC 34 and SFR-DC 35. (for more discussion see NUREG-0968 components, (2) the operability and performance of the system (Ref. 5) and NUREG-1368 (Ref. 4)) However, the staff components, and (3) the operability of the system as a whole acknowledges that this may not be the case for every advanced and, under conditions as close to design as practical, the reactor design. Therefore, to allow current and future non-LWR performance of the full operational sequence that brings the designers the flexibility to provide a single system or multiple system into operation, including operation of any associated systems to perform residual heat removal and emergency core systems and interfaces necessary to transfer decay heat to the cooling, the staff decided to keep the SFR-DC 34 and SFR-DC 35 ultimate heat sink. separate in lieu of combining them into a single criterion. The staffs approach to provide two separate criteria is consistent with the approach taken in the LWR GDCs.

The SFR-DC has slightly different wording than the GDC to clarify the scope of the criteria. Any system, or portions of a system, credited with an emergency core cooling function during postulated accidents (for example, a system that performs both the residual heat removal function and the emergency core cooling function) would need to meet SFR-DC 37.

Specific mention of pressure testing has been removed yet remains a potential requirement should it be necessary as a Appendix B to DG-1330, Page B-18

APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA IV. Fluid Systems Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC component of appropriate periodic functional testing... of cooling systems.

A non-leaktight system may be acceptable for some designs provided that (1) the system leakage does not impact safety functions under all conditions, and (2) defense in depth is not impacted by system leakage.

Active has been deleted in item (2) as appropriate operability and performance system component testing are required, regardless of an active or passive nature.

Reference to the operation of applicable portions of the protection system, cooling water system, and power transfers is considered part of the more general associated systems. Together with the ultimate heat sink, they are part of the operability testing of the system as a whole.

The GDC reference to electric power was removed. Refer to SFR-DC17 concerning those systems that require electric power.

38 Containment heat removal. Same as ARDC as necessary is meant to condition an SFR-DC 38 A system to remove heat from the reactor containment shall be application to designs requiring heat removal for conventional provided as necessary to maintain the containment pressure and containments that are found to require heat removal measures.

temperature within acceptable limits following postulated accidents. The LOCA reference has been removed to provide for any postulated accident that might affect the containment structure.

Suitable redundancy in components and features and suitable interconnections, leak detection, isolation, and containment Containment structure safety system redundancy is addressed in the capabilities shall be provided to ensure that the system safety second paragraph.

function can be accomplished, assuming a single failure.

39 Inspection of containment heat removal system. Examples were deleted to make the SFR-DC technology neutral.

Same as ARDC Appendix B to DG-1330, Page B-19

APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA IV. Fluid Systems Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC The containment heat removal system shall be designed to permit appropriate periodic inspection of important components to ensure the integrity and capability of the system.

40 Testing of containment heat removal system. Specific mention of pressure testing has been removed yet Same as ARDC remains a potential requirement should it be necessary as a The containment heat removal system shall be designed to component of appropriate periodic functional testing... of permit appropriate periodic functional testing to ensure (1) the containment heat removal.

structural and leaktight integrity of its components, (2) the operability and performance of the system components, and A non-leaktight system may be acceptable for some designs (3) the operability of the system as a whole, and under provided that (1) the system leakage does not impact safety conditions as close to the design as practical, the performance functions under all conditions, and (2) defense in depth is not of the full operational sequence that brings the system into impacted by system leakage.

operation, including the operation of associated systems.

Reference to the operation of applicable portions of the protection system, structural and equipment cooling water systems, and power transfers is considered part of the more general associated systems for operability testing of the system as a whole.

The GDC reference to electric power was removed. Refer to SFR-DC17 concerning those systems that require electric power.

41 Containment atmosphere cleanup. Advanced reactors offer potential for reaction product generation Same as ARDC that is different from that associated with clad metal-water Systems to control fission products and other substances that interactions. Therefore, the terms hydrogen and oxygen are may be released into the reactor containment shall be provided removed while other substances is retained to allow for as necessary to reduce, consistent with the functioning of other exceptions.

associated systems, the concentration and quality of fission products released to the environment following postulated The GDC reference to electric power was removed. Refer to SFR-accidents and to control the concentration of other substances in DC17 concerning those systems that require electric power.

the containment atmosphere following postulated accidents to ensure that containment integrity is maintained.

Appendix B to DG-1330, Page B-20

APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA IV. Fluid Systems Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC Each system shall have suitable redundancy in components and features and suitable interconnections, leak detection, isolation, and containment capabilities to ensure that its safety function can be accomplished, assuming a single failure.

42 Inspection of containment atmosphere cleanup systems.

Same as GDC 43 Testing of containment atmosphere cleanup systems. Active has been deleted in item (2), as appropriate operability Same as ARDC and performance testing of system components is required The containment atmosphere cleanup systems shall be designed regardless of an active or passive nature, as are cited examples of to permit appropriate periodic functional testing to ensure active system components.

(1) the structural and leaktight integrity of its components, (2) the operability and performance of the system components, Examples of active systems under item (2) have been deleted, both and (3) the operability of the systems as a whole and, under to conform to similar wording in ARDC 37 and 40 and ensure that conditions as close to design as practical, the performance of passive as well as active system components are considered.

the full operational sequence that brings the systems into operation, including the operation of associated systems. Specific mention of pressure testing has been removed yet remains a potential requirement should it be necessary as a component of appropriate periodic functional testing... of cooling systems. A non-leaktight system may be acceptable for some designs provided that (1) the system leakage does not impact safety functions under all conditions and (2) defense in depth is not impacted by system leakage.

The GDC reference to electric power was removed. Refer to SFR-DC17 concerning those systems that require electric power.

44 Structural and equipment cooling. This renamed SFR-DC accounts for advanced reactor design Same as ARDC system differences to include cooling requirements for SSCs, if A system and components important to safety to an ultimate applicable; this SFR-DC does not address the residual heat removal heat sink shall be provided, as necessary, to transfer the system required under SFR-DC 34, and ECCS system under SFR-combined heat load of these structures, systems, and DC 35.

components under normal operating and accident conditions.

Appendix B to DG-1330, Page B-21

APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA IV. Fluid Systems Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC Suitable redundancy in components and features and suitable The GDC reference to electric power was removed. Refer to SFR-interconnections, leak detection, and isolation capabilities shall DC17 concerning those systems that require electric power.

be provided to ensure that the system safety function can be accomplished, assuming a single failure.

45 Inspection of structural and equipment cooling systems. This renamed ARDC accounts for advanced reactor system design Same as ARDC differences to include possible cooling requirements for SSCs The structural and equipment cooling systems shall be designed important to safety.

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

46 Testing of structural and equipment cooling systems. This renamed ARDC accounts for advanced reactor system design Same as ARDC differences to include possible cooling requirements for SSCs The structural and equipment cooling systems shall be designed important to safety. Specific mention of pressure testing has been to permit appropriate periodic functional testing to ensure removed yet remains a potential requirement should it be necessary (1) the structural and leaktight integrity of their components, as a component of appropriate periodic functional testing... of (2) the operability and performance of the system components, cooling systems. A non-leaktight system may be acceptable for and (3) the operability of the systems as a whole and, under some designs provided that (1) the system leakage does not impact conditions as close to design as practical, the performance of safety functions under all conditions and (2) defense in depth is not the full operational sequences that bring the systems into impacted by system leakage.

operation for reactor shutdown and postulated accidents, including the operation of associated systems. Active has been deleted in item (2) because appropriate operability and performance tests of system components are required regardless of their active or passive nature. The LOCA reference has been removed to provide for any postulated accident that might affect subject SSCs.

The GDC reference to electric power was removed. Refer to SFR-DC17 concerning those systems that require electric power.

Appendix B to DG-1330, Page B-22

APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA V. Reactor Containment Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC 50 Containment design basis. SFR-DC 50 specifically addresses a containment structure in the Same as ARDC opening sentence and SFR-DC 51-57 support the containment The reactor containment structure, including access openings, structures design basis. Therefore, SFR-DC 51-57 are modified by penetrations, and the containment heat removal system shall be adding the word structure to highlight the containment structure-designed so that the containment structure and its internal specific criteria.

compartments can accommodate, without exceeding the design leakage rate and with sufficient margin, the calculated pressure The phrase loss-of-coolant accident is LWR specific because this and temperature conditions resulting from postulated accidents. is understood to be the limiting containment structure accident for This margin shall reflect consideration of (1) the effects of an LWR design. It is replaced by the phrase postulated accident potential energy sources that have not been included in the to allow for consideration of the design-specific containment determination of the peak conditions, (2) the limited experience structure limiting accident for non-LWR designs.

and experimental data available for defining accident phenomena and containment responses, and (3) the The example at the end of subpart 1 of the ARDC is LWR specific conservatism of the calculational model and input parameters and therefore deleted 51 Fracture prevention of containment pressure boundary. SFR-DC 51-57 support SFR-DC 50, which specifically applies to Same as ARDC non-LWR designs that use a fixed containment structure.

The boundary of the reactor containment structure shall be Therefore, the word structure is added to each of these SFR-DC designed with sufficient margin to ensure that, under operating, to clearly convey the understanding that this criterion applies to maintenance, testing, and postulated accident conditions, (1) its designs employing containment structures.

materials behave in a nonbrittle manner and (2) the probability The term ferritic was removed to avoid limiting the scope of the of rapidly propagating fracture is minimized. The design shall criterion to ferritic materials. With this revision, the staff believes reflect consideration of service temperatures and other that this criterion is more broadly applicable to all non-LWR conditions of the containment boundary materials during designs.

operation, maintenance, testing, and postulated accident conditions, and the uncertainties in determining (1) material The word pressure was left in the title to reflect that, while a properties, (2) residual, steady-state, and transient stresses, and design might not have a high-pressure containment like a (3) size of flaws. traditional LWR, the containment still serves a pressure-retaining function.

