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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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 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.

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.

DG-1330, Page 2 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).

DG-1330, Page 3 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.

DG-1330, Page 4 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 5 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.

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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 6 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.

DG-1330, Page 7 Title 10 CFR 50.341 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).

DG-1330, Page 8 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.

DG-1330, Page 9 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.

DG-1330, Page 10 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.

DG-1330, Page 11 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.

DG-1330, Page 12 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)

TABLE 1: NON-LIGHT-WATER-REACTOR CROSSWALK DG-1330, Page 13 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 against natural phenomena.

Same as GDC Same as GDC Same as GDC 3

Fire protection.

Fire protection.

Modified for ARDC Same as ARDC Same as ARDC 4

Environmental and dynamic effects design bases.

Environmental and dynamic effects design bases.

Modified for ARDC Environmental and dynamic effects design bases.

Modified for SFR-DC Environmental and dynamic effects design bases.

Modified for mHTGR-DC 5

Sharing of structures, systems, and components.

Same as GDC Same as GDC Same as GDC 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.

Modified for ARDC Same as ARDC Same as ARDC 12 Suppression of reactor power oscillations.

Suppression of reactor power oscillations.

Modified for ARDC Same as ARDC Suppression of reactor power oscillations.

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

Modified for ARDC Instrumentation and control.

Modified for SFR-DC Instrumentation and control.

Modified for mHTGR-DC

TABLE 1: NON-LIGHT-WATER-REACTOR CROSSWALK DG-1330, Page 14 II.

Multiple Barriers Criterion Current GDC Title ARDC Title/Status SFR-DC Title/Status mHTGR-DC Title/Status 14 Reactor coolant pressure boundary.

Reactor coolant pressure boundary.

Modified for ARDC Primary coolant pressure boundary.

Modified for SFR-DC Reactor helium coolant pressure boundary.

Modified for mHTGR-DC 15 Reactor coolant system design.

Reactor coolant system design.

Modified for ARDC Primary Reactor coolant system design.

Modified for SFR-DC Reactor helium pressure boundary coolant system design.

Modified for mHTGR-DC 16 Containment design.

Same as GDC Containment design.

Modified for SFR-DC Containment design.

Modified for mHTGR-DC 17 Electric power systems.

Electric power systems.

Modified for ARDC Electric power systems.

Modified for SFR-DC Electric power systems.

Modified for mHTGR-DC 18 Inspection and testing of electric power systems.

Inspection and testing of electric power systems.

Modified for ARDC Same as ARDC Same as GDC 19 Control room.

Control room.

Modified for ARDC Control room.

Modified for SFR-DC Same as ARDC 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 and testability.

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

Same as GDC Same as GDC Same as GDC

TABLE 1: NON-LIGHT-WATER-REACTOR CROSSWALK DG-1330, Page 15 III.

Reactivity Control Criterion Current GDC Title ARDC Title/Status SFR-DC Title/Status mHTGR-DC Title/Status 23 Protection system failure modes.

Same as GDC Protection system failure modes.

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

Same as GDC Same as GDC Same as GDC 25 Protection system requirements for reactivity control malfunctions.

Protection system requirements for reactivity control malfunctions.

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

Modified for mHTGR-DC 26 Reactivity control system redundancy and capability.

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

Reactivity limits.

Modified for ARDC Reactivity limits.

Modified for SFR-DC Reactivity limits.

Modified for mHTGR-DC 29 Protection against anticipated operational occurrences.

Same as GDC Same as GDC Same as GDC

TABLE 1: NON-LIGHT-WATER-REACTOR CROSSWALK DG-1330, Page 16 IV.

Fluid Systems Criterion Current GDC Title ARDC Title/Status SFR-DC Title/Status mHTGR-DC Title/Status 30 Quality of reactor coolant pressure boundary.

Quality of reactor coolant pressure boundary.

Modified for ARDC Quality of reactor primary coolant pressure boundary.

Modified for SFR-DC Quality of reactor helium coolant pressure boundary.

Modified for mHTGR-DC 31 Fracture prevention of reactor coolant pressure boundary.

Fracture prevention of reactor coolant pressure boundary.

Modified for ARDC Fracture prevention of reactor primary coolant pressure boundary.

Modified for SFR-DC Fracture prevention of reactor helium coolant pressure boundary.

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

Inspection of reactor coolant pressure boundary.

Modified for ARDC Inspection of reactor primary coolant pressure boundary.

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

Modified for mHTGR-DC 33 Reactor coolant makeup.

Reactor coolant inventory maintenance.makeup Modified for ARDC Reactor Primary coolant inventory maintenance makeup.

Modified for SFR-DC Not applicable to mHTGR.

34 Residual heat removal.

Residual heat removal.

Modified for ARDC Residual heat removal.

Modified for SFR-DC Passive residual heat removal.

Modified for mHTGR-DC 35 Emergency core cooling.

Emergency core cooling.

Modified for ARDC Same as ARDC Not applicable to mHTGR.

36 Inspection of emergency core cooling system.

Inspection of emergency core cooling system.

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

Modified for mHTGR-DC 37 Testing of emergency core cooling system.

Testing of emergency core cooling system.

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

Modified for mHTGR-DC 38 Containment heat removal.

Containment heat removal.

Modified for ARDC Same as ARDC Not applicable to mHTGR.

TABLE 1: NON-LIGHT-WATER-REACTOR CROSSWALK DG-1330, Page 17 IV.

Fluid Systems Criterion Current GDC Title ARDC Title/Status SFR-DC Title/Status mHTGR-DC Title/Status 39 Inspection of containment heat removal system.

Inspection of containment heat removal system.

Modified for ARDC Same as ARDC Not applicable to mHTGR.

40 Testing of containment heat removal system.

Testing of containment heat removal system.

Modified for ARDC Same as ARDC Not applicable to mHTGR.

41 Containment atmosphere cleanup.

Containment atmosphere cleanup.

Modified for ARDC Same as ARDC Not applicable to mHTGR.

42 Inspection of containment atmosphere cleanup systems.

Same as GDC Same as GDC Not applicable to mHTGR.

43 Testing of containment atmosphere cleanup systems.

Testing of containment atmosphere cleanup systems.

Modified for ARDC Same as ARDC Not applicable to mHTGR.

44 Cooling water.

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

Inspection of structural and equipment cooling water systems.

Modified for ARDC Same as ARDC Same as ARDC 46 Testing of cooling water system.

Testing of structural and equipment cooling water systems.

Modified for ARDC Same as ARDC Same as ARDC

TABLE 1: NON-LIGHT-WATER-REACTOR CROSSWALK DG-1330, Page 18 V.

Reactor Containment Criterion Current GDC Title ARDC Title/Status SFR-DC Title/Status mHTGR-DC Title/Status 50 Containment design basis.

Containment design basis.

Modified for ARDC Same as ARDC Not applicable to mHTGR.

51 Fracture prevention of containment pressure boundary.

Fracture prevention of containment pressure boundary.

Modified for ARDC Same as ARDC Not applicable to mHTGR.

52 Capability for containment leakage rate testing.

Capability for containment leakage rate testing.

Modified for ARDC Same as ARDC Not applicable to mHTGR.

53 Provisions for containment testing and inspection.

Provisions for containment testing and inspection.

Modified for ARDC Same as ARDC Not applicable to mHTGR.

54 Piping systems penetrating containment.

Piping systems penetrating containment.

Modified for ARDC Piping systems penetrating containment.

Modified for SFR-DC Not applicable to mHTGR.

55 Reactor coolant pressure boundary penetrating containment.

Reactor coolant pressure boundary penetrating containment.

Modified for ARDC Reactor Primary coolant pressure boundary penetrating containment.

Modified for SFR-DC Not applicable to mHTGR.

56 Primary containment isolation.

Primary Containment isolation.

Modified for ARDC Same as ARDC Not applicable to mHTGR.

57 Closed system isolation valves.

Closed system isolation valves.

Modified for ARDC Closed system isolation valves.

Modified for SFR-DC Not applicable to mHTGR.

TABLE 1: NON-LIGHT-WATER-REACTOR CROSSWALK DG-1330, Page 19 VI.

Fuel and Radioactivity Control Criterion Current GDC Title ARDC Title/Status SFR-DC Title/Status mHTGR-DC Title/Status 60 Control of releases of radioactive materials to the environment.

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

Fuel storage and handling and radioactivity control.

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

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

Same as GDC Same as GDC Same as GDC 64 Monitoring radioactivity releases.

Monitoring radioactivity releases.

Modified for ARDC Monitoring radioactivity releases.

Modified for SFR-DC Monitoring radioactivity releases.

Modified for mHTGR-DC

TABLE 1: NON-LIGHT-WATER-REACTOR CROSSWALK DG-1330, Page 20 VII. Additional Technology-Specific Design Criteria Criterion Current GDC Title ARDC Title/Status SFR-DC Title/Status mHTGR-DC Title/Status 70 N/A N/A Intermediate coolant system.

Reactor vessel and reactor system structural design basis.

71 N/A N/A Primary coolant and cover gas purity control.

Reactor building design basis.

72 N/A N/A Sodium heating systems.

Provisions for periodic reactor building inspection.

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

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

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

N/A 76 N/A N/A Fracture prevention of the intermediate coolant boundary. N/A 77 N/A N/A Inspection of the intermediate coolant boundary.

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

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

N/A

DG-1330, Page 20 D. IMPLEMENTATION The purpose of this section is to provide information on how applicants and licensees2 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 voluntarily3 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 21 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 22 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 23 REFERENCES4

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 24

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 25

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=770896.

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/S03064549150058009.

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 2

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).

Appendix A to DG-1330, Page A-1 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.

ADVANCED REACTOR DESIGN CRITERIA Appendix A to DG-1330, Page A-2 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.

Structures, systems, and components important to safety shall be designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions. Noncombustible and fire-resistant materials shall be used wherever practical throughout the unit, particularly in 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.

The phrase containing examples where noncombustible and heat-resistant materials must be used has been broadened to apply to all advanced reactor designs.

Instead of and, the phrase locations with structures, systems, and components (SSCs) important to safety uses or, which is logically correct in this case.

4 Environmental and dynamic effects design bases.

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

Reference to pipe whip may not be applicable to designs that operate at low pressure.

APPENDIX A.

ADVANCED REACTOR DESIGN CRITERIA Appendix A to DG-1330, Page A-3 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 reactor core and associated systems that contribute to reactivity feedback shall be designed so that, in the power operating range, the net effect of the prompt inherent nuclear feedback characteristics tends to compensate for a rapid increase in reactivity.