52 Capability for containment leakage rate testing. SFR-DC 51-57 support SFR-DC 50, which specifically applies to Same as ARDC non-LWR designs that use a fixed containment structure.

The reactor containment structure and other equipment that Therefore, the word structure is added to each of these SFR-DC may be subjected to containment test conditions shall be Appendix B to DG-1330, Page B-23

APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA V. Reactor Containment Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC designed so that periodic integrated leakage rate testing can be to clearly convey the understanding that this criterion applies to conducted at containment design pressure. designs employing containment structures.

53 Provisions for containment testing and inspection. SFR-DC 51-57 support SFR-DC 50, which specifically applies to Same as ARDC non-LWR designs that use a fixed containment structure.

The reactor containment structure shall be designed to permit Therefore, the word structure is added to each of these SFR-DC (1) appropriate periodic inspection of all important areas, such to clearly convey the understanding that this criterion only applies as penetrations, (2) an appropriate surveillance program, and to designs employing containment structures.

(3) periodic testing at containment design pressure of the leak-tightness of penetrations that have resilient seals and expansion bellows.

54 Piping systems penetrating containment. The word structure was added to this SFR-DC to clearly convey Piping systems penetrating the reactor containment structure the understanding that this criterion only applies to designs shall be provided with leak detection, isolation, and employing containment structures.

containment capabilities that have redundancy, reliability, and performance capabilities necessary to perform the containment Not all penetrations will provide a release path to the atmosphere.

safety function and that reflect the importance to safety of Piping that may be of interest in the case of an SFR design is for preventing radioactivity releases from containment through the intermediate heat transport system and the residual heat these piping systems. Such piping systems shall be designed removal system. Based on stakeholder input, a designer may be with the capability to verify, by testing, the operational able to satisfactorily demonstrate that containment isolation valves readiness of any isolation valves and associated apparatus are not required for an SFR design. This rewording for the SFR-DC periodically and to confirm that valve leakage is within provides a designer the opportunity to present the safety case acceptable limits. without containment isolation valves and the associated need for testing. Otherwise, NUREG-1368 (page 3-51) indicated that GDC 54 was applicable as written.

American National Standards Institute/American Nuclear Society (ANSI/ANS)-54.1-1989 recommended revising the phrase containment capabilities having redundancy, reliability, and performance capabilities which reflect the importance to safety of isolating these piping systems. to containment capabilities as required to perform the containment safety function.

Appendix B to DG-1330, Page B-24

APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA V. Reactor Containment Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC The adjustment to the last sentence enhances the clarity of the sentence with respect to the latest terminology used for valve periodic verification and operational readiness. It also removes the introductory statement, as the definition of required could be confusingthe designer will present the safety case for what is necessary, and the NRC staff will review it.

The American Society of Mechanical Engineers (ASME) Operation and Maintenance of Nuclear Power Plants, Division 1: OM Code:

Section IST (ASME OM Code) defines operational readiness as the ability of a component to perform its specified functions. The ASME OM Code is incorporated by reference in the NRC regulations in 10 CFR 50.55a, including the definition of operational readiness for pumps, valves, and dynamic restraints.

55 Primary coolant boundary penetrating containment. The word structure was added to this SFR-DC to clearly convey Each line that is part of the primary coolant boundary and that the understanding that this criterion only applies to designs penetrates the reactor containment structure shall be provided employing containment structures. In some cases, the word the with containment isolation valves as follows, unless it can be was also added to make the phrase grammatically correct.

demonstrated that the containment isolation provisions for a specific class of lines, such as instrument lines, are acceptable The title of SFR-DC 55 is the Primary coolant boundary on some other defined basis: penetrating containment. The SFR intermediate loop is a separate (1) One locked closed isolation valve inside and one locked closed system that does not allow any direct mixing of intermediate closed isolation valve outside containment; or fluid with the primary coolant sodium. The tubing of the IHX and (2) One automatic isolation valve inside and one locked closed associated intermediate loop piping inside the reactor vessel are a isolation valve outside containment; or part of the primary coolant boundary. SFR-DC 57, Closed system (3) One locked closed isolation valve inside and one automatic isolation valves, addresses closed systems that penetrate isolation valve outside containment. A simple check valve containment and would be the appropriate place to address a closed may not be used as the automatic isolation valve outside system, such as an intermediate loop, that penetrates containment containment; or and is not part of the primary coolant boundary (in its entirety).

(4) One automatic isolation valve inside and one automatic This is similar to the treatment of the main steam system and the isolation valve outside containment. A simple check valve steam generator in a pressurized-water reactor.

Appendix B to DG-1330, Page B-25

APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA V. Reactor Containment Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC may not be used as the automatic isolation valve outside Reactor coolant pressure boundary has been relabeled as containment. primary coolant boundary to reflect that the SFR primary system operates at low pressure and to conform to standard terms used in Isolation valves outside containment shall be located as close to the LMR industry. The use of the term primary implies the containment as practical and, upon loss of actuating power, SFR-DC are applicable to the primary cooling system, not the automatic isolation valves shall be designed to take the position intermediate cooling system.

that provides greater safety.

Other appropriate requirements to minimize the probability or consequences of an accidental rupture of these lines or of lines connected to them shall be provided as necessary to ensure adequate safety. Determination of the appropriateness of these requirements, such as higher quality in design, fabrication, and testing, additional provisions for inservice inspection, protection against more severe natural phenomena, and additional isolation valves and containment, shall include consideration of the population density, use characteristics, and physical characteristics of the site environs.

Appendix B to DG-1330, Page B-26

APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA V. Reactor Containment Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC 56 Containment isolation. SFR-DC 51-57 support SFR-DC 50, which specifically applies to Same as ARDC non-LWR designs that use a fixed containment structure.

Each line that connects directly to the containment atmosphere Therefore, the word structure is added to each of these SFR-DC and penetrates the containment structure shall be provided with to clearly convey the understanding that this criterion only applies containment isolation valves as follows, unless it can be to designs employing containment structures. The word primary demonstrated that the containment isolation provisions for a in the title and the text was removed, and the word reactor was specific class of lines, such as instrument lines, are acceptable also removed because the containment is a barrier between the on some other defined basis: fission products and the environment. There are diverse advanced (1) One locked closed isolation valve inside and one locked reactor designs and, hence, there is no single containment concept.

closed isolation valve outside containment; or In all cases, the rules for containment penetrations to fulfill (2) One automatic isolation valve inside and one locked closed containment isolation would apply. How this is accomplished isolation valve outside containment; or should be left to the designer of the particular advanced reactor (3) One locked closed isolation valve inside and one automatic design, without being too prescriptive as to whether it is a primary isolation valve outside containment. A simple check valve or secondary or reactor containment. There may be a need for a may not be used as the automatic isolation valve outside containment structure outside the reactor region. For example, in containment; or the MSR design, some of the molten fuel salt is drawn off to a (4) One automatic isolation valve inside and one automatic processing system to clean it up and remove fission products before isolation valve outside containment. A simple check valve returning it to the reactor. The molten fuel salt is highly radioactive may not be used as the automatic isolation valve outside and would need a containment around the entire system.

containment. Alternatively, in an SFR, the guard vessel would be the primary containment and, in the case of the PRISM design, a dome-shaped structure above it that would be the secondary containment. The secondary containment also has penetrations and needs containment isolation requirements to be fulfilled.

57 Closed system isolation valves. The word structure was added to this SFR-DC to clearly convey Each line that penetrates the reactor containment structure and the understanding that this criterion applies to designs employing is neither part of the primary coolant boundary nor connected containment structures. In some cases, the word the was also directly to the containment atmosphere shall have at least one added to make the phrase grammatically correct.

containment isolation valve unless it can be demonstrated that the containment safety function can be met without an isolation Reactor coolant pressure boundary has been relabeled as valve and assuming failure of a single active component. The primary coolant boundary to reflect that the SFR primary system isolation valve, if required, shall be either automatic, or locked operates at low-pressure and to conform to standard terms used in Appendix B to DG-1330, Page B-27

APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA V. Reactor Containment Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC closed, or capable of remote manual operation. This valve shall the LMR industry. The use of the term primary implies the SFR-be outside containment and located as close to the containment DC are applicable to the primary cooling system, not the as practical. A simple check valve may not be used as the intermediate cooling system.

automatic isolation valve.

VI. Fuel and Reactivity Control Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC 60 Control of releases of radioactive materials to the environment.

Same as GDC 61 Fuel storage and handling and radioactivity control. The underlying concept of establishing functional requirements for Same as ARDC radioactivity control in fuel storage and fuel handling systems is The fuel storage and handling, radioactive waste, and other independent of the design of non-LWR reactors. However, some systems that may contain radioactivity shall be designed to advanced designs may use dry fuel storage that incorporates ensure adequate safety under normal and postulated accident cooling jackets that can be liquid cooled or air cooled to remove conditions. These systems shall be designed (1) with a heat. This modification to this GDC allows for both liquid and air capability to permit appropriate periodic inspection and testing cooling of the dry fuel storage containers.

of components important to safety, (2) with suitable shielding for radiation protection, (3) with appropriate containment, 62 Prevention of criticality in fuel storage and handling.

Same as GDC 63 Monitoring fuel and waste storage.

Same as GDC 64 Monitoring radioactivity releases. In NUREG-1368, Table 3.3 (page 3-25), the NRC staff Means shall be provided for monitoring the reactor containment recommended deleting the GDC 64 phrase spaces containing atmosphere, spaces containing components for primary system components for recirculation of loss-of-coolant accident fluids.

sodium and cover gas cleanup and processing, effluent Otherwise, the NRC staff noted that criterion requirements are discharge paths, and the plant environs for radioactivity that independent of the design of SFRs (page 3-55).

Appendix B to DG-1330, Page B-28

APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA VI. Fuel and Reactivity Control Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC may be released from normal operations, including anticipated The staff added text to identify other SFR plant areas that should operational occurrences, and from postulated accidents. also be included to maintain consideration of all potential discharge paths and areas subject to monitoring. Therefore, primary system sodium and cover gas cleanup systems that may be outside containment and effluent processing systems are considered in place of the current text.

VII. Additional SFR-DC Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC 70 Intermediate coolant system. NUREG-1368 (Ref. 4) (page 3-57), Section 3.2.4.5, suggested the need for a separate criterion for the intermediate coolant system.

If an intermediate coolant system is provided, then the system Also, separate criteria were included in NUREG-0968 (Ref. 5) shall be designed to transport heat from the primary coolant (Criterion 31, Design of Intermediate Cooling System, and system to the energy conversion system as required. Criterion 33, Inspection of Intermediate Cooling System).

The intermediate coolant system shall be designed with The staff revised SFR-DC 70 to focus on the function of the sufficient margin to assure that (1) the design conditions of the intermediate coolant system, and to use language that is consistent intermediate coolant boundary are not exceeded during normal with other design criteria. The discussion related to sodium leakage operations, including anticipated occupational occurrences, and and required barriers was moved to a new SFR-DC 78.

(2) the integrity of the primary coolant boundary is maintained during intermediate coolant system accidents. Assurance that components of the intermediate coolant system are also designed, as necessary, to prevent the transport of radionuclides between the primary coolant system and the energy conversion system is provided by the design criteria proposed for the intermediate coolant boundary (SFR-DC 75, SFR-DC 76, and SFR-DC 77).

Examples of intermediate coolant system accidents would include:

rupture (including at a location in the steam-sodium generator), loss of flow, overcooling conditions, and undercooling conditions.

Appendix B to DG-1330, Page B-29

APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA VII. Additional SFR-DC Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC 71 Primary coolant & cover gas purity control. The NRC considered DOEs proposed SFR-DC 71 and made changes based on the Response to NRC Staff Questions on the Systems shall be provided as necessary to maintain the purity of U.S. Department of Energy Report, Guidance for Developing primary coolant sodium and cover gas within specified design Principal Design Criteria for Advanced Non-Light Water limits. These limits shall be based on consideration of Reactors (pages 12-13).

(1) chemical attack, (2) fouling and plugging of passages, and (3) radionuclide concentrations, and (4) air or moisture ingress NUREG 1368 (Ref. 4) (page 3-57), Section 3.2.4.6, suggested the as a result of a leak of cover gas. need for a separate criterion for a sodium and cover gas purity control. Also a separate criterion was included in NUREG-0968 (Ref. 5) (Criterion 34, Reactor and Intermediate Coolant and Cover Gas Purity Control).

72 Sodium heating systems. The NRC considered DOEs proposed SFR-DC 72 and made Heating systems shall be provided for systems and components changes based on the Response to NRC Staff Questions on the important to safety, which contain or could be required to U.S. Department of Energy Report, Guidance for Developing contain sodium. These heating systems and their controls shall Principal Design Criteria for Advanced Non-Light Water be appropriately designed to ensure that the temperature Reactors (pages 13-14).

distribution and rate of change of temperature in systems and components containing sodium are maintained within design NUREG-1368 (Ref. 4) (page 3-56)), Section 3.2.4.2, suggested the limits assuming a single failure. If plugging of any cover gas need for a separate criterion for sodium heating system. Also, a line due to condensation or plate out of sodium aerosol or vapor separate criterion was included in NUREG-0968 (Ref. 5) could prevent accomplishing a safety function, the temperature (Criterion 7, Sodium Heating Systems).

control and the relevant corrective measures associated with that line shall be considered important to safety. The phrase and the relevant corrective measures has been added, in case the cover gas line design includes a feature for clearing an obstruction resulting from condensation or plate out of sodium aerosol or vapor.

73 Sodium leakage detection and reaction prevention and The NRC considered DOEs proposed SFR-DC 73 and made mitigation. changes based on the Response to NRC Staff Questions on the Means to detect sodium leakage and to limit and control the U.S. Department of Energy Report, Guidance for Developing extent of sodium-air and sodium-concrete reactions and to Principal Design Criteria for Advanced Non-Light Water mitigate the effects of fires resulting from these sodium-air and Reactors (pages 15-16).

sodium-concrete reactions shall be provided to ensure that the Appendix B to DG-1330, Page B-30

APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA VII. Additional SFR-DC Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC safety functions of structures, systems, and components NUREG-1368 (Ref. 4) (page 3-56), Section 3.2.4.1, suggested the important to safety are maintained. Special features, such as need for a separate criterion for protection against sodium reactions.

inerted enclosures or guard vessels, shall be provided for Also, a separate criterion was included in NUREG-0968 (Ref. 5) systems containing sodium. (Criterion 4, Protection against Sodium and NaK reactions).

74 Sodium/water reaction prevention/mitigation. The NRC considered DOEs proposed SFR-DC 74 and made Structures, systems, and components containing sodium shall be changes based on the Response to NRC Staff Questions on the designed and located to avoid contact between sodium and U.S. Department of Energy Report, Guidance for Developing water and to limit the adverse effects of chemical reactions Principal Design Criteria for Advanced Non-Light Water between sodium and water on the capability of any structure, Reactors (pages 16-18). NUREG-1368 (Ref 4) (page 3-56),

system, or component to perform any of its intended safety Section 3.2.4.1, suggested the need for a separate criterion for functions. If steam-water is used for energy conversion, to protection against sodium reactions. Also, a separate criterion was prevent loss of any plant safety function, the sodium-steam included in NUREG-0968 (Ref. 5) (Criterion 4, Protection against generator system shall be designed to detect and contain Sodium and NaK reactions). Fire considerations are added for sodium-water reactions and limit the effects of the energy and consistency with SFR-DC 73.

reaction products released by such reactions, including mitigation of the effects of any resulting fire involving sodium.

75 Quality of the intermediate coolant boundary. This criterion is similar to GDC 30 in 10 CFR Part 50, Appendix A, Components that are part of the intermediate coolant boundary and is intended to ensure that, similar to the reactor coolant pressure shall be designed, fabricated, erected, and tested to quality boundary, the intermediate coolant boundary is designed, standards commensurate with the importance of the safety fabricated, and tested using quality standards and controls sufficient functions to be performed. to ensure that failure of the intermediate system would be unlikely.

76 Fracture prevention of the intermediate coolant boundary. This criterion is similar to GDC 31 in 10 CFR Part 50, Appendix A, The intermediate coolant boundary shall be designed with and is intended to ensure that, similar to the reactor coolant pressure sufficient margin to ensure that, when stressed under operating, boundary, the intermediate coolant boundary is designed to avoid maintenance, testing, and postulated accident conditions, (1) the brittle and rapidly propagating facture modes.

boundary behaves in a nonbrittle manner and (2) the probability of rapidly propagating fracture is minimized. The second sentence related to required analyses is removed to make the criteria more generic. In this manner, the design considerations may include, but are not limited to, those previously stated in the design criteria.

Appendix B to DG-1330, Page B-31

APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA VII. Additional SFR-DC Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC 77 Inspection of the intermediate coolant boundary. This criterion is similar to GDC 32 in 10 CFR Part 50, Appendix A, Components that are part of the intermediate coolant boundary and is intended to ensure that, similar to the reactor coolant pressure shall be designed to permit (1) periodic inspection and boundary, the intermediate coolant boundary is designed to avoid functional testing of important areas and features to assess their brittle and rapidly propagating fracture modes.

structural and leaktight integrity commensurate with the A non-leaktight system may be acceptable for some designs systems importance to safety, and (2) an appropriate material provided that (1) the system leakage does not impact safety surveillance program for the intermediate coolant boundary. functions under all conditions, and (2) defense in depth is not Means shall be provided for detecting and, to the extent impacted by system leakage.

practical, identifying the location of the source of coolant leakage. The staff added commensurate with the systems importance to safety. If leakage of the intermediate system constitutes a significant risk to the plant, then the appropriate inspection of the intermediate coolant boundary is necessary to ensure that the structural integrity of the boundary is maintained.

The requirement for an appropriate surveillance program is maintained to ensure that such a program is provided, as needed, to ensure that the integrity of the intermediate boundary is maintained.

At this time, the staff generally does not expect that the projected fluence on the intermediate boundary will be at levels that would necessitate a materials surveillance program that focuses on the impacts of irradiation embrittlement. However, the staff recognizes that this may not be the case for every design. In addition, a materials surveillance program may be used to monitor the effect of other environmental conditions on the boundary materials.

78 Primary Coolant System Interfaces The consequence of leakage between the primary coolant system and a heat removal system (i.e. RHR system, intermediate coolant When the primary coolant system interfaces with a structure, system) is more significant for primary coolant system (potentially system, or component containing fluid that is chemically impacting the fuel design limits or integrity of the primary coolant incompatible with the primary coolant, the interface location boundary) than it is for the heat removal system (coolant drawdown shall be designed to ensure that the primary coolant is separated or introduction of radioactive sodium).

from the chemically incompatible fluid by two redundant, Appendix B to DG-1330, Page B-32

APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA VII. Additional SFR-DC Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC passive barriers. When the primary coolant system interfaces Rather than creating two parallel requirements for the two systems, with a structure, system, or component containing fluid that is SFR-78 was created to discuss leakage and required barriers as a chemically compatible with the primary coolant, then the generic criterion. The criterion allows for double walled steam interface location may be a single passive barrier provided that generators, intermediate coolant systems connected to steam power the following conditions are met: system, and systems similar to the PRISM Direct Reactor Auxiliary Cooling System (DRACS).