The wording has been changed to broaden the applicability from coolant systems to additional factors (including structures or other fluids) that may contribute to reactivity feedback. These systems are to be designed to compensate for rapid reactivity increase.

12 Suppression of reactor power oscillations.

The reactor core; associated structures; and associated coolant, control, and protection systems shall be designed to ensure that 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.

The word structures was added because items such as reflectors, which could be considered either outside or not part of the reactor core, may affect susceptibility of the core to power oscillations.

13 Instrumentation and control.

Instrumentation shall be provided to monitor variables and systems over their anticipated ranges for normal operation, for anticipated operational occurrences, and for accident conditions, as appropriate to ensure adequate safety, including those variables and systems that can affect the fission process, the integrity of the reactor core, the reactor coolant boundary, and 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.

APPENDIX A.

ADVANCED REACTOR DESIGN CRITERIA Appendix A to DG-1330, Page A-4 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.

The reactor coolant boundary shall be designed, fabricated, erected, and tested so as to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture.

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.

15 Reactor coolant system design.

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

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.

16 Containment design.

Same as GDC For non-LWR technologies other than SFRs and mHTGRs, 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.

Electric power systems shall be provided to permit functioning of structures, systems, and components important to safety. The safety function for the systems shall be to provide sufficient capacity, capability, and reliability to ensure that (1) specified acceptable fuel design limits and design conditions of the reactor coolant boundary are not exceeded as a result of A reliable power system is required for SSCs during postulated accident conditions. Power systems shall be sufficient in capacity, capability, and reliability to ensure vital safety functions are maintained. The emphasis is placed on requiring reliability of power sources rather than prescribing how such reliability can be attained.

Reference to onsite vs. offsite electric power systems was deleted to

APPENDIX A.

ADVANCED REACTOR DESIGN CRITERIA Appendix A to DG-1330, Page A-5 II.

Multiple Barriers Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC anticipated operational occurrences and (2) vital functions that rely on electric power are maintained in the event of postulated accidents.

The onsite electric power systems shall have sufficient independence, redundancy, and testability to perform their safety functions, assuming a single failure.

provide for those reactor designs that do not depend on offsite power for the functioning of SSCs important to safety.

Text related to supplies, including batteries, and the onsite distribution system, was deleted to allow increased flexibility in the design of offsite power systems for advanced reactor designs.

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.

Electric power systems important to safety shall be designed to permit appropriate periodic inspection and testing of important areas and features, such as wiring, insulation, connections, and switchboards, to assess the continuity of the systems and the condition of their components. The systems shall be designed GDC 18 is a design-independent companion criterion to GDC 17.

Wording pertaining to additional system examples has been deleted to allow increased flexibility associated with various designs.

APPENDIX A.

ADVANCED REACTOR DESIGN CRITERIA Appendix A to DG-1330, Page A-6 II.

Multiple Barriers Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC 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.

The text related to the nuclear power unit, offsite power system, and onsite power system was deleted to be consistent with ARDC 17.

19 Control room.

A control room shall be provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions. Adequate radiation protection shall be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem total effective dose equivalent as defined in

§ 50.2 for the duration of the accident.

Adequate habitability measures shall be provided to permit access and occupancy of the control room during normal operations and under 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.

The criterion was updated to remove specific emphasis on LOCAs, which may be not appropriate.

Reference to whole body, or its equivalent to any part of the body has been updated to the current total effective dose equivalent standard as defined in § 50.2.

A control room habitability requirement beyond that associated with radiation protection has been added to address the concern that nonradionuclide accidents may also affect control room access and occupancy.

The last paragraph of the GDC has been eliminated for the ARDC because it is not applicable to future applicants.

APPENDIX A.

ADVANCED REACTOR DESIGN CRITERIA Appendix A to DG-1330, Page A-7 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 malfunctions.

The protection system shall be designed to ensure that specified acceptable fuel design limits are not exceeded during any anticipated operational occurrence accounting for a single malfunction of the reactivity control systems.

Text has been added to clarify that the protection system is designed to protect the specified acceptable fuel design limits for anticipated operational occurrences (AOOs) in combination with a single failure; the protection system does not have to protect the specified acceptable fuel design limits during a postulated accident in combination with a single failure. The example was deleted to make the ARDC technology neutral.

26 Reactivity control systems.

Reactivity control systems shall include the following capabilities:

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

(2) A means of shutting down the reactor and maintaining a safe shutdown under design-basis event conditions, with appropriate margin for malfunctions, shall be provided. A Recent licensing activity associated with the application of GDC 26 and GDC 27 to new reactor designs Response to Gap Analysis Summary Report for Reactor System Issues, (Ref. 26) and Response to NuScale Gap Analysis Summary Report for Reactivity Control Systems, Addressing Gap 11, General Design Criteria 26, (Ref. 27), revealed that additional clarity could be provided in the area of reactivity control requirements. ARDC 26 combines the scope of GDC 26 and GDC 27. The development of 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);

the current GDC 26 and 27; the definition of safety-related SSC in 10 CFR 50.2; and SECY-94-084, Policy and Technical Issues Associated with the Regulatory Treatment of Non-Safety Systems

APPENDIX A.

ADVANCED REACTOR DESIGN CRITERIA Appendix A to DG-1330, Page A-8 III.

Reactivity Control Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC second means of reactivity control shall be provided that is independent, diverse, and capable of achieving and maintaining safe shutdown under design-basis event conditions.

(3) A system for holding the reactor subcritical under cold conditions shall be provided.

in Passive Plant Designs (Ref. 29); and the prior application of reactivity control requirements.

Current GDC 26, first sentence, states that two reactivity control systems of different design principles shall be provided. In addition, the NRC has not licensed a power reactor that did not provide two 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.

ADVANCED REACTOR DESIGN CRITERIA Appendix A to DG-1330, Page A-9 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.

ADVANCED REACTOR DESIGN CRITERIA Appendix A to DG-1330, Page A-10 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.

The reactivity control systems shall be designed with appropriate limits on the potential amount and rate of reactivity increase to ensure that the effects of postulated reactivity accidents can neither (1) result in damage to the reactor coolant boundary greater than limited local yielding nor (2) sufficiently 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.

APPENDIX A.

ADVANCED REACTOR DESIGN CRITERIA Appendix A to DG-1330, Page A-11 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.

Components that are part of the reactor coolant boundary shall be designed, fabricated, erected, and tested to the highest quality standards practical. Means shall be provided for detecting and, to the extent practical, identifying the location of the source of reactor coolant leakage.

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.

31 Fracture prevention of reactor coolant boundary.

The reactor coolant boundary shall be designed with sufficient margin to ensure that when stressed under operating, 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 design shall reflect service temperatures, 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.

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.

Specific examples are added to the ARDC to account for the high design and operating temperatures and unique potential coolants.

APPENDIX A.

ADVANCED REACTOR DESIGN CRITERIA Appendix A to DG-1330, Page A-12 IV. Fluid Systems Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC 32 Inspection of reactor coolant boundary.

Components that are part of the reactor coolant boundary shall be designed to permit (1) periodic inspection and functional testing of important areas and features to assess their structural and leaktight integrity, and (2) an appropriate material surveillance program for the reactor vessel.

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 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.

A system to maintain reactor coolant inventory for protection against small breaks in the reactor coolant boundary shall be provided as necessary to ensure that specified acceptable fuel design limits are not exceeded as a result of reactor coolant inventory loss due to leakage from the reactor coolant boundary and rupture of small piping or other small components that are part of the boundary. The system shall be designed to ensure that the system safety function can be accomplished using the piping, pumps, and valves used to maintain reactor coolant inventory during normal reactor operation.

ARDC 33 was relabeled as inventory maintenance to provide more flexibility for advanced reactor designs. The first sentence is modified so that it ends with...shall be provided as necessary and is combined with the second sentence as necessary to ensure (without the opening phrase The system safety function shall be) to recognize that the inventory control system may be unnecessary for some designs to maintain safety functions that ensure fuel design limits are not exceeded.

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 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.

ADVANCED REACTOR DESIGN CRITERIA Appendix A to DG-1330, Page A-13 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.

A system to remove residual heat shall be provided. For normal operations and anticipated operational occurrences, the system safety function shall be to transfer fission product decay heat and other residual heat from the reactor core at a rate such that specified acceptable fuel design limits and the design conditions of the reactor coolant boundary are not exceeded.

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.

In most advanced reactor designs, a single system (i.e the residual heat removal system) is provided to perform both the residual heat removal and emergency core cooling functions. In this case, the single system would be designed to meet the requirements of ARDC 34 and ARDC 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 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.

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.

ADVANCED REACTOR DESIGN CRITERIA Appendix A to DG-1330, Page A-14 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.

A system to provide sufficient emergency core cooling shall be provided. The system safety function shall be to transfer heat from the reactor core such that effective core cooling is maintained and fuel damage is limited.

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

In most advanced reactor designs, a single system (i.e the residual heat removal system) is provided to perform both the residual heat removal and emergency core cooling functions. In this case, the single system would be designed to meet the requirements of ARDC 34 and ARDC 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 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.

A system that provides emergency core cooling shall be designed to permit appropriate periodic inspection of important components to ensure the integrity and capability of the system.

In most advanced reactor designs, a single system (i.e the residual heat removal system) is provided to perform both the residual heat removal and emergency core cooling functions. In this case, the single system would be designed to meet the requirements of ARDC 34 and ARDC 35. (for more discussion see NUREG-0968

APPENDIX A.

ADVANCED REACTOR DESIGN CRITERIA Appendix A to DG-1330, Page A-15 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.

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

APPENDIX A.

ADVANCED REACTOR DESIGN CRITERIA Appendix A to DG-1330, Page A-16 IV. Fluid Systems Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC and, under conditions as close to design as practical, the performance of the full operational sequence that brings the system into operation, including operation of any associated systems and interfaces necessary to transfer decay heat to the ultimate heat sink.

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 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.

ADVANCED REACTOR DESIGN CRITERIA Appendix A to DG-1330, Page A-17 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.

A system to remove heat from the reactor containment shall be provided as necessary to maintain the containment pressure and temperature within acceptable limits following postulated accidents.

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

as necessary is meant to condition an ARDC 38 application to designs requiring heat removal for conventional containments that are found to require heat removal measures.

The LOCA reference has been removed to provide for any postulated accident that might affect the containment structure.