(1) postulated leakage at the interface location does not result in failure of the intended safety functions of A paragraph from NUREG 1368 (page 3-41) was added describing structures, systems or components important to safety or the characteristics of the residual heat removal working fluid and its result in exceeding the fuel design limits associated operating pressure. This SFR-DC has been worded to explain that an intermediate coolant system may be used if the (2) the fluid contained in the structure, system, or primary coolant is not chemically compatible with the energy component is maintained at a higher pressure than the conversion system coolant.

primary coolant during normal operation, AOOs, shutdown, and accident conditions. A single passive barrier is adequate defense in depth when the heat removal working fluid is chemically compatible with the primary coolant, such that postulated leakage between the two systems does not result in the failure of any intended safety function of any SSC important to safety or cause fuel design limits to be exceeded.

An example is a heat removal system with liquid sodium potassium (NaK). A liquid sodium primary coolant system that is contaminated with NaK may have phase changes (e.g.,

solidification, boiling) at different temperatures, without adversely affecting the overall system. The postulated leakage may be based upon a leak-before-break analysis or the ability to detect leakage between the primary and intermediate coolant systems. If the working fluids are not chemically compatible, at least two passive barriers must separate the two systems.

The higher pressure requirement is to ensure any leakage in the interface between the two systems does not result in a release of Appendix B to DG-1330, Page B-33

APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA VII. Additional SFR-DC Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC radioactive primary coolant to the nonradioactive part of the heat transport system.

A sentence has been added to explain that this differential pressure requirement must be satisfied during AOOs and design-basis accidents, as well as during normal operating and shutdown conditions.

79 Cover gas inventory maintenance. This criterion is similar to GDC 33 in 10 CFR Part 50, Appendix A and SFR-DC 33 in this document. GDC 33 and SFR-DC 33 focus A system to maintain cover gas inventory shall be provided as on the effects of primary coolant (sodium) loss. A leak in a SFR necessary to ensure that the primary coolant sodium design primary coolant system may expel the cover gas rather than the limits are not exceeded as a result of cover gas loss due to primary coolant. The cover gas in the SFR performs an important to leakage from the primary coolant boundary and rupture of small safety function by protecting the sodium coolant from chemical piping or other small components that are part of the primary reactions. The staff created a new SFR-DC rather than adding the coolant boundary. cover gas in the term primary coolant. The term primary coolant sodium design limits is used to maintain consistent terminology with SFR-DC 71. The primary coolant sodium design limits consider the possibility of interactions between the primary coolant sodium and the primary coolant boundary or the fuel due to changes in the chemistry of the primary coolant sodium. The considerations include the possibility of (1) chemical attack, (2) fouling and plugging of passages, (3) radionuclide concentrations, and (4) air or moisture ingress as a result of a leak of cover gas.

The term as necessary is retained from SFR-DC 33 to permit designer flexibility if leakage of the system does not challenge the design limits of the primary coolant (for instance, an inerted containment filled with Argon).

Reactor coolant pressure boundary has been relabeled as primary coolant boundary to conform to standard terms used in the LMR industry.

Appendix B to DG-1330, Page B-34

APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA VII. Additional SFR-DC Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC The use of the term primary indicates that the SFR-DC are applicable only to the primary cooling system, not the intermediate cooling system.

The cover gas boundary is included as part of the primary coolant boundary (referred to as RCPB by PRISM) per NUREG-1368 (page 3-38).

The GDC reference to electric power was removed. Refer to SFR-DC17 concerning those systems that require electric power.

Appendix B to DG-1330, Page B-35

APPENDIX C MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGN CRITERIA The table below contains the modular high-temperature gas-cooled reactor design criteria (mHTGR-DC). These criteria are applicable to mHTGRs meeting the definition in the Glossary section of this RG. Applicants/designers may use the mHTGR-DC in this appendix to develop all or part of the principal design criteria (PDC) and may choose among the advanced reactor design criteria (ARDC)

(Appendix A), sodium-cooled fast reactor design criteria (SFR-DC) (Appendix B), or mHTGR-DC (Appendix C) to develop each PDC. Applicants/designers may also develop entirely new PDC as needed to address unique design features in their respective designs.

To develop these mHTGR-DC, the staff of the U.S. Nuclear Regulatory Commission (NRC) reviewed each general design criterion (GDC) in Appendix A, General Design Criteria for Nuclear Power Plants, to Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Domestic Licensing of Production and Utilization Facilities, to determine its applicability to mHTGR designs. The NRC staff then determined what if any adaptation was appropriate for mHTGRs. The results are included in column 2 of the table below. The table also includes the NRC staffs rationale for the adaptations. In many cases, the rationale refers to changes made to the language of the GDC. To fully understand the context of the rationale, the user of this RG should refer to the appropriate GDC. Where the NRC staff determined that the current GDC or the ARDC were applicable to the mHTGR-DC, the table denotes Same as GDC, or Same as ARDC, respectively. In many cases, the NRC staff determined the design criteria were not applicable to mHTGR designs. In these instances, the table denotes Not applicable to mHTGR.

The table consists of three columns:

Column 1Criterion Number Column 2mHTGR-DC Title and Content Column 3Staff Rationale for Adaptations to GDC The table is further divided into seven sections similar to those in 10 CFR Part 50, Appendix A:

Section IOverall Requirements (Criteria 1-5)

Section IIMultiple Barriers (Criteria 10-19)

Section IIIReactivity Control (Criteria 20-29)

Section IVFluid Systems (Criteria 30-46)

Section VReactor Containment (Criteria 50-57)

Section VIFuel and Radioactivity Control (Criteria 60-64)

Section VIIAdditional mHTGR-DC (Criteria 70-72)

Appendix C to DG-1330, Page C-1

APPENDIX C. MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGN CRITERIA I. Overall Requirements Criterion mHTGR-DC Title and Content NRC Rationale for Adaptions to GDC 1 Quality standards and records.

Same as GDC 2 Design bases for protection against natural phenomena.

Same as GDC 3 Fire protection. The phrase containing examples where noncombustible and fire-Same as ARDC resistant materials must be used has been broadened to apply to all Structures, systems, and components important to safety shall advanced reactor designs.

be designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and Instead of and, the phrase locations with structures, systems, explosions. Noncombustible and fire-resistant materials shall and components (SSCs) important to safety uses or, which is be used wherever practical throughout the unit, particularly in logically correct in this case.

locations with structures, systems, or components important to safety. Fire detection and fighting systems of appropriate capacity and capability shall be provided and designed to minimize the adverse effects of fires on structures, systems, and components important to safety. Firefighting systems shall be designed to ensure that their rupture or inadvertent operation does not significantly impair the safety capability of these structures, systems, and components.

4 Environmental and dynamic effects design bases. This change removes the light-water reactor (LWR) emphasis on Structures, systems, and components important to safety shall loss-of-coolant accidents (LOCAs) that may not apply to every be designed to accommodate the effects of and to be design. For example, helium is not needed in a mHTGR to remove compatible with the environmental conditions associated with heat from the core during postulated accidents and does not have normal operation, maintenance, testing, and postulated the same importance as water does to LWR designs to ensure that accidents. These structures, systems, and components shall be fuel integrity is maintained. Therefore, a specific reference to appropriately protected against dynamic effects, including the LOCAs is not applicable to all designs. LOCAs may still require effects of missiles originating both inside and outside the analysis in conjunction with postulated accidents if they are reactor helium pressure boundary, pipe whipping, and relevant to the design.

discharging fluids, that may result from equipment failures and from events and conditions outside the nuclear power unit. If an mHTGR design proposes using a direct power cycle in which However, dynamic effects associated with postulated pipe one or more very high-speed, very high-energy gas turbines are Appendix C to DG-1330, Page C-2

APPENDIX C. MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGN CRITERIA I. Overall Requirements Criterion mHTGR-DC Title and Content NRC Rationale for Adaptions to GDC ruptures in nuclear power units may be excluded from the located inside the reactor helium pressure boundary. The presence design basis when analyses reviewed and approved by the of one or more very high-energy turbines inside the primary Commission demonstrate that the probability of fluid system helium pressure boundary creates the potential that a catastrophic piping rupture is extremely low under conditions consistent dynamic failure of the gas turbine (e.g., at power) could result in with the design basis for the piping. the consequential catastrophic failure of the primary system pressure boundary caused by the failure of rotating turbine components. To account for the possibility of an mHTGR design that locates high-energy gas turbines inside the reactor helium pressure boundary, the mHTGR-DC language in the area of prevention, protection, and mitigation of turbine dynamic failure is strengthened to support such a power conversion system design approach.

5 Sharing of structures, systems, and components.

Same as GDC II. Multiple Barriers Criterion mHTGR-DC Title and Content NRC Rationale for Adaptions to GDC 10 Reactor design. The concept of specified acceptable fuel design limits, which The reactor system and associated heat removal, control, and prevent additional fuel failures during anticipated operational protection systems shall be designed with appropriate margin to occurrences (AOOs), has been replaced with that of the specified ensure that specified acceptable system radionuclide release acceptable system radionuclide release design limits (SARRDL),

design limits are not exceeded during any condition of normal which limits the amount of radionuclide inventory that is released operation, including the effects of anticipated operational by the fuel and surfaces within the helium coolant boundary under occurrences. normal and AOO conditions. The system refers to the components and internals of the mHTGR helium pressure boundary. Design features within the reactor system must ensure that the SARRDLS are not exceeded during normal operations and AOOs.

Appendix C to DG-1330, Page C-3

APPENDIX C. MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGN CRITERIA II. Multiple Barriers Criterion mHTGR-DC Title and Content NRC Rationale for Adaptions to GDC The tristructural isotropic (TRISO) fuel used in the mHTGR design is the primary fission product barrier and is expected to have a very low incremental fission product release during AOOs.

As noted in NUREG-1338 (Ref. 3) and in the NRC staffs feedback on the Next Generation Nuclear Plant (NGNP) project white paper, Next-Generation Nuclear Plant - Assessment of Key Licensing Issues (Ref. 11) the TRISO fuel fission product transport and retention behavior under all expected operating conditions is the key to meeting dose limits, as a different approach to defense in depth is employed in an mHTGR. The SARRDL concept allows for some small increase in circulating radionuclide inventory during an AOO. To ensure the SARRDL is not violated during an AOO, a normal operation radionuclide inventory limit must also be established (i.e., appropriate margin). The radionuclide activity circulating within the helium coolant boundary is continuously monitored such that the normal operation limits and SARRDLs are not exceeded.

The SARRDLs will be established so that the most limiting license-basis event does not exceed the siting regulatory dose limits criteria at the exclusion area boundary (EAB) and low-population zone (LPZ), and also so that the 10 CFR 20.1301 annualized dose limits to the public are not exceeded at the EAB for normal operation and AOOs.

The NRC has not approved the concept of replacing specified acceptable fuel design limits with SARRDLs. The concept of the TRISO fuel being the primary fission product barrier is intertwined with the concept of a functional containment for mHTGR technologies. See the rationale for mHTGR-DC 16 for further information on the Commissions current position.

Appendix C to DG-1330, Page C-4

APPENDIX C. MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGN CRITERIA II. Multiple Barriers Criterion mHTGR-DC Title and Content NRC Rationale for Adaptions to GDC The word coolant has been replaced with heat removal, as helium coolant inventory control for normal operation and AOOs is not necessary to meet the SARRDLs, due to the reactor system design.

11 Reactor inherent protection. The wording has been changed to broaden the applicability from Same as ARDC coolant systems to additional factors (including structures or The reactor core and associated systems that contribute to other fluids) that may contribute to reactivity feedback. These reactivity feedback shall be designed so that, in the power systems are to be designed to compensate for rapid reactivity operating range, the net effect of the prompt inherent nuclear increase.

feedback characteristics tends to compensate for a rapid increase in reactivity.

12 Suppression of reactor power oscillations. Helium in the mHTGR does not affect reactor core susceptibility to The reactor core and associated control and protection systems coolant-induced power oscillations; therefore, a separate mHTGR-shall be designed to ensure that power oscillations that can specific DC is appropriate. The word coolant was deleted and the result in conditions exceeding specified acceptable system specified acceptable fuel design limits were replaced by radionuclide release design limits are not possible or can be SARRDLs. The discussion on the SARRDL is given in reliably and readily detected and suppressed. mHTGR-DC 10.

13 Instrumentation and control. Reactor coolant pressure boundary has been relabeled as reactor Instrumentation shall be provided to monitor variables and helium pressure boundary to conform to standard terms used for systems over their anticipated ranges for normal operation, for mHTGRs.

anticipated operational occurrences, and for accident conditions, as appropriate, to ensure adequate safety, including The criterion has been modified to reflect the use of the mHTGR those variables and systems that can affect the fission process functional containment. See the mHTGR-DC 16 rationale.

and the integrity of the reactor core, reactor helium pressure boundary, and functional containment. Appropriate controls shall be provided to maintain these variables and systems within prescribed operating ranges.

14 Reactor helium pressure boundary. Reactor coolant pressure boundary has been relabeled as reactor The reactor helium pressure boundary shall be designed, helium pressure boundary to conform to standard terms used for fabricated, erected, and tested so as to have an extremely low mHTGRs.

probability of abnormal leakage, of rapidly propagating failure, Appendix C to DG-1330, Page C-5

APPENDIX C. MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGN CRITERIA II. Multiple Barriers Criterion mHTGR-DC Title and Content NRC Rationale for Adaptions to GDC of gross rupture, and of unacceptable ingress of moisture, air, The mHTGR-DC 14 addresses the need to consider leakage of secondary coolant, or other fluids. contaminants into the helium used to transport heat from the reactor to the heat exchangers for power production, residual heat removal, and process heat. The phrase reactor helium pressure boundary encompasses the entire volume containing helium used to cool the reactor, not just the volume within the reactor vessel.

For consistency, a specific requirement is appended to mHTGR-DC 30 for a means of detecting ingress of moisture, air, secondary coolant, or other fluids. Although other fluids could be interpreted as including water and steam, for emphasis, the word moisture is included in the list of contaminants in both mHTGR-DC 14 and mHTGR-DC 30.

15 Reactor helium pressure boundary system design. Reactor coolant system has been relabeled as reactor helium All systems that are part of the reactor helium pressure pressure boundary to conform to standard terms used for boundary, such as the reactor system, vessel system, and heat mHTGRs.

removal systems, and the associated auxiliary, control, and protection systems, shall be designed with sufficient margin to ensure that the design conditions of the reactor helium pressure boundary are not exceeded during any condition of normal operation, including anticipated operational occurrences.

16 Containment design. The term functional containment is applicable to advanced A reactor functional containment, consisting of multiple non-LWRs without a pressure retaining containment structure.

barriers internal and/or external to the reactor and its cooling A functional containment can be defined as a barrier, or set of system, shall be provided to control the release of radioactivity barriers taken together, that effectively limit the physical transport to the environment and to ensure that the functional and release of radionuclides to the environment across a full range containment design conditions important to safety are not of normal operating conditions, AOOs, and accident conditions.

exceeded for as long as postulated accident conditions require.

Functional containment is relied upon to ensure that dose at the site boundary as a consequence of postulated accidents meets regulatory limits. Traditional containment structures also provide the reactor and SSCs important to safety inside the containment Appendix C to DG-1330, Page C-6

APPENDIX C. MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGN CRITERIA II. Multiple Barriers Criterion mHTGR-DC Title and Content NRC Rationale for Adaptions to GDC structure protection against accidents related to external hazards (e.g., turbine missiles, flooding, aircraft).

The mHTGR functional containment safety design objective is to meet 10 CFR 50.34, 52.79, 52.137, or 52.157 offsite dose requirements at the plants exclusion area boundary (EAB) with margins.

The NRC staff has brought the issue of functional containment to the Commission, and the Commission has found it generally acceptable, as indicated in the staff requirements memoranda (SRM) to SECY-93-092 (Ref. 8) and SECY-03-0047 (Ref. 9). In the SRM to SECY-03-0047 (Ref. 10), the Commission instructed the staff to develop performance requirements and criteria working closely with industry experts (e.g., designers, EPRI, etc.)

and other stakeholders regarding options in this area, taking into account such features as core, fuel, and cooling systems design, and directed the staff to submit options and recommendations to the Commission for a policy decision.

The NRC staff also provided feedback to the DOE on this issue as part of the NGNP project. In the NRC staffs Summary Feedback on Four Licensing Issues NGNP (Ref. 11), the area on functional containment and fuel development and qualification noted that approval of the proposed approach to functional containment for the mHTGR concept, with its emphasis on passive safety features and radionuclide retention within the fuel over a broad spectrum of off-normal conditions, would necessitate that the required fuel particle performance capabilities be demonstrated with a high degree of certainty.

GDC 38, 39, 40, 41, 42, 43, 50, 51, 52, 53, 54, 55, 56, and 57 are not applicable to the mHTGR design, since they address design Appendix C to DG-1330, Page C-7

APPENDIX C. MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGN CRITERIA II. Multiple Barriers Criterion mHTGR-DC Title and Content NRC Rationale for Adaptions to GDC criteria for pressure-retaining containments in the traditional LWR sense. Requirements for the performance of the mHTGR reactor building are addressed by new Criterion 71 (design basis) and Criterion 72 (provisions for periodic testing and inspection).

17 Electric power systems. A reliable power system is required for SSCs during postulated Electric power systems shall be provided to permit functioning accident conditions. Power systems shall be sufficient in capacity, of structures, systems, and components important to safety. The capability, and reliability to ensure vital safety functions are safety function for the systems shall be to provide sufficient maintained. The emphasis is placed on requiring reliability of capacity, capability, and reliability to ensure that (1) specified power sources rather than prescribing how such reliability can be acceptable system radionuclide release design limits and design attained. The reference to onsite vs. offsite electric power systems conditions of the reactor helium pressure boundary are not was deleted to provide for those reactor designs that do not depend exceeded as a result of anticipated operational occurrences and on offsite power for the functioning of SSCs important to safety.

(2) vital functions that rely on electric power are maintained in the event of postulated accidents. The text related to supplies, including batteries, and the onsite The onsite electric power systems shall have sufficient distribution system, was deleted to allow increased flexibility in independence, redundancy, and testability to perform their the design of offsite power systems for advanced reactor designs.

safety functions, assuming a single failure. However, such onsite systems are still expected to remain capable of performing assigned safety functions during accidents as a condition of requisite reliability. Reactor coolant pressure boundary has been relabeled as reactor helium pressure boundary to conform to standard terms used for mHTGRs.

The specified acceptable fuel design limit has been replaced with the SARRDL. The discussion on the change to SARRDL is given in mHTGR-DC 10.

The existing single switchyard allowance remains available under ARDC 17. If a particular advanced design requires the use of GDC single switchyard allowance wording, the designer should look to GDC 17 for guidance when developing PDC.

If electrical power is not required to permit the functioning of SSCs important to safety, the requirements in the mHTGR-DC are not applicable to the design. In this case, the functionality of SSCs Appendix C to DG-1330, Page C-8

APPENDIX C. MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGN CRITERIA II. Multiple Barriers Criterion mHTGR-DC Title and Content NRC Rationale for Adaptions to GDC important to safety must be fully evaluated and documented in the design bases.