Containment structure safety system redundancy is addressed in the second paragraph.

39 Inspection of containment heat removal system.

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.

Examples were deleted to make the ARDC technology neutral.

40 Testing of containment heat removal system.

The containment heat removal system shall be designed to permit appropriate periodic functional testing to ensure (1) the structural and leaktight integrity of its components, (2) the operability and performance of the system components, and (3) the operability of the system as a whole, and under conditions as close to the design as practical, the performance of the full operational sequence that brings the system 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 containment heat removal.

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.

Reference to the operation of applicable portions of the protection system, structural and equipment cooling, and power transfers is

APPENDIX A.

ADVANCED REACTOR DESIGN CRITERIA Appendix A to DG-1330, Page A-18 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.

Systems to control fission products and other substances that may be released into the reactor containment shall be provided as necessary to reduce, consistent with the functioning of other associated systems, the concentration and quality of fission products released to the environment following postulated accidents and to control the concentration of other substances in the containment atmosphere following postulated accidents to ensure that containment integrity and other safety functions are maintained.

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.

Advanced reactors offer potential for reaction product generation that is different from that associated with clad metal-water interactions. Therefore, the terms hydrogen and oxygen are removed while other substances is retained to allow for exceptions.

Considering that a passive containment cooling system may be used or that the containment may have an additional safety function other than radionuclide retention, additional wording for maintaining safety-functions is added.

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

42 Inspection of containment atmosphere cleanup systems.

Same as GDC 43 Testing of containment atmosphere cleanup systems.

The containment atmosphere cleanup systems shall be designed to permit appropriate periodic functional testing to ensure (1) the structural and leaktight integrity of its components, (2) the operability and performance of the system components, and (3) the operability of the systems as a whole and, under conditions as close to design as practical, the performance of the full operational sequence that brings the systems into operation, including the operation of associated systems.

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

Examples of active systems under item (2) have been deleted, both to conform to similar wording in ARDC 37 and 40 and ensure that passive as well as active system components are considered.

APPENDIX A.

ADVANCED REACTOR DESIGN CRITERIA Appendix A to DG-1330, Page A-19 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.

A system to transfer heat from structures, systems, and components important to safety to an ultimate heat sink shall be provided, as necessary, to transfer the combined heat load of these structures, systems, and components under normal operating and accident conditions.

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.

This renamed ARDC accounts for advanced reactor design system differences to include cooling requirements for SSCs, if applicable; this ARDC does not address the residual heat removal system required under ARDC 34, and ECCS system under ARDC 35 The GDC reference to electric power was removed. Refer to ARDC 17 concerning those systems that require electric power.

45 Inspection of structural and equipment cooling systems.

The structural and equipment cooling systems shall be designed to permit appropriate periodic inspection of important components, such as heat exchangers and piping, to ensure the integrity and capability of the systems.

This renamed ARDC accounts for advanced reactor system design differences to include possible cooling requirements for SSCs important to safety.

46 Testing of structural and equipment cooling systems.

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

APPENDIX A.

ADVANCED REACTOR DESIGN CRITERIA Appendix A to DG-1330, Page A-20 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 conditions as close to design as practical, the performance of the full operational sequences that bring the systems into operation for reactor shutdown and postulated accidents, including the operation of associated systems.

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) 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.

The reactor containment structure, including access openings, penetrations, and the containment heat removal system shall be designed so that the containment structure and its internal compartments can accommodate, without exceeding the design leakage rate and with sufficient margin, the calculated pressure and temperature conditions resulting from postulated accidents.

This margin shall reflect consideration of (1) the effects of potential energy sources that have not been included in the determination of the peak conditions, (2) the limited experience and experimental data available for defining accident phenomena and containment responses, and (3) the conservatism of the calculational model and input parameters.

ARDC 50 specifically addresses a containment structure in the opening sentence and ARDC 51-57 support the containment structures design basis. Therefore, ARDC 51-57 are modified by adding the word structure to highlight the containment structure-specific criteria.

The phrase loss-of-coolant accident is LWR specific because this is understood to be the limiting containment structure accident for an LWR design. It is replaced by the phrase postulated accident to allow for consideration of the design-specific containment structure limiting accident for non-LWR designs.

The example at the end of subpart 1 of the ARDC is LWR specific and therefore deleted.

APPENDIX A.

ADVANCED REACTOR DESIGN CRITERIA Appendix A to DG-1330, Page A-21 V.

Reactor Containment Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC 51 Fracture prevention of containment pressure boundary.

The boundary of the reactor containment structure shall be designed with sufficient margin to ensure that, under operating, maintenance, testing, and postulated accident conditions, (1) its materials behave in a nonbrittle manner and (2) the probability of rapidly propagating fracture is minimized. The design shall reflect consideration of service temperatures and other conditions of the containment boundary materials during operation, maintenance, testing, and postulated accident conditions, and the uncertainties in determining (1) material properties, (2) residual, steady-state, and transient stresses, and (3) size of flaws.

ARDC 51-57 support ARDC 50, which specifically applies to non-LWR designs that use a fixed containment structure. Therefore, the word structure is added to each of these ARDC to clearly convey the understanding that this criterion applies to designs employing containment structures.

The term ferritic was removed to avoid limiting the scope of the criterion to ferritic materials. With this revision, the staff believes that this criterion is more broadly applicable to all non-LWR designs.

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.

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

ARDC 51-57 support ARDC 50, which specifically applies to non-LWR designs that use a fixed containment structure. Therefore, the word structure is added to each of these ARDC to clearly convey the understanding that this criterion applies to designs employing containment structures.

53 Provisions for containment testing and inspection.

The reactor containment structure shall be designed to permit (1) appropriate periodic inspection of all important areas, such as penetrations, (2) an appropriate surveillance program, and (3) periodic testing at containment design pressure of the leak-tightness of penetrations that have resilient seals and expansion bellows.

ARDC 51-57 support ARDC 50, which specifically applies to non-LWR designs that use a fixed containment structure. Therefore, the word structure is added to each of these ARDC to clearly convey the understanding that this criterion only applies to designs employing containment structures.

54 Piping systems penetrating containment.

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

APPENDIX A.

ADVANCED REACTOR DESIGN CRITERIA Appendix A to DG-1330, Page A-22 V.

Reactor Containment Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC these piping systems. Such piping systems shall be designed with the capability to verify, by testing, the operational readiness of any isolation valves and associated apparatus periodically and to confirm that valve leakage is within acceptable limits.

removed because the containment is a barrier between the fission products and the environment. There are diverse advanced reactor designs and, hence, there is no single containment concept. In all cases, the rules for containment penetrations to fulfill containment 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.

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

APPENDIX A.

ADVANCED REACTOR DESIGN CRITERIA Appendix A to DG-1330, Page A-23 V.

Reactor Containment Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC containment isolation valves, as follows, unless it can be demonstrated that the containment isolation provisions for a specific class of lines, such as instrument lines, are acceptable on some other defined basis:

(1) One locked closed isolation valve inside and one locked closed isolation valve outside containment; or (2) One automatic isolation valve inside and one locked closed isolation valve outside containment; or (3) One locked closed isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment; or (4) One automatic isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment.

Isolation valves outside containment shall be located as close to containment as practical and upon loss of actuating power, automatic isolation valves shall be designed to take the position 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.

the understanding that this ARDC only applies to designs employing containment structures. The word reactor was removed because the containment is a barrier between the fission products and the environment. There are diverse advanced reactor designs and, hence, there is no single containment concept. In all cases, the rules for containment penetrations to fulfill containment 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.

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.

APPENDIX A.

ADVANCED REACTOR DESIGN CRITERIA Appendix A to DG-1330, Page A-24 V.

Reactor Containment Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC 56 Containment isolation.

Each line that connects directly to the containment atmosphere and penetrates the containment structure shall be provided with containment isolation valves as follows, unless it can be demonstrated that the containment isolation provisions for a specific class of lines, such as instrument lines, are acceptable on some other defined basis:

(1) One locked closed isolation valve inside and one locked closed isolation valve outside containment; or (2) One automatic isolation valve inside and one locked closed isolation valve outside containment; or (3) One locked closed isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment; or (4) One automatic isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment.

Isolation valves outside containment shall be located as close to the containment as practical and upon loss of actuating power, automatic isolation valves shall be designed to take the position that provides greater safety.

ARDC 51-57 support ARDC 50, which specifically applies to non-LWR designs that use a fixed containment structure. Therefore, the word structure is added to each of these ARDC to clearly convey the understanding that this criterion only applies to designs employing containment structures. The word primary in the title and the text was removed, and the word reactor was also removed because the containment is a barrier between the fission products and the environment. There are diverse advanced reactor designs and, hence, there is no single containment concept. In all cases, the rules for containment penetrations to fulfill containment 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.

57 Closed system isolation valves.

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

APPENDIX A.

ADVANCED REACTOR DESIGN CRITERIA Appendix A to DG-1330, Page A-25 V.

Reactor Containment Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC closed, or capable of remote manual operation. This valve shall be outside containment and located as close to the containment as practical. A simple check valve may not be used as the automatic isolation valve.

cases, the rules for containment penetrations to fulfill containment 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.

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.

ADVANCED REACTOR DESIGN CRITERIA Appendix A to DG-1330, Page A-26 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 fuel storage and handling, radioactive waste, and other systems that may contain radioactivity shall be designed to ensure adequate safety under normal and postulated accident conditions. These systems shall be designed (1) with a capability to permit appropriate periodic inspection and testing of 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.

The underlying concept of establishing functional requirements for radioactivity control in fuel storage and fuel handling systems is independent of the design of non-LWR advanced reactors.

However, some advanced designs may use dry fuel storage that incorporates cooling jackets that can be liquid cooled or air cooled to remove heat. This modification to this GDC allows for both liquid and air cooling of the dry fuel storage containers.

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.

Means shall be provided for monitoring the reactor containment atmosphere, effluent discharge paths, and plant environs for radioactivity that may be released from normal operations, including anticipated operational occurrences, and from postulated accidents.

The phrase spaces containing components for recirculation of loss-of-coolant accident fluids was removed to allow for plant designs that do not have LOCA fluids but may have other similar equipment in spaces where radioactivity should be monitored.

Appendix B to DG-1330, Page B-1 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.

SODIUM-COOLED FAST REACTOR DESIGN CRITERIA Appendix B to DG-1330, Page B-2 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.