18 Inspection and testing of electric power systems. GDC 18 is a design-independent companion criterion to GDC 17.

Same as ARDC Electric power systems important to safety shall be designed to Wording pertaining to additional system examples has been deleted permit appropriate periodic inspection and testing of important to allow increased flexibility associated with various designs.

areas and features, such as wiring, insulation, connections, and switchboards, to assess the continuity of the systems and the The text related to the nuclear power unit, offsite power system, condition of their components. The systems shall be designed and onsite power system was deleted to be consistent with with a capability to test periodically (1) the operability and mHTGR-DC 17.

functional performance of the components of the systems, such as onsite power sources, relays, switches, and buses, and (2) the operability of the systems as a whole and, under conditions as close to design as practical, the full operation sequence that brings the systems into operation, including operation of applicable portions of the protection system, and the transfer of power among systems.

19 Control room. The criterion was updated to remove specific emphasis on LOCAs, Same as ARDC which may be not appropriate for advanced designs such as the A control room shall be provided from which actions can be mHTGR.

taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident Reference to whole body, or its equivalent to any part of the conditions. Adequate radiation protection shall be provided to body has been updated to the current total effective dose permit access and occupancy of the control room under equivalent standard as defined in § 50.2.

accident conditions without personnel receiving radiation exposures in excess of 5 rem total effective dose equivalent as A control room habitability requirement beyond that associated defined in § 50.2 for the duration of the accident. with radiation protection has been added to address the concern that non-radionuclide accidents may also affect control room Adequate habitability measures shall be provided to permit access and occupancy.

access and occupancy of the control room during normal operations and under accident conditions. Equipment at The last paragraph of the GDC has been eliminated for the appropriate locations outside the control room shall be provided mHTGR-DC because it is not applicable to future applicants.

Appendix C to DG-1330, Page C-9

APPENDIX C. MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGN CRITERIA II. Multiple Barriers Criterion mHTGR-DC Title and Content NRC Rationale for Adaptions to GDC (1) with a design capability for prompt hot shutdown of the reactor, including necessary instrumentation and controls to maintain the unit in a safe condition during hot shutdown, and (2) with a potential capability for subsequent cold shutdown of the reactor through the use of suitable procedures.

III. Reactivity Control Criterion mHTGR-DC Title and Content NRC Rationale for Adaptions to GDC 20 Protection system functions. Specified acceptable fuel design limits has been replaced with The protection system shall be designed (1) to initiate SARRDLs. The concept of using SARRDLs is discussed for automatically the operation of appropriate systems, including GDC 10. The quantitative value of the SARRDL will be design the reactivity control systems, to ensure that the specified specific. The protection aspect of automatic operation, to protect acceptable system radionuclide release design limits is not normal operation and AOO limits, to sense accident conditions, and exceeded as a result of anticipated operational occurrences and to initiate mitigating equipment has been preserved.

(2) to sense accident conditions and to initiate the operation of systems and components important to safety.

21 Protection system reliability and testability.

Same as GDC 22 Protection system independence.

Same as GDC 23 Protection system failure modes.

Same as GDC 24 Separation of protection and control systems.

Same as GDC Appendix C to DG-1330, Page C-10

APPENDIX C. MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGN CRITERIA III. Reactivity Control Criterion mHTGR-DC Title and Content NRC Rationale for Adaptions to GDC 25 Protection system requirements for reactivity control Specified acceptable fuel design limits is replaced with malfunctions. SARRDLs. The concept of using SARRDLs is discussed for The protection system shall be designed to ensure that specified GDC 10.

acceptable system radionuclide release design limits are not exceeded during any anticipated operational occurrence, accounting for a single malfunction of the reactivity control systems.

26 Reactivity control systems. Recent licensing activity associated with the application of GDC 26 Same as ARDC and GDC 27 to new reactor designs Response to Gap Analysis Reactivity control systems shall include the following Summary Report for Reactor System Issues, (Ref. 26) and capabilities: Response to NuScale Gap Analysis Summary Report for Reactivity Control Systems, Addressing Gap 11, General Design (1) A means of shutting down the reactor shall be provided to Criteria 26, (Ref. 27), revealed that additional clarity could be ensure that, under conditions of normal operation, provided in the area of reactivity control requirements. ARDC 26 including anticipated operational occurrences, and with combines the scope of GDC 26 and GDC 27. The development of appropriate margin for malfunctions, design limits for ARDC 26 is informed by the proposed General Design Criteria of fission product barriers are not exceeded. 1965, AEC-R 2/49 and November 5, 1967 (32 FR 10216) (Ref.

28); the current GDC 26 and 27; the definition of safety-related (2) A means of shutting down the reactor and maintaining a SSC in 10 CFR 50.2; and SECY-94-084, Policy and Technical safe shutdown under design-basis event conditions, with Issues Associated with the Regulatory Treatment of Non-Safety appropriate margin for malfunctions, shall be provided. A Systems in Passive Plant Designs (Ref. 29); and the prior second means of reactivity control shall be provided that is application of reactivity control requirements.

independent, diverse, and capable of achieving and maintaining safe shutdown under design-basis event Current GDC 26, first sentence, states that two reactivity control conditions. systems of different design principles shall be provided. In addition, the NRC has not licensed a power reactor that did not (3) A system for holding the reactor subcritical under cold provide two independent means of shutting down the reactor.

conditions shall be provided.

(1) Current GDC 26, second sentence, states that one of the reactivity control systems shall use control rods and shall be capable of reliably controlling reactivity changes to ensure that, under conditions of normal operation, including AOOs, and with Appendix C to DG-1330, Page C-11

APPENDIX C. MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGN CRITERIA III. Reactivity Control Criterion mHTGR-DC Title and Content NRC Rationale for Adaptions to GDC appropriate margin for malfunctions such as stuck rods, specified acceptable fuel design limits are not exceeded. The staff recognizes that specifying control rods may not be suitable for advanced reactors. Additionally, reliably controlling reactivity, as required by GDC 26, has been interpreted as ensuring the control rods are capable of rapidly (i.e., within a few seconds) shutting down the reactor (Ref. 27).

The staff changed control rods to means in recognition that advanced reactor designs may not rely on control rods to rapidly shut down the reactor (e.g., alternative system designs or inherent feedback mechanisms may be relied upon to perform this function).

Additionally, specified acceptable fuel design limits is replaced with design limits for fission product barriers to be consistent with the AOO acceptance criteria. ARDC 10 and ARDC 15 provide the appropriate design limits for the fuel and reactor coolant boundary, respectively. A non-LWR may not necessarily shut down rapidly (within seconds) but the shutdown should occur in a time frame such that the fission product barrier design limits are not exceeded. In regards to safety class, the capability to shut down the reactor is identified as a function performed by safety-related SSCs in the 10 CFR 50.2 definition of safety-related SSCs.

(2) Current GDC 27 states that the reactivity control systems shall be designed to have a combined capability of reliably controlling reactivity changes to assure that, under postulated accident conditions and with appropriate margin for stuck rods, the capability to cool the core is maintained. Reliably controlling reactivity, as required by GDC 27, requires that the reactor achieve and maintain safe, stable conditions, including subcriticality, using only safety related equipment with margin for stuck rods (Ref. 26).

The first sentence of ARDC 26 (2) refers to the safety-related Appendix C to DG-1330, Page C-12

APPENDIX C. MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGN CRITERIA III. Reactivity Control Criterion mHTGR-DC Title and Content NRC Rationale for Adaptions to GDC means (systems and/or mechanisms) to achieve and maintain safe shutdown. Maintain safe shutdown indicates subcriticality in the long term or an equilibrium condition naturally achieved by the design.

The staff changed reactivity control systems to means in recognition that advanced reactor designs may rely on a system, inherent feedback mechanism, or some combination thereof to shut down the reactor and maintain a safe shutdown under design-basis event conditions. SECY-94-084, Policy and Technical Issues Associated with the Regulatory Treatment of Non-Safety Systems in Passive Plant Designs (Ref. 29), describes the characteristics of a safe shutdown condition as reactor subcriticality, decay heat removal, and radioactive materials containment. The staff replaced postulated accident conditions with design-basis event conditions, to emphasize that plants are required to maintain a safe shutdown following AOOs as well as postulated accidents.

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 serves as a backup to the safety-related 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).

Independent indicates no shared systems or components with the safety-related means and diverse indicates a different design than the safety-related means. The purpose of an independent and diverse means of controlling reactivity is to preclude a potential common cause failure affecting both means of reactivity control, which would lead to the inability to shut down the reactor. The second means of reactivity control does not have to demonstrate that design limits for fission product barriers are met.

Appendix C to DG-1330, Page C-13

APPENDIX C. MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGN CRITERIA III. Reactivity Control Criterion mHTGR-DC Title and Content NRC Rationale for Adaptions to GDC Additionally, the current GDC 26, third sentence, states that the second reactivity control system shall be capable of reliably controlling the rate of changes resulting from planned, normal power changes (including xenon burnout) to assure acceptable fuel design limits are not exceeded. Staff has identified this as an operational requirement that is not necessary to ensure reactor safety provided a design complies with ARDC 26(1). Therefore, this sentence is not retained in ARDC 26.

27 Combined reactivity control systems capability.

Same as ARDC DELETEDInformation incorporated into ARDC 26 28 Reactivity limits. Reactor coolant pressure boundary has been relabeled as reactor The reactor core, including the reactivity control systems, shall helium pressure boundary to conform to standard terms used for be designed with appropriate limits on the potential amount and mHTGRs.

rate of reactivity increase to ensure that the effects of postulated reactivity accidents can neither (1) result in damage to the The list of postulated reactivity accidents has been deleted. Each reactor helium pressure boundary greater than limited local design will have to determine its postulated reactivity accidents yielding, nor (2) sufficiently disturb the core, its support based on the specific design and associated risk evaluation.

structures, or other reactor vessel internals to impair significantly the capability to cool the core.