Same as ARDC Structures, systems, and components important to safety shall be designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions. Noncombustible and fire-resistant materials shall be used wherever practical throughout the unit, particularly in 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.

The phrase containing examples where noncombustible and fire-resistant materials must be used has been broadened to apply to all advanced reactor designs.

Instead of and, the phrase locations with structures, systems, and components (SSCs) important to safety uses or, which is logically correct in this case.

4 Environmental and dynamic effects design bases.

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

The phrase the environmental conditions associated with anticipated operational occurrences has been added to ensure that

APPENDIX B.

SODIUM-COOLED FAST REACTOR DESIGN CRITERIA Appendix B to DG-1330, Page B-3 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 excluded from the design basis when analyses reviewed and approved by the Commission demonstrate that the probability of fluid system piping rupture is extremely low under conditions consistent with the design basis for the piping.

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.

the criterion would apply to all SFR design-basis events, as suggested in NUREG-1368.

A new sentence is added to ensure the designer considers the effects of sodium leakage and associated chemical reactions with SSCs 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.

Same as ARDC The reactor core and associated systems that contribute to reactivity feedback shall be designed so that, in the power operating range, the net effect of the prompt inherent nuclear feedback characteristics tends to compensate for a rapid increase in reactivity.

The wording has been changed to broaden the applicability from coolant systems to additional factors (including structures or other fluids) that may contribute to reactivity feedback. These systems are to be designed to compensate for rapid reactivity increase.

12 Suppression of reactor power oscillations.

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

APPENDIX B.

SODIUM-COOLED FAST REACTOR DESIGN CRITERIA Appendix B to DG-1330, Page B-4 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.

Instrumentation shall be provided to monitor variables and systems over their anticipated ranges for normal operation, for anticipated operational occurrences, and for accident conditions, as appropriate to ensure adequate safety, including those variables and systems that can affect the fission process, the integrity of the reactor core, the primary coolant boundary, and the containment and its associated systems. Appropriate controls shall be provided to maintain these variables and systems within prescribed operating ranges.

Reactor coolant pressure boundary has been relabeled as primary coolant boundary to conform to standard terms used in the liquid-metal reactor (LMR) industry.

The use of the term primary indicates that the SFR-DC are applicable to the primary cooling system, not the intermediate cooling system.

14 Primary coolant boundary.

The primary coolant boundary shall be designed, fabricated, erected, and tested so as to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture.

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

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.

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

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

The use of the term primary indicates that the SFR-DC are applicable only to the primary cooling system, not the intermediate cooling system.

APPENDIX B.

SODIUM-COOLED FAST REACTOR DESIGN CRITERIA Appendix B to DG-1330, Page B-5 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.

A reactor containment consisting of a high-strength, low-leakage, pressure-retaining structure surrounding the reactor and its primary cooling system shall be provided to control the release of radioactivity to the environment and to ensure that the reactor containment design conditions important to safety are not exceeded for as long as postulated accident conditions require.

The containment leakage shall be restricted to be less than that needed to meet the acceptable onsite and offsite dose consequence limits, as specified in 10 CFR 50.34 for postulated accidents.

The Commission approved the staffs recommendation to restrict the leakage of the containment to be less than that needed to meet the acceptable onsite and offsite dose consequence limits in SECY-93-092 (Ref. 7). Therefore, the Commission agreed that the containment leakage for advanced reactors, similar to and including PRISM, NUREG-1368 (Ref. 4) should not be required to meet the essentially leaktight statement in GDC 16.

Furthermore, all past, current, and planned SFR designs use a high-strength, low-leakage, pressure-retaining containment concept, which aims to provide a barrier to contain the fission products and other substances and to control the release of radioactivity to the 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.

SODIUM-COOLED FAST REACTOR DESIGN CRITERIA Appendix B to DG-1330, Page B-6 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.

Electric power systems shall be provided to permit functioning of structures, systems, and components important to safety. The safety function for the systems shall be to provide sufficient capacity, capability, and reliability to ensure that (1) specified acceptable fuel design limits and design conditions of the primary coolant boundary are not exceeded as a result of anticipated operational occurrences and (2) vital functions that rely on electric power are maintained in the event of postulated accidents.

A reliable power system is required for SSCs during postulated accident conditions. Power systems shall be sufficient in capacity, capability, and reliability to ensure vital safety functions are maintained. The emphasis is placed on requiring reliability of power sources rather than prescribing how such reliability can be attained.

Reference to onsite vs. offsite electric power systems was deleted to provide for those reactor designs that do not depend on offsite power for the functioning of SSCs important to safety.

The text related to supplies, including batteries, and the onsite distribution system, was deleted to allow increased flexibility in the design of onsite power systems for advanced reactor designs.

APPENDIX B.

SODIUM-COOLED FAST REACTOR DESIGN CRITERIA Appendix B to DG-1330, Page B-7 II.

Multiple Barriers Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC The onsite electric power systems shall have sufficient independence, redundancy, and testability to perform their safety functions, assuming a single failure.

However, such onsite systems are 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 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.

Same as ARDC.

Electric power systems important to safety shall be designed to permit appropriate periodic inspection and testing of important areas and features, such as wiring, insulation, connections, and switchboards, to assess the continuity of the systems and the condition of their components. The systems shall be designed 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.

GDC 18 is a design-independent companion criterion to GDC 17.

Wording pertaining to additional system examples has been deleted to allow increased flexibility associated with various designs.

The text related to the nuclear power unit, offsite power system, and onsite power system was deleted to be consistent with ARDC 17.

APPENDIX B.

SODIUM-COOLED FAST REACTOR DESIGN CRITERIA Appendix B to DG-1330, Page B-8 II.

Multiple Barriers Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC 19 Control room.

A control room shall be provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions. Adequate radiation protection shall be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem total effective dose equivalent, as defined in

§ 50.2 for the duration of the accident.

Adequate habitability measures shall be provided to permit access and occupancy of the control room during normal operations and under accident conditions.

Adequate protection against sodium aerosols shall be provided to permit access and occupancy of the control room under 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.

The criterion was updated to remove specific emphasis on LOCAs, which may be not appropriate for advanced designs such as the SFR.

Reference to whole body, or its equivalent to any part of the body has been updated to the current total effective dose equivalent standard as defined in § 50.2.

A control room habitability requirement beyond that associated with radiation protection has been added to address the concern that nonradionuclide accidents, including accidental sodium leakage and sodium fire, which could release sodium aerosols, may also affect control room access and occupancy.

The last paragraph of the GDC has been eliminated for the SFR-DC because it is not applicable to future applicants.

APPENDIX B.

SODIUM-COOLED FAST REACTOR DESIGN CRITERIA Appendix B to DG-1330, Page B-9 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.

The protection system shall be designed to fail into a safe state or into a state demonstrated to be acceptable on some other defined basis, if conditions such as disconnection of the system, loss of energy (e.g., electric power, instrument air), or postulated adverse environments (e.g., extreme heat or cold, fire, sodium and sodium reaction products, pressure, steam, water, and radiation) are experienced.

In NUREG-1368, Table 3.3 (page 3-21) (Ref. 4), the NRC staff recommended adding the phrase sodium and sodium reaction products to the list of postulated adverse environments in the GDC.

Therefore, sodium and sodium reaction products are added to the second list of examples in parentheses in SFR-DC 23.

24 Separation of protection and control systems.

Same as GDC 25 Protection system requirements for reactivity control malfunctions.

Same as ARDC The protection system shall be designed to ensure that specified acceptable fuel design limits are not exceeded during any anticipated operational occurrence accounting for a single malfunction of the reactivity control systems.

Text has been added to clarify that the protection system is designed to protect the specified acceptable fuel design limits for AOOs in combination with a single failure; the protection system does not have to protect the specified acceptable fuel design limits during a postulated accident in combination with a single failure. The example was deleted to make the SFR technology neutral.

26 Reactivity control systems.

Same as ARDC Reactivity control systems shall include the following capabilities:

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

APPENDIX B.

SODIUM-COOLED FAST REACTOR DESIGN CRITERIA Appendix B to DG-1330, Page B-10 III.

Reactivity Control Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC anticipated operational occurrences, and with appropriate margin for malfunctions, design limits for fission product barriers are not exceeded.

(2) A means of shutting down the reactor and maintaining a safe shutdown under design-basis event conditions, with appropriate margin for malfunctions, shall be provided. A second means of reactivity control shall be provided that is independent, diverse, and capable of achieving and maintaining safe shutdown under design-basis event conditions.

(3) A system for holding the reactor subcritical under cold conditions shall be provided.

combines the scope of GDC 26 and GDC 27. The development of 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);

the current GDC 26 and 27; the definition of safety-related SSC in 10 CFR 50.2; and SECY-94-084, Policy and Technical Issues Associated with the Regulatory Treatment of Non-Safety Systems in Passive Plant Designs (Ref. 29); and the prior application of reactivity control requirements.

Current GDC 26, first sentence, states that two reactivity control systems of different design principles shall be provided. In addition, the NRC has not licensed a power reactor that did not provide two 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

APPENDIX B.

SODIUM-COOLED FAST REACTOR DESIGN CRITERIA Appendix B to DG-1330, Page B-11 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.

SODIUM-COOLED FAST REACTOR DESIGN CRITERIA Appendix B to DG-1330, Page B-12 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.

SODIUM-COOLED FAST REACTOR DESIGN CRITERIA Appendix B to DG-1330, Page B-13 III.

Reactivity Control Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC 28 Reactivity limits.

The reactivity control systems shall be designed with appropriate limits on the potential amount and rate of reactivity increase to ensure that the effects of postulated reactivity accidents can neither (1) result in damage to the primary coolant boundary greater than limited local yielding nor (2) sufficiently disturb the core, its support structures or other reactor vessel internals to impair significantly the capability to cool the core.

Reactor coolant pressure boundary has been relabeled as primary coolant boundary to conform to standard terms used in the LMR industry. The use of the term primary indicates that the SFR-DC are applicable to the primary cooling system, not the intermediate cooling system.

The list of postulated reactivity accidents has been deleted.

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.

Components that are part of the primary coolant boundary shall be designed, fabricated, erected, and tested to the highest quality standards practical. Means shall be provided for detecting and, to the extent practical, identifying the location of the source of primary coolant leakage.

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

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 reactor primary coolant boundary per NUREG-1368 (page 3-38).

31 Fracture prevention of primary coolant boundary.

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

APPENDIX B.

SODIUM-COOLED FAST REACTOR DESIGN CRITERIA Appendix B to DG-1330, Page B-14 IV.

Fluid Systems Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC design shall reflect consideration of service temperatures, 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.

The use of the term primary indicates that the SFR-DC are applicable only to the primary cooling system, not the intermediate cooling system.

Specific examples are added to the SFR-DC to account for the high design and operating temperatures and sodium coolant.

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.

Components that are part of the primary coolant boundary shall be designed to permit (1) periodic inspection and functional testing of important areas and features to assess their structural and leaktight integrity, and (2) an appropriate material surveillance program for the reactor vessel.

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

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 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.

A system to maintain primary coolant inventory for protection against small breaks in the primary coolant boundary shall be provided as necessary to ensure that specified acceptable fuel design limits are not exceeded as a result of primary coolant inventory loss due to leakage from the primary coolant boundary and rupture of small piping or other small This SFR-DC was retitled as inventory maintenance to provide more flexibility for advanced reactor designs.

The first sentence is modified so that it ends with...shall be provided as necessary and is combined with the second sentence as necessary to ensure (without the opening phrase, The system safety function shall be) to recognize that the inventory

APPENDIX B.

SODIUM-COOLED FAST REACTOR DESIGN CRITERIA Appendix B to DG-1330, Page B-15 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 designed to ensure that the system safety function can be accomplished using the piping, pumps, and valves used to maintain primary coolant inventory during normal reactor operation.

control system may be unnecessary for some designs to maintain safety functions that ensure fuel design limits are not exceeded.

Reactor coolant pressure boundary has been relabeled as 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.

A system to remove residual heat shall be provided. For normal operations and anticipated operational occurrences, the system safety function shall be to transfer fission product decay heat and other residual heat from the reactor core at a rate such that specified acceptable fuel design limits and the design conditions of the primary coolant boundary are not exceeded.

Suitable redundancy in components and features and suitable interconnections leak detection, and isolation capabilities, shall In most advanced reactor designs the residual heat removal system is designed to meet the requirements of SFR-DC 34 and SFR-DC 35 (for more discussion see NUREG-0968 (Ref. 5) and NUREG-1368 (Ref. 4)).

It is anticipated that the RHR system for non-LWRs will have the same regulatory treatment as the current LWR fleet.

Reactor coolant pressure boundary has been relabeled as primary coolant boundary to reflect that the SFR primary system operates at low-pressure and to conform to standard terms used in

APPENDIX B.

SODIUM-COOLED FAST REACTOR DESIGN CRITERIA Appendix B to DG-1330, Page B-16 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 accomplished, assuming a single failure.

the LMR industry. The use of the term primary indicates that the 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.

Same as ARDC A system to provide sufficient emergency core cooling shall be provided. The system safety function shall be to transfer heat from the reactor core such that effective core cooling is maintained and fuel damage is limited.

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

In most advanced reactor designs, a single system (i.e the residual heat removal system) is provided to perform both the residual heat removal and emergency core cooling functions. In this case, the single system would be designed to meet the requirements of SFR-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. 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.

SODIUM-COOLED FAST REACTOR DESIGN CRITERIA Appendix B to DG-1330, Page B-17 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.

Same as ARDC A system that provides emergency core cooling shall be designed to permit appropriate periodic inspection of important components to ensure the integrity and capability of the system.

In most advanced reactor designs, a single system (i.e the residual heat removal system) is provided to perform both the residual heat removal and emergency core cooling functions. In this case, the single system would be designed to meet the requirements of SFR-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.

SODIUM-COOLED FAST REACTOR DESIGN CRITERIA Appendix B to DG-1330, Page B-18 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.

Same as ARDC A system that provides emergency core cooling shall be designed to permit appropriate periodic functional testing to ensure (1) the structural and leaktight integrity of its components, (2) the operability and performance of the system components, and (3) the operability of the system as a whole and, under conditions as close to design as practical, the performance of the full operational sequence that brings the system into operation, including operation of any associated systems and interfaces necessary to transfer decay heat to the ultimate heat sink.

In most advanced reactor designs, a single system (i.e the residual heat removal system) is provided to perform both the residual heat removal and emergency core cooling functions. In this case, the single system would be designed to meet the requirements of SFR-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 37.

Specific mention of pressure testing has been removed yet remains a potential requirement should it be necessary as a

APPENDIX B.

SODIUM-COOLED FAST REACTOR DESIGN CRITERIA Appendix B to DG-1330, Page B-19 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 A system to remove heat from the reactor containment shall be provided as necessary to maintain the containment pressure and temperature within acceptable limits following postulated accidents.

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

as necessary is meant to condition an SFR-DC 38 application to designs requiring heat removal for conventional containments that are found to require heat removal measures.

The LOCA reference has been removed to provide for any postulated accident that might affect the containment structure.

Containment structure safety system redundancy is addressed in the second paragraph.

39 Inspection of containment heat removal system.

Same as ARDC Examples were deleted to make the SFR-DC technology neutral.

APPENDIX B.

SODIUM-COOLED FAST REACTOR DESIGN CRITERIA Appendix B to DG-1330, Page B-20 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.

Same as ARDC The containment heat removal system shall be designed to permit appropriate periodic functional testing to ensure (1) the structural and leaktight integrity of its components, (2) the operability and performance of the system components, and (3) the operability of the system as a whole, and under conditions as close to the design as practical, the performance of the full operational sequence that brings the system 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 containment heat removal.

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.

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.

Same as ARDC Systems to control fission products and other substances that may be released into the reactor containment shall be provided as necessary to reduce, consistent with the functioning of other associated systems, the concentration and quality of fission products released to the environment following postulated accidents and to control the concentration of other substances in the containment atmosphere following postulated accidents to ensure that containment integrity is maintained.

Advanced reactors offer potential for reaction product generation that is different from that associated with clad metal-water interactions. Therefore, the terms hydrogen and oxygen are removed while other substances is retained to allow for exceptions.

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

APPENDIX B.

SODIUM-COOLED FAST REACTOR DESIGN CRITERIA Appendix B to DG-1330, Page B-21 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.

Same as ARDC The containment atmosphere cleanup systems shall be designed to permit appropriate periodic functional testing to ensure (1) the structural and leaktight integrity of its components, (2) the operability and performance of the system components, and (3) the operability of the systems as a whole and, under conditions as close to design as practical, the performance of the full operational sequence that brings the systems into operation, including the operation of associated systems.

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

Examples of active systems under item (2) have been deleted, both to conform to similar wording in ARDC 37 and 40 and ensure that passive as well as active system components are considered.

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.

Same as ARDC A system and components important to safety to an ultimate heat sink shall be provided, as necessary, to transfer the combined heat load of these structures, systems, and components under normal operating and accident conditions.

This renamed SFR-DC accounts for advanced reactor design system differences to include cooling requirements for SSCs, if applicable; this SFR-DC does not address the residual heat removal system required under SFR-DC 34, and ECCS system under SFR-DC 35.

APPENDIX B.

SODIUM-COOLED FAST REACTOR DESIGN CRITERIA Appendix B to DG-1330, Page B-22 IV.

Fluid Systems Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC 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.

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

45 Inspection of structural and equipment cooling systems.

Same as ARDC The structural and equipment cooling systems shall be designed to permit appropriate periodic inspection of important components, such as heat exchangers and piping, to ensure the integrity and capability of the systems.

This renamed ARDC accounts for advanced reactor system design differences to include possible cooling requirements for SSCs important to safety.

46 Testing of structural and equipment cooling systems.

Same as ARDC The structural and equipment cooling systems shall be designed to permit appropriate periodic functional testing to ensure (1) the structural and leaktight integrity of their components, (2) the operability and performance of the system components, and (3) the operability of the systems as a whole and, under conditions as close to design as practical, the performance of the full operational sequences that bring the systems into operation for reactor shutdown and postulated accidents, including the operation of associated systems.

This renamed ARDC accounts for advanced reactor system design differences to include possible cooling requirements for SSCs important to safety. 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) 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.

SODIUM-COOLED FAST REACTOR DESIGN CRITERIA Appendix B to DG-1330, Page B-23 V.

Reactor Containment Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC 50 Containment design basis.

Same as ARDC The reactor containment structure, including access openings, penetrations, and the containment heat removal system shall be designed so that the containment structure and its internal compartments can accommodate, without exceeding the design leakage rate and with sufficient margin, the calculated pressure and temperature conditions resulting from postulated accidents.

This margin shall reflect consideration of (1) the effects of potential energy sources that have not been included in the determination of the peak conditions, (2) the limited experience and experimental data available for defining accident phenomena and containment responses, and (3) the conservatism of the calculational model and input parameters SFR-DC 50 specifically addresses a containment structure in the opening sentence and SFR-DC 51-57 support the containment structures design basis. Therefore, SFR-DC 51-57 are modified by adding the word structure to highlight the containment structure-specific criteria.

The phrase loss-of-coolant accident is LWR specific because this is understood to be the limiting containment structure accident for an LWR design. It is replaced by the phrase postulated accident to allow for consideration of the design-specific containment structure limiting accident for non-LWR designs.

The example at the end of subpart 1 of the ARDC is LWR specific and therefore deleted 51 Fracture prevention of containment pressure boundary.

Same as ARDC The boundary of the reactor containment structure shall be designed with sufficient margin to ensure that, under operating, maintenance, testing, and postulated accident conditions, (1) its materials behave in a nonbrittle manner and (2) the probability of rapidly propagating fracture is minimized. The design shall reflect consideration of service temperatures and other conditions of the containment boundary materials during operation, maintenance, testing, and postulated accident conditions, and the uncertainties in determining (1) material properties, (2) residual, steady-state, and transient stresses, and (3) size of flaws.

SFR-DC 51-57 support SFR-DC 50, which specifically applies to non-LWR designs that use a fixed containment structure.

Therefore, the word structure is added to each of these SFR-DC to clearly convey the understanding that this criterion applies to designs employing containment structures.

The term ferritic was removed to avoid limiting the scope of the criterion to ferritic materials. With this revision, the staff believes that this criterion is more broadly applicable to all non-LWR designs.

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.

Same as ARDC The reactor containment structure and other equipment that may be subjected to containment test conditions shall be SFR-DC 51-57 support SFR-DC 50, which specifically applies to non-LWR designs that use a fixed containment structure.

Therefore, the word structure is added to each of these SFR-DC

APPENDIX B.