29 Protection against anticipated operational occurrences.

Same as GDC Appendix C to DG-1330, Page C-14

APPENDIX C. MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGN CRITERIA IV. Fluid Systems Criterion mHTGR-DC Title and Content NRC Rationale for Adaptions to GDC 30 Quality of reactor helium pressure boundary. Reactor coolant pressure boundary has been relabeled as reactor Components that are part of the reactor helium pressure helium pressure boundary to conform to standard terms used for boundary shall be designed, fabricated, erected, and tested to mHTGRs.

the highest quality standards practical. Means shall be provided for detecting and, to the extent practical, identifying the The mHTGR-DC 14 addresses the need to consider leakage of location of the source of reactor helium leakage. Means shall be contaminants into the helium used to transport heat from the reactor provided for detecting ingress of moisture, air, secondary to the heat exchangers for power production, residual heat removal, coolant, or other fluids to within the reactor helium pressure and process heat. The phrase reactor helium pressure boundary boundary. encompasses the entire volume containing helium used to cool the reactor, not just the volume within the reactor vessel. For consistency, a specific requirement is appended to mHTGR-DC 30 for a means of detecting ingress of moisture, air, secondary coolant, or other fluids. Although other fluids could be interpreted as including water and steam, for emphasis, the word moisture is included in the list of contaminants in both mHTGR-DC 14 and mHTGR-DC 30.

31 Fracture prevention of reactor helium pressure boundary. Reactor coolant pressure boundary has been relabeled as reactor The reactor helium pressure boundary shall be designed with helium pressure boundary to conform to standard terms used for sufficient margin to ensure that, when stressed under operating, mHTGRs.

maintenance, testing, and postulated accident conditions, (1) the boundary behaves in a nonbrittle manner and (2) the Specific examples are added to the mHTGR-DC to account for the probability of rapidly propagating fracture is minimized. The high design and operating temperatures and unique potential design shall reflect consideration of service temperatures, coolants.

service degradation of material properties, creep, fatigue, stress rupture, and other conditions of the boundary material under operating, maintenance, testing, and postulated accident conditions and the uncertainties in determining (1) material properties, (2) the effects of irradiation and coolant chemistry on material properties, (3) residual, steady-state, and transient stresses, and (4) size of flaws.

Appendix C to DG-1330, Page C-15

APPENDIX C. MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGN CRITERIA IV. Fluid Systems Criterion mHTGR-DC Title and Content NRC Rationale for Adaptions to GDC 32 Inspection of reactor helium pressure boundary. Reactor coolant pressure boundary has been relabeled as reactor Components that are part of the reactor helium pressure helium pressure boundary to conform to standard terms used for boundary shall be designed to permit (1) periodic inspection mHTGRs.

and functional testing of important areas and features to assess their structural and leaktight integrity, and (2) an appropriate The staff modified the LWR GDC by replacing the term reactor material surveillance program for the reactor vessel. pressure vessel with reactor vessel, which the staff believes is a more generically applicable term.

A non-leaktight system may be acceptable for some designs provided that (1) the system leakage does not impact safety functions under all conditions, and (2) leakage is consistent with SARRDL.

33 Reactor coolant makeup. The mHTGR does not require reactor coolant inventory Not applicable to mHTGR. maintenance for small leaks to meet the SARRDLs, which replaces the concept of the specified acceptable fuel design limits, as discussed in GDC 10. Therefore, ARDC 33 is not applicable to the mHTGR design.

34 Passive residual heat removal.

A passive system to remove residual heat shall be provided. For The word passive was added, based on the definition of a normal operations and anticipated operational occurrences, the mHTGR. In definitions Section 3.1 of the DOE report titled system safety function shall be to transfer fission product decay Guidance for Developing Principal Design Criteria for Advanced heat and other residual heat from the reactor core to an ultimate (Non-Light-Water) Reactors (Ref. 17), the mHTGR design has a heat sink at a rate such that specified acceptable system low power density and hence residual heat is removed by a passive radionuclide release design limits and the design conditions of system.

the reactor helium pressure boundary are not exceeded.

Ultimate heat sink has been added to explain that, if During postulated accidents, the system safety function shall mHTGR-DC 44 is deemed not applicable to the design, the residual provide effective core cooling. heat removal system is then required to provide the heat removal path to the ultimate heat sink.

Suitable redundancy in components and features and suitable interconnections, leak detection, and isolation capabilities shall Appendix C to DG-1330, Page C-16

APPENDIX C. MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGN CRITERIA IV. Fluid Systems Criterion mHTGR-DC Title and Content NRC Rationale for Adaptions to GDC be provided to ensure the system safety function can be Reactor coolant pressure boundary has been relabeled as reactor accomplished, assuming a single failure. helium pressure boundary to conform to standard terms used for mHTGRs.

The SARRDL replaces the ARDC specified acceptable fuel design limits as described in the rationale to mHTGR-DC 10.

The mHTGR-DC 34 incorporates the postulated accident residual heat removal requirements contained in GDC 35.

Effective core cooling under postulated accident conditions is defined as maintaining fuel temperature limits below design values to help ensure the siting regulatory dose limits criteria at the exclusion area boundary (EAB) and low-population zone (LPZ) are not exceeded and a geometry is preserved which supports residual heat removal.

The GDC reference to electric power was removed. Refer to the rationale for ARDC 17 on electric power systems.

35 Emergency core cooling. In the mHTGR design the power density and large length to Not applicable to mHTGR. diameter ratio are such that maintaining the helium coolant inventory is not necessary to maintain effective core cooling.

Postulated accident heat removal is accomplished by the residual heat removal system described in mHTGR DC 34.

Appendix C to DG-1330, Page C-17

APPENDIX C. MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGN CRITERIA IV. Fluid Systems Criterion mHTGR-DC Title and Content NRC Rationale for Adaptions to GDC 36 Inspection of passive residual heat removal system. The word passive was added, based on the definition of a mHTGR. In definitions Section 3.1 of DOE report titled Guidance The passive residual heat removal shall be designed to permit for Developing Principal Design Criteria for Advanced (Non-appropriate periodic inspection of important components to Light-Water) Reactors (Ref. 17), the mHTGR design has a low ensure the integrity and capability of the system. power density and hence residual heat is removed by a passive system.

The GDC 36 system is renamed and revised to provide for inspection of the residual heat removal systems as required for mHTGR-DC 34.

The list of examples was deleted, as they apply to LWR designs and each specific design will have different important components associated with residual heat removal.

37 Testing of passive residual heat removal system. Criterion 37 has been renamed and revised for testing the passive residual heat removal system required by mHTGR-DC 34.

The passive residual heat removal system shall be designed to permit appropriate periodic functional testing to ensure (1) the Section 2.3.4 of INL/EXT-10-17997, Mechanistic Source Terms structural and leaktight integrity of its components, (2) the White Paper, (Ref. 33) notes that the passive reactor cavity operability and performance of the system components, and cooling system (RCCS) (using either air or water as heat transfer (3) the operability of the system as a whole and, under fluid) contributes to the mHTGR safety basis and is subject to conditions as close to design as practical, the performance of component integrity testing. However, Section 6.1 of the full operational sequence that brings the system into INL/EXT-11-22708, Modular HTGR Safety Basis and operation, including operation of associated systems and Approach, (Ref. 34), indicates that RCCS performance does not interfaces with an ultimate heat sink and the transition from the require leaktight conditions. For an RCCS which is an open active normal operation mode to the passive operation mode system, the normal and expected loss of RCCS coolant through relied upon during postulated accidents, including the operation the exhaust structure would not be considered leakage. Abnormal of applicable portions of the protection system and the leakage of RCCS coolant to locations other than the exhaust operation of the associated structural and equipment cooling structure may be acceptable provided that (1) the RCCS leakage water system. does not impact safety functions under all conditions, and (2) functional containment is not impacted by RCCS leakage.

Appendix C to DG-1330, Page C-18

APPENDIX C. MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGN CRITERIA IV. Fluid Systems Criterion mHTGR-DC Title and Content NRC Rationale for Adaptions to GDC Some mHTGR RCCS designs will provide continuous passive operation without need for a requirement to test the operation sequence that brings the system into operation; if applicable is included to recognize this contingency.

Reference to the operation of applicable portions of the protection system, structural and equipment cooling water systems, and power transfers is considered part of the more general associated systems for operability testing of the system as a whole.

The criterion was modified to reflect the passive nature of the mHTGR RCCS and the need to verify the ability to transition the RCCS from active mode (if present) to passive mode during postulated accidents.

38 Containment heat removal. This criterion is not applicable to the mHTGR. The mHTGR Not applicable to mHTGR. designs do not have a pressure retaining reactor containment structure but instead rely on a multibarrier functional containment configuration to control the release of radionuclides. See the mHTGR DC 16 rationale.

39 Inspection of containment heat removal system. This criterion is not applicable to the mHTGR. The mHTGR Not applicable to mHTGR. designs do not have a pressure retaining reactor containment structure but instead rely on a multibarrier functional containment configuration to control the release of radionuclides. See the mHTGR-DC 16 rationale.

40 Testing of containment heat removal system. This criterion is not applicable to the mHTGR. The mHTGR Not applicable to mHTGR. designs do not have a pressure retaining reactor containment structure but instead rely on a multibarrier functional containment configuration to control the release of radionuclides. See the mHTGR-DC 16 rationale.

41 Containment atmosphere cleanup. This criterion is not applicable to the mHTGR. The mHTGR Not applicable to mHTGR. designs do not have a pressure retaining reactor containment Appendix C to DG-1330, Page C-19

APPENDIX C. MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGN CRITERIA IV. Fluid Systems Criterion mHTGR-DC Title and Content NRC Rationale for Adaptions to GDC structure but instead rely on a multibarrier functional containment configuration to control the release of radionuclides. See the mHTGR-DC 16 rationale.

42 Inspection of containment atmosphere cleanup systems. This criterion is not applicable to the mHTGR. The mHTGR Not applicable to mHTGR. designs do not have a pressure retaining reactor containment structure but instead rely on a multibarrier functional containment configuration to control the release of radionuclides. See the mHTGR-DC 16 rationale.