SODIUM-COOLED FAST REACTOR DESIGN CRITERIA Appendix B to DG-1330, Page B-24 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 conducted at containment design pressure.

to clearly convey the understanding that this criterion applies to designs employing containment structures.

53 Provisions for containment testing and inspection.

Same as ARDC The reactor containment structure shall be designed to permit (1) appropriate periodic inspection of all important areas, such as penetrations, (2) an appropriate surveillance program, and (3) periodic testing at containment design pressure of the leak-tightness of penetrations that have resilient seals and expansion bellows.

SFR-DC 51-57 support SFR-DC 50, which specifically applies to non-LWR designs that use a fixed containment structure.

Therefore, the word structure is added to each of these SFR-DC to clearly convey the understanding that this criterion only applies to designs employing containment structures.

54 Piping systems penetrating containment.

Piping systems penetrating the reactor containment structure shall be provided with leak detection, isolation, and containment capabilities that have redundancy, reliability, and performance capabilities necessary to perform the containment safety function and that reflect the importance to safety of preventing radioactivity releases from containment through these piping systems. Such piping systems shall be designed with the capability to verify, by testing, the operational readiness of any isolation valves and associated apparatus periodically and to confirm that valve leakage is within acceptable limits.

The word structure was added to this SFR-DC to clearly convey the understanding that this criterion only applies to designs employing containment structures.

Not all penetrations will provide a release path to the atmosphere.

Piping that may be of interest in the case of an SFR design is for the intermediate heat transport system and the residual heat removal system. Based on stakeholder input, a designer may be able to satisfactorily demonstrate that containment isolation valves are not required for an SFR design. This rewording for the SFR-DC provides a designer the opportunity to present the safety case 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.

SODIUM-COOLED FAST REACTOR DESIGN CRITERIA Appendix B to DG-1330, Page B-25 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.

Each line that is part of the primary coolant boundary and that penetrates the reactor containment structure shall be provided with containment isolation valves as follows, unless it can be demonstrated that the containment isolation provisions for a specific class of lines, such as instrument lines, are acceptable on some other defined basis:

(1) One locked closed isolation valve inside and one locked closed isolation valve outside containment; or (2) One automatic isolation valve inside and one locked closed isolation valve outside containment; or (3) One locked closed isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment; or (4) One automatic isolation valve inside and one automatic isolation valve outside containment. A simple check valve The word structure was added to this SFR-DC to clearly convey the understanding that this criterion only applies to designs employing containment structures. In some cases, the word the was also added to make the phrase grammatically correct.

The title of SFR-DC 55 is the Primary coolant boundary penetrating containment. The SFR intermediate loop is a separate closed system that does not allow any direct mixing of intermediate fluid with the primary coolant sodium. The tubing of the IHX and associated intermediate loop piping inside the reactor vessel are a part of the primary coolant boundary. SFR-DC 57, Closed system isolation valves, addresses closed systems that penetrate containment and would be the appropriate place to address a closed system, such as an intermediate loop, that penetrates containment and is not part of the primary coolant boundary (in its entirety).

This is similar to the treatment of the main steam system and the steam generator in a pressurized-water reactor.

APPENDIX B.

SODIUM-COOLED FAST REACTOR DESIGN CRITERIA Appendix B to DG-1330, Page B-26 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 containment.

Isolation valves outside containment shall be located as close to containment as practical and, upon loss of actuating power, automatic isolation valves shall be designed to take the position 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.

Reactor coolant pressure boundary has been relabeled as 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 use of the term primary implies the SFR-DC are applicable to the primary cooling system, not the intermediate cooling system.

APPENDIX B.

SODIUM-COOLED FAST REACTOR DESIGN CRITERIA Appendix B to DG-1330, Page B-27 V.

Reactor Containment Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC 56 Containment isolation.

Same as ARDC Each line that connects directly to the containment atmosphere and penetrates the containment structure shall be provided with containment isolation valves as follows, unless it can be demonstrated that the containment isolation provisions for a specific class of lines, such as instrument lines, are acceptable on some other defined basis:

(1) One locked closed isolation valve inside and one locked closed isolation valve outside containment; or (2) One automatic isolation valve inside and one locked closed isolation valve outside containment; or (3) One locked closed isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment; or (4) One automatic isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment.

SFR-DC 51-57 support SFR-DC 50, which specifically applies to non-LWR designs that use a fixed containment structure.

Therefore, the word structure is added to each of these SFR-DC to clearly convey the understanding that this criterion only applies to designs employing containment structures. The word primary in the title and the text was removed, and the word reactor was also removed because the containment is a barrier between the fission products and the environment. There are diverse advanced reactor designs and, hence, there is no single containment concept.

In all cases, the rules for containment penetrations to fulfill containment 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.

57 Closed system isolation valves.

Each line that penetrates the reactor containment structure and is neither part of the primary coolant boundary nor connected directly to the containment atmosphere shall have at least one containment isolation valve unless it can be demonstrated that the containment safety function can be met without an isolation valve and assuming failure of a single active component. The isolation valve, if required, shall be either automatic, or locked The word structure was added to this SFR-DC to clearly convey the understanding that this criterion applies to designs employing containment structures. In some cases, the word the was also added to make the phrase grammatically correct.

Reactor coolant pressure boundary has been relabeled as primary coolant boundary to reflect that the SFR primary system operates at low-pressure and to conform to standard terms used in

APPENDIX B.

SODIUM-COOLED FAST REACTOR DESIGN CRITERIA Appendix B to DG-1330, Page B-28 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 be outside containment and located as close to the containment as practical. A simple check valve may not be used as the automatic isolation valve.

the LMR industry. The use of the term primary implies the SFR-DC are applicable to the primary cooling system, not the intermediate cooling system.

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.

Same as ARDC The fuel storage and handling, radioactive waste, and other systems that may contain radioactivity shall be designed to ensure adequate safety under normal and postulated accident conditions. These systems shall be designed (1) with a capability to permit appropriate periodic inspection and testing of components important to safety, (2) with suitable shielding for radiation protection, (3) with appropriate containment, The underlying concept of establishing functional requirements for radioactivity control in fuel storage and fuel handling systems is independent of the design of non-LWR reactors. However, some advanced designs may use dry fuel storage that incorporates cooling jackets that can be liquid cooled or air cooled to remove heat. This modification to this GDC allows for both liquid and air cooling of the dry fuel storage containers.

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.

Means shall be provided for monitoring the reactor containment atmosphere, spaces containing components for primary system sodium and cover gas cleanup and processing, effluent discharge paths, and the plant environs for radioactivity that In NUREG-1368, Table 3.3 (page 3-25), the NRC staff recommended deleting the GDC 64 phrase spaces containing components for recirculation of loss-of-coolant accident fluids.

Otherwise, the NRC staff noted that criterion requirements are independent of the design of SFRs (page 3-55).

APPENDIX B.

SODIUM-COOLED FAST REACTOR DESIGN CRITERIA Appendix B to DG-1330, Page B-29 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 operational occurrences, and from postulated accidents.

The staff added text to identify other SFR plant areas that should 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.

If an intermediate coolant system is provided, then the system shall be designed to transport heat from the primary coolant system to the energy conversion system as required.

The intermediate coolant system shall be designed with sufficient margin to assure that (1) the design conditions of the intermediate coolant boundary are not exceeded during normal operations, including anticipated occupational occurrences, and (2) the integrity of the primary coolant boundary is maintained during intermediate coolant system accidents.

NUREG-1368 (Ref. 4) (page 3-57), Section 3.2.4.5, suggested the need for a separate criterion for the intermediate coolant system.

Also, separate criteria were included in NUREG-0968 (Ref. 5)

(Criterion 31, Design of Intermediate Cooling System, and Criterion 33, Inspection of Intermediate Cooling System).

The staff revised SFR-DC 70 to focus on the function of the intermediate coolant system, and to use language that is consistent with other design criteria. The discussion related to sodium leakage and required barriers was moved to a new SFR-DC 78.

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.

SODIUM-COOLED FAST REACTOR DESIGN CRITERIA Appendix B to DG-1330, Page B-30 VII. Additional SFR-DC Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC 71 Primary coolant & cover gas purity control.

Systems shall be provided as necessary to maintain the purity of primary coolant sodium and cover gas within specified design limits. These limits shall be based on consideration of (1) chemical attack, (2) fouling and plugging of passages, and (3) radionuclide concentrations, and (4) air or moisture ingress as a result of a leak of cover gas.

The NRC considered DOEs proposed SFR-DC 71 and made changes based on the 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 (pages 12-13).

NUREG 1368 (Ref. 4) (page 3-57), Section 3.2.4.6, suggested the 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.

Heating systems shall be provided for systems and components important to safety, which contain or could be required to contain sodium. These heating systems and their controls shall be appropriately designed to ensure that the temperature distribution and rate of change of temperature in systems and components containing sodium are maintained within design limits assuming a single failure. If plugging of any cover gas line due to condensation or plate out of sodium aerosol or vapor could prevent accomplishing a safety function, the temperature control and the relevant corrective measures associated with that line shall be considered important to safety.

The NRC considered DOEs proposed SFR-DC 72 and made changes based on the 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 (pages 13-14).

NUREG-1368 (Ref. 4) (page 3-56)), Section 3.2.4.2, suggested the need for a separate criterion for sodium heating system. Also, a separate criterion was included in NUREG-0968 (Ref. 5)

(Criterion 7, Sodium Heating Systems).

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 mitigation.

Means to detect sodium leakage and to limit and control the extent of sodium-air and sodium-concrete reactions and to mitigate the effects of fires resulting from these sodium-air and sodium-concrete reactions shall be provided to ensure that the The NRC considered DOEs proposed SFR-DC 73 and made changes based on the 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 (pages 15-16).

APPENDIX B.

SODIUM-COOLED FAST REACTOR DESIGN CRITERIA Appendix B to DG-1330, Page B-31 VII. Additional SFR-DC Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC safety functions of structures, systems, and components important to safety are maintained. Special features, such as inerted enclosures or guard vessels, shall be provided for systems containing sodium.

NUREG-1368 (Ref. 4) (page 3-56), Section 3.2.4.1, suggested the need for a separate criterion for protection against sodium reactions.

Also, a separate criterion was included in NUREG-0968 (Ref. 5)

(Criterion 4, Protection against Sodium and NaK reactions).

74 Sodium/water reaction prevention/mitigation.

Structures, systems, and components containing sodium shall be designed and located to avoid contact between sodium and water and to limit the adverse effects of chemical reactions between sodium and water on the capability of any structure, system, or component to perform any of its intended safety functions. If steam-water is used for energy conversion, to prevent loss of any plant safety function, the sodium-steam generator system shall be designed to detect and contain sodium-water reactions and limit the effects of the energy and reaction products released by such reactions, including mitigation of the effects of any resulting fire involving sodium.

The NRC considered DOEs proposed SFR-DC 74 and made changes based on the 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 (pages 16-18). NUREG-1368 (Ref 4) (page 3-56),

Section 3.2.4.1, suggested the need for a separate criterion for protection against sodium reactions. Also, a separate criterion was included in NUREG-0968 (Ref. 5) (Criterion 4, Protection against Sodium and NaK reactions). Fire considerations are added for consistency with SFR-DC 73.

75 Quality of the intermediate coolant boundary.

Components that are part of the intermediate coolant boundary shall be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed.

This criterion is similar to GDC 30 in 10 CFR Part 50, Appendix A, and is intended to ensure that, similar to the reactor coolant pressure boundary, the intermediate coolant boundary is designed, fabricated, and tested using quality standards and controls sufficient to ensure that failure of the intermediate system would be unlikely.

76 Fracture prevention of the intermediate coolant boundary.

The intermediate coolant boundary shall be designed with sufficient margin to ensure that, when stressed under operating, maintenance, testing, and postulated accident conditions, (1) the boundary behaves in a nonbrittle manner and (2) the probability of rapidly propagating fracture is minimized.

This criterion is similar to GDC 31 in 10 CFR Part 50, Appendix A, and is intended to ensure that, similar to the reactor coolant pressure boundary, the intermediate coolant boundary is designed to avoid brittle and rapidly propagating facture modes.

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.

SODIUM-COOLED FAST REACTOR DESIGN CRITERIA Appendix B to DG-1330, Page B-32 VII. Additional SFR-DC Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC 77 Inspection of the intermediate coolant boundary.

Components that are part of the intermediate coolant boundary shall be designed to permit (1) periodic inspection and functional testing of important areas and features to assess their structural and leaktight integrity commensurate with the systems importance to safety, and (2) an appropriate material surveillance program for the intermediate coolant boundary.

Means shall be provided for detecting and, to the extent practical, identifying the location of the source of coolant leakage.

This criterion is similar to GDC 32 in 10 CFR Part 50, Appendix A, and is intended to ensure that, similar to the reactor coolant pressure boundary, the intermediate coolant boundary is designed to avoid brittle and rapidly propagating fracture modes.

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

APPENDIX B.

SODIUM-COOLED FAST REACTOR DESIGN CRITERIA Appendix B to DG-1330, Page B-33 VII. Additional SFR-DC Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC passive barriers. When the primary coolant system interfaces with a structure, system, or component containing fluid that is chemically compatible with the primary coolant, then the interface location may be a single passive barrier provided that the following conditions are met:

(1) postulated leakage at the interface location does not result in failure of the intended safety functions of structures, systems or components important to safety or result in exceeding the fuel design limits (2) the fluid contained in the structure, system, or component is maintained at a higher pressure than the primary coolant during normal operation, AOOs, shutdown, and accident conditions.

Rather than creating two parallel requirements for the two systems, SFR-78 was created to discuss leakage and required barriers as a generic criterion. The criterion allows for double walled steam generators, intermediate coolant systems connected to steam power system, and systems similar to the PRISM Direct Reactor Auxiliary Cooling System (DRACS).

A paragraph from NUREG 1368 (page 3-41) was added describing the characteristics of the residual heat removal working fluid and its associated operating pressure. This SFR-DC has been worded to explain that an intermediate coolant system may be used if the primary coolant is not chemically compatible with the energy conversion system coolant.

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.

SODIUM-COOLED FAST REACTOR DESIGN CRITERIA Appendix B to DG-1330, Page B-34 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.

A system to maintain cover gas inventory shall be provided as necessary to ensure that the primary coolant sodium design limits are not exceeded as a result of cover gas loss due to leakage from the primary coolant boundary and rupture of small piping or other small components that are part of the primary coolant boundary.

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 on the effects of primary coolant (sodium) loss. A leak in a SFR primary coolant system may expel the cover gas rather than the primary coolant. The cover gas in the SFR performs an important to safety function by protecting the sodium coolant from chemical reactions. The staff created a new SFR-DC rather than adding the 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.

SODIUM-COOLED FAST REACTOR DESIGN CRITERIA Appendix B to DG-1330, Page B-35 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 C to DG-1330, Page C-1 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.

MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGN CRITERIA Appendix C to DG-1330, Page C-2 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.

Same as ARDC Structures, systems, and components important to safety shall be designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions. Noncombustible and fire-resistant materials shall be used wherever practical throughout the unit, particularly in 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.

The phrase containing examples where noncombustible and fire-resistant materials must be used has been broadened to apply to all advanced reactor designs.

Instead of and, the phrase locations with structures, systems, and components (SSCs) important to safety uses or, which is logically correct in this case.

4 Environmental and dynamic effects design bases.

Structures, systems, and components important to safety shall be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents. These structures, systems, and components shall be appropriately protected against dynamic effects, including the effects of missiles originating both inside and outside the reactor helium pressure boundary, pipe whipping, and discharging fluids, that may result from equipment failures and from events and conditions outside the nuclear power unit.

However, dynamic effects associated with postulated pipe This change removes the light-water reactor (LWR) emphasis on loss-of-coolant accidents (LOCAs) that may not apply to every design. For example, helium is not needed in a mHTGR to remove heat from the core during postulated accidents and does not have the same importance as water does to LWR designs to ensure that fuel integrity is maintained. Therefore, a specific reference to LOCAs is not applicable to all designs. LOCAs may still require analysis in conjunction with postulated accidents if they are relevant to the design.

If an mHTGR design proposes using a direct power cycle in which one or more very high-speed, very high-energy gas turbines are

APPENDIX C.

MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGN CRITERIA Appendix C to DG-1330, Page C-3 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 design basis when analyses reviewed and approved by the Commission demonstrate that the probability of fluid system piping rupture is extremely low under conditions consistent with the design basis for the piping.

located inside the reactor helium pressure boundary. The presence of one or more very high-energy turbines inside the primary helium pressure boundary creates the potential that a catastrophic dynamic failure of the gas turbine (e.g., at power) could result in 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 reactor system and associated heat removal, control, and protection systems shall be designed with appropriate margin to ensure that specified acceptable system radionuclide release design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.

The concept of specified acceptable fuel design limits, which prevent additional fuel failures during anticipated operational occurrences (AOOs), has been replaced with that of the specified acceptable system radionuclide release design limits (SARRDL),

which limits the amount of radionuclide inventory that is released by the fuel and surfaces within the helium coolant boundary under 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.

MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGN CRITERIA Appendix C to DG-1330, Page C-4 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.

MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGN CRITERIA Appendix C to DG-1330, Page C-5 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.

Same as ARDC The reactor core and associated systems that contribute to reactivity feedback shall be designed so that, in the power operating range, the net effect of the prompt inherent nuclear feedback characteristics tends to compensate for a rapid increase in reactivity.

The wording has been changed to broaden the applicability from coolant systems to additional factors (including structures or other fluids) that may contribute to reactivity feedback. These systems are to be designed to compensate for rapid reactivity increase.

12 Suppression of reactor power oscillations.

The reactor core and associated control and protection systems shall be designed to ensure that power oscillations that can result in conditions exceeding specified acceptable system radionuclide release design limits are not possible or can be reliably and readily detected and suppressed.

Helium in the mHTGR does not affect reactor core susceptibility to coolant-induced power oscillations; therefore, a separate mHTGR-specific DC is appropriate. The word coolant was deleted and the specified acceptable fuel design limits were replaced by SARRDLs. The discussion on the SARRDL is given in mHTGR-DC 10.

13 Instrumentation and control.

Instrumentation shall be provided to monitor variables and systems over their anticipated ranges for normal operation, for anticipated operational occurrences, and for accident conditions, as appropriate, to ensure adequate safety, including those variables and systems that can affect the fission process 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.

Reactor coolant pressure boundary has been relabeled as reactor helium pressure boundary to conform to standard terms used for mHTGRs.

The criterion has been modified to reflect the use of the mHTGR functional containment. See the mHTGR-DC 16 rationale.

14 Reactor helium pressure boundary.

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

APPENDIX C.

MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGN CRITERIA Appendix C to DG-1330, Page C-6 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, secondary coolant, or other fluids.

The mHTGR-DC 14 addresses the need to consider leakage of 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.

All systems that are part of the reactor helium pressure boundary, such as the reactor system, vessel system, and heat 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.

Reactor coolant system has been relabeled as reactor helium pressure boundary to conform to standard terms used for mHTGRs.

16 Containment design.

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

The term functional containment is applicable to advanced non-LWRs without a pressure retaining containment structure.

A functional containment can be defined as a barrier, or set of barriers taken together, that effectively limit the physical transport and release of radionuclides to the environment across a full range of normal operating conditions, AOOs, and accident conditions.

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.

MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGN CRITERIA Appendix C to DG-1330, Page C-7 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.

MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGN CRITERIA Appendix C to DG-1330, Page C-8 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.

Electric power systems shall be provided to permit functioning of structures, systems, and components important to safety. The safety function for the systems shall be to provide sufficient capacity, capability, and reliability to ensure that (1) specified acceptable system radionuclide release design limits and design conditions of the reactor helium pressure boundary are not exceeded as a result of anticipated operational occurrences and (2) vital functions that rely on electric power are maintained in the event of postulated accidents.

The onsite electric power systems shall have sufficient independence, redundancy, and testability to perform their safety functions, assuming a single failure.

A reliable power system is required for SSCs during postulated accident conditions. Power systems shall be sufficient in capacity, capability, and reliability to ensure vital safety functions are maintained. The emphasis is placed on requiring reliability of power sources rather than prescribing how such reliability can be attained. The reference to onsite vs. offsite electric power systems was deleted to provide for those reactor designs that do not depend on offsite power for the functioning of SSCs important to safety.

The text related to supplies, including batteries, and the onsite distribution system, was deleted to allow increased flexibility in the design of offsite power systems for advanced reactor designs.

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.

MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGN CRITERIA Appendix C to DG-1330, Page C-9 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.

Same as ARDC Electric power systems important to safety shall be designed to permit appropriate periodic inspection and testing of important areas and features, such as wiring, insulation, connections, and switchboards, to assess the continuity of the systems and the condition of their components. The systems shall be designed 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.