43 Testing of containment atmosphere cleanup systems. This criterion is not applicable to the mHTGR. The mHTGR Not applicable to mHTGR. designs do not have a pressure retaining reactor containment structure but instead rely on a multibarrier functional containment configuration to control the release of radionuclides. See the mHTGR-DC 16 rationale.

44 Structural and equipment cooling. This mHTGR-DC accounts for advanced reactor design system In addition to the heat rejection capability of the passive differences to include cooling requirements for SSCs important to residual heat removal system, systems to transfer heat from safety, if applicable; this mHTGR-DC does not address the residual structures, systems, and components important to safety to an heat removal system required under ARDC 34.

ultimate heat sink shall be provided, as necessary, to transfer the combined heat load of these structures, systems, and The staff inserted passive based on the system design for residual components under normal operating and accident conditions. heat removal. If a specific mHTGR design can demonstrate that the reactor cavity cooling system (RCCS) provides indefinite core Suitable redundancy in components and features and suitable cooling capability, then structural and equipment cooling systems interconnections leak detection, and isolation capabilities shall would not be needed.

be provided to ensure that the system safety function can be accomplished, assuming a single failure. The GDC reference to electric power was removed. Refer to the rationale for ARDC 17 on electric power systems.

45 Inspection of structural and equipment cooling systems. This renamed mHTGR-DC accounts for advanced reactor system Same as ARDC design differences to include possible cooling requirements for The structural and equipment cooling systems shall be designed SSCs important to safety.

to permit appropriate periodic inspection of important Appendix C to DG-1330, Page C-20

APPENDIX C. MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGN CRITERIA IV. Fluid Systems Criterion mHTGR-DC Title and Content NRC Rationale for Adaptions to GDC components, such as heat exchangers and piping, to assure the integrity and capability of the systems.

46 Testing of structural and equipment cooling systems. This renamed mHTGR-DC accounts for advanced reactor system Same as ARDC design differences to include possible cooling requirements for The structural and equipment cooling systems shall be designed SSCs important to safety. Specific mention of pressure testing to permit appropriate periodic functional testing to assure (1) has been removed yet remains a potential requirement should it be the structural and leaktight integrity of their components, (2) necessary as a component of appropriate periodic functional the operability and the performance of the system components, testing... of cooling systems. A non-leaktight system may be and (3) the operability of the systems as a whole and, under acceptable for some designs provided that (1) the system leakage conditions as close to design as practical, the performance of does not impact safety functions under all conditions, and (2) the full operational sequences that bring the systems into defense in depth is not impacted by system leakage.

operation for reactor shutdown and postulated accidents, including operation of associated systems. Active has been deleted in item (2) because appropriate operability and performance tests of system components are required regardless of their active or passive nature. The LOCA reference has been removed to provide for any postulated accident that might affect subject SSCs.

The GDC reference to electric power was removed. Refer to the rationale for ARDC 17 regarding electric power systems.

V. Reactor Containment Criterion mHTGR-DC Title and Content NRC Rationale for Adaptions to GDC 50 Containment design basis. This criterion is not applicable to the mHTGR. The mHTGR Not applicable to mHTGR. designs do not have a pressure retaining reactor containment structure but instead rely on a multibarrier functional containment configuration to control the release of radionuclides. See the mHTGR-DC 16 rationale.

Appendix C to DG-1330, Page C-21

APPENDIX C. MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGN CRITERIA V. Reactor Containment Criterion mHTGR-DC Title and Content NRC Rationale for Adaptions to GDC 51 Fracture prevention of containment pressure boundary. This criterion is not applicable to the mHTGR. The mHTGR Not applicable to mHTGR. designs do not have a pressure retaining reactor containment structure but instead rely on a multibarrier functional containment configuration to control the release of radionuclides. See the mHTGR-DC 16 rationale.

52 Capability for containment leakage rate testing. This criterion is not applicable to the mHTGR. The mHTGR Not applicable to mHTGR. designs do not have a pressure retaining reactor containment structure but instead rely on a multibarrier functional containment configuration to control the release of radionuclides. See the mHTGR-DC 16 rationale.

53 Provisions for containment testing and inspection. This criterion is not applicable to the mHTGR. The mHTGR Not applicable to mHTGR. designs do not have a pressure retaining reactor containment structure but instead rely on a multibarrier functional containment configuration to control the release of radionuclides. See the mHTGR-DC 16 rationale.

54 Piping systems penetrating containment. This criterion is not applicable to the mHTGR. The mHTGR Not applicable to mHTGR. designs do not have a pressure retaining reactor containment structure but instead rely on a multibarrier functional containment configuration to control the release of radionuclides. See the mHTGR-DC 16 rationale.

55 Reactor coolant boundary penetrating containment. This criterion is not applicable to the mHTGR. The mHTGR Not applicable to mHTGR. designs do not have a pressure retaining reactor containment structure but instead rely on a multibarrier functional containment configuration to control the release of radionuclides. See the mHTGR-DC 16 rationale.

56 Primary Containment isolation. This criterion is not applicable to the mHTGR. The mHTGR Not applicable to mHTGR. designs do not have a pressure retaining reactor containment structure but instead rely on a multibarrier functional containment configuration to control the release of radionuclides. See the mHTGR-DC 16 rationale.

Appendix C to DG-1330, Page C-22

APPENDIX C. MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGN CRITERIA V. Reactor Containment Criterion mHTGR-DC Title and Content NRC Rationale for Adaptions to GDC 57 Closed system isolation valves. This criterion is not applicable to the mHTGR. The mHTGR Not applicable to mHTGR. designs do not have a pressure retaining reactor containment structure but instead rely on a multibarrier functional containment configuration to control the release of radionuclides. See the mHTGR-DC 16 rationale.

VI. Fuel and Reactivity Control Criterion mHTGR-DC Title and Content NRC Rationale for Adaptions to GDC 60 Control of releases of radioactive materials to the environment.

Same as GDC 61 Fuel storage and handling and radioactivity control. The underlying concept of establishing functional requirements for Same as ARDC radioactivity control in fuel storage and fuel handling systems is The fuel storage and handling, radioactive waste, and other independent of the design of non-LWR advanced reactors.

systems which may contain radioactivity shall be designed to However, some advanced designs may use dry fuel storage that assure adequate safety under normal and postulated accident incorporates cooling jackets that can be liquid-cooled or air-cooled conditions. These systems shall be designed (1) with a capability to remove heat. This modification to this GDC allows for both to permit appropriate periodic inspection and testing of liquid and air-cooling of the dry fuel storage containers.

components important to safety, (2) with suitable shielding for radiation protection, (3) with appropriate containment, confinement, and filtering systems, (4) with a residual heat removal capability having reliability and testability that reflects the importance to safety of decay heat and other residual heat removal, and (5) to prevent significant reduction in fuel storage cooling under accident conditions.

62 Prevention of criticality in fuel storage and handling.

Same as GDC 63 Monitoring fuel and waste storage.

Same as GDC Appendix C to DG-1330, Page C-23

APPENDIX C. MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGN CRITERIA VI. Fuel and Reactivity Control Criterion mHTGR-DC Title and Content NRC Rationale for Adaptions to GDC 64 Monitoring radioactivity releases. The underlying concept of monitoring radioactivity releases from Means shall be provided for monitoring the reactor building the mHTGR particle fuel to the reactor building, effluent discharge atmosphere, effluent discharge paths, and plant environs for paths, and plant environs applies. High radioactivity in the reactor radioactivity that may be released from normal operations, building provides input to the plant protection system. In addition, including anticipated operational occurrences, and from the reactor building atmosphere is monitored for personnel postulated accidents. protection. Recirculation of loss-of-coolant fluids (i.e., water) does not apply to the mHTGR.

The descriptions of the associated atmospheres and spaces that are required to be monitored are revised to reflect the mHTGRs different design configuration and functional containment arrangement.

VII. Additional mHTGR-DC Criterion mHTGR-DC Title and Content NRC Rationale for Adaptions to GDC 70 Reactor vessel and reactor system structural design basis. New mHTGR design-specific GDC are necessary to ensure that the The design of the reactor vessel and reactor system shall be such reactor vessel and reactor system (including the fuel, reflector, that their integrity is maintained during postulated accidents control rods, core barrel, and structural supports) integrity is (1) to ensure the geometry for passive removal of residual heat preserved for passive heat removal and for the insertion of neutron from the reactor core to the ultimate heat sink and (2) to permit absorbers.

sufficient insertion of the neutron absorbers to provide for reactor shutdown.

71 Reactor building design basis. The reactor building functions are to protect and maintain passive The design of the reactor building shall be such that, during cooling geometry and to provide a pathway for the release of postulated accidents, it structurally protects the geometry for helium from the building in the case of a line break in the reactor passive removal of residual heat from the reactor core to the helium pressure boundary. This newly established criterion ensures ultimate heat sink and provides a pathway for the release of that these safety functions are provided.

reactor helium from the building in the event of depressurization It is noted that the reactor building is not relied upon to meet the accidents. offsite dose requirements of 10 CFR 50.34 (10 CFR 52.79).

Appendix C to DG-1330, Page C-24

APPENDIX C. MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGN CRITERIA VII. Additional mHTGR-DC Criterion mHTGR-DC Title and Content NRC Rationale for Adaptions to GDC 72 Provisions for periodic reactor building inspection. This newly established criterion on periodic inspection and The reactor building shall be designed to permit (1) appropriate surveillance provides assurance that the reactor building will periodic inspection of all important structural areas and the perform its safety functions of protecting and maintaining the depressurization pathway, and (2) an appropriate surveillance configuration needed for passive cooling and providing a discharge program. pathway for helium depressurization events.

Appendix C to DG-1330, Page C-25