GDC 18 is a design-independent companion criterion to GDC 17.

Wording pertaining to additional system examples has been deleted to allow increased flexibility associated with various designs.

The text related to the nuclear power unit, offsite power system, and onsite power system was deleted to be consistent with mHTGR-DC 17.

19 Control room.

Same as ARDC A control room shall be provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions. Adequate radiation protection shall be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem total effective dose equivalent as defined in § 50.2 for the duration of the accident.

Adequate habitability measures shall be provided to permit access and occupancy of the control room during normal operations and under accident conditions. Equipment at appropriate locations outside the control room shall be provided The criterion was updated to remove specific emphasis on LOCAs, which may be not appropriate for advanced designs such as the mHTGR.

Reference to whole body, or its equivalent to any part of the body has been updated to the current total effective dose equivalent standard as defined in § 50.2.

A control room habitability requirement beyond that associated with radiation protection has been added to address the concern that non-radionuclide accidents may also affect control room access and occupancy.

The last paragraph of the GDC has been eliminated for the mHTGR-DC because it is not applicable to future applicants.

APPENDIX C.

MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGN CRITERIA Appendix C to DG-1330, Page C-10 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.

The protection system shall be designed (1) to initiate automatically the operation of appropriate systems, including the reactivity control systems, to ensure that the specified acceptable system radionuclide release design limits is not exceeded as a result of anticipated operational occurrences and (2) to sense accident conditions and to initiate the operation of systems and components important to safety.

Specified acceptable fuel design limits has been replaced with SARRDLs. The concept of using SARRDLs is discussed for GDC 10. The quantitative value of the SARRDL will be design specific. The protection aspect of automatic operation, to protect normal operation and AOO limits, to sense accident conditions, and to initiate mitigating equipment has been preserved.

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.

MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGN CRITERIA Appendix C to DG-1330, Page C-11 III.

Reactivity Control Criterion mHTGR-DC Title and Content NRC Rationale for Adaptions to GDC 25 Protection system requirements for reactivity control malfunctions.

The protection system shall be designed to ensure that specified acceptable system radionuclide release design limits are not exceeded during any anticipated operational occurrence, accounting for a single malfunction of the reactivity control systems.

Specified acceptable fuel design limits is replaced with SARRDLs. The concept of using SARRDLs is discussed for GDC 10.

26 Reactivity control systems.

Same as ARDC Reactivity control systems shall include the following capabilities:

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

(2) A means of shutting down the reactor and maintaining a safe shutdown under design-basis event conditions, with appropriate margin for malfunctions, shall be provided. A second means of reactivity control shall be provided that is independent, diverse, and capable of achieving and maintaining safe shutdown under design-basis event conditions.

(3) A system for holding the reactor subcritical under cold conditions shall be provided.

Recent licensing activity associated with the application of GDC 26 and GDC 27 to new reactor designs Response to Gap Analysis Summary Report for Reactor System Issues, (Ref. 26) and Response to NuScale Gap Analysis Summary Report for Reactivity Control Systems, Addressing Gap 11, General Design Criteria 26, (Ref. 27), revealed that additional clarity could be provided in the area of reactivity control requirements. ARDC 26 combines the scope of GDC 26 and GDC 27. The development of 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); the current GDC 26 and 27; the definition of safety-related SSC in 10 CFR 50.2; and SECY-94-084, Policy and Technical Issues Associated with the Regulatory Treatment of Non-Safety Systems in Passive Plant Designs (Ref. 29); and the prior application of reactivity control requirements.

Current GDC 26, first sentence, states that two reactivity control systems of different design principles shall be provided. In addition, the NRC has not licensed a power reactor that did not provide two 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

APPENDIX C.

MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGN CRITERIA Appendix C to DG-1330, Page C-12 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.

MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGN CRITERIA Appendix C to DG-1330, Page C-13 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.

MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGN CRITERIA Appendix C to DG-1330, Page C-14 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.

The reactor core, including the reactivity control systems, shall be designed with appropriate limits on the potential amount and rate of reactivity increase to ensure that the effects of postulated reactivity accidents can neither (1) result in damage to the reactor helium pressure boundary greater than limited local yielding, nor (2) sufficiently disturb the core, its support structures, or other reactor vessel internals to impair significantly the capability to cool the core.

Reactor coolant pressure boundary has been relabeled as reactor helium pressure boundary to conform to standard terms used for mHTGRs.

The list of postulated reactivity accidents has been deleted. Each design will have to determine its postulated reactivity accidents based on the specific design and associated risk evaluation.

29 Protection against anticipated operational occurrences.

Same as GDC

APPENDIX C.

MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGN CRITERIA Appendix C to DG-1330, Page C-15 IV.

Fluid Systems Criterion mHTGR-DC Title and Content NRC Rationale for Adaptions to GDC 30 Quality of reactor helium pressure boundary.

Components that are part of the reactor helium pressure boundary shall be designed, fabricated, erected, and tested to the highest quality standards practical. Means shall be provided for detecting and, to the extent practical, identifying the location of the source of reactor helium leakage. Means shall be provided for detecting ingress of moisture, air, secondary coolant, or other fluids to within the reactor helium pressure boundary.

Reactor coolant pressure boundary has been relabeled as reactor helium pressure boundary to conform to standard terms used for mHTGRs.

The mHTGR-DC 14 addresses the need to consider leakage of 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.

31 Fracture prevention of reactor helium pressure boundary.

The reactor helium pressure boundary shall be designed with sufficient margin to ensure that, when stressed under operating, 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 design shall reflect consideration of service temperatures, 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.

Reactor coolant pressure boundary has been relabeled as reactor helium pressure boundary to conform to standard terms used for mHTGRs.

Specific examples are added to the mHTGR-DC to account for the high design and operating temperatures and unique potential coolants.

APPENDIX C.

MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGN CRITERIA Appendix C to DG-1330, Page C-16 IV.

Fluid Systems Criterion mHTGR-DC Title and Content NRC Rationale for Adaptions to GDC 32 Inspection of reactor helium pressure boundary.

Components that are part of the reactor helium pressure boundary shall be designed to permit (1) periodic inspection and functional testing of important areas and features to assess their structural and leaktight integrity, and (2) an appropriate material surveillance program for the reactor vessel.

Reactor coolant pressure boundary has been relabeled as reactor helium pressure boundary to conform to standard terms used for mHTGRs.

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.

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.

Not applicable to mHTGR.

The mHTGR does not require reactor coolant inventory 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 normal operations and anticipated operational occurrences, the system safety function shall be to transfer fission product decay heat and other residual heat from the reactor core to an ultimate heat sink at a rate such that specified acceptable system radionuclide release design limits and the design conditions of the reactor helium pressure boundary are not exceeded.

During postulated accidents, the system safety function shall provide effective core cooling.

Suitable redundancy in components and features and suitable interconnections, leak detection, and isolation capabilities shall The word passive was added, based on the definition of a mHTGR. In definitions Section 3.1 of the DOE report titled Guidance for Developing Principal Design Criteria for Advanced (Non-Light-Water) Reactors (Ref. 17), the mHTGR design has a low power density and hence residual heat is removed by a passive system.

Ultimate heat sink has been added to explain that, if mHTGR-DC 44 is deemed not applicable to the design, the residual heat removal system is then required to provide the heat removal path to the ultimate heat sink.

APPENDIX C.

MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGN CRITERIA Appendix C to DG-1330, Page C-17 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 accomplished, assuming a single failure.

Reactor coolant pressure boundary has been relabeled as reactor 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.

Not applicable to mHTGR.

In the mHTGR design the power density and large length to 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.

MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGN CRITERIA Appendix C to DG-1330, Page C-18 IV.

Fluid Systems Criterion mHTGR-DC Title and Content NRC Rationale for Adaptions to GDC 36 Inspection of passive residual heat removal system.

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

The word passive was added, based on the definition of a mHTGR. In definitions Section 3.1 of DOE report titled Guidance for Developing Principal Design Criteria for Advanced (Non-Light-Water) Reactors (Ref. 17), the mHTGR design has a low 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.

The passive residual heat removal system shall be designed to permit appropriate periodic functional testing to ensure (1) the structural and leaktight integrity of its components, (2) the operability and performance of the system components, and (3) the operability of the system as a whole and, under conditions as close to design as practical, the performance of the full operational sequence that brings the system into operation, including operation of associated systems and interfaces with an ultimate heat sink and the transition from the active normal operation mode to the passive operation mode relied upon during postulated accidents, including the operation of applicable portions of the protection system and the operation of the associated structural and equipment cooling water system.

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

Section 2.3.4 of INL/EXT-10-17997, Mechanistic Source Terms White Paper, (Ref. 33) notes that the passive reactor cavity cooling system (RCCS) (using either air or water as heat transfer fluid) contributes to the mHTGR safety basis and is subject to component integrity testing. However, Section 6.1 of INL/EXT-11-22708, Modular HTGR Safety Basis and Approach, (Ref. 34), indicates that RCCS performance does not require leaktight conditions. For an RCCS which is an open system, the normal and expected loss of RCCS coolant through the exhaust structure would not be considered leakage. Abnormal leakage of RCCS coolant to locations other than the exhaust structure may be acceptable provided that (1) the RCCS leakage does not impact safety functions under all conditions, and (2) functional containment is not impacted by RCCS leakage.

APPENDIX C.

MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGN CRITERIA Appendix C to DG-1330, Page C-19 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.

Not applicable to mHTGR.

This criterion is not applicable to the mHTGR. The 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.

Not applicable to mHTGR.

This criterion is not applicable to the mHTGR. The 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.

Not applicable to mHTGR.

This criterion is not applicable to the mHTGR. The 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.

Not applicable to mHTGR.

This criterion is not applicable to the mHTGR. The mHTGR designs do not have a pressure retaining reactor containment

APPENDIX C.

MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGN CRITERIA Appendix C to DG-1330, Page C-20 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.

Not applicable to mHTGR.

This criterion is not applicable to the mHTGR. The 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.

Not applicable to mHTGR.

This criterion is not applicable to the mHTGR. The 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.

In addition to the heat rejection capability of the passive residual heat removal system, systems to transfer heat from structures, systems, and components important to safety to an ultimate heat sink shall be provided, as necessary, to transfer the combined heat load of these structures, systems, and components under normal operating and accident conditions.

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.

This mHTGR-DC accounts for advanced reactor design system differences to include cooling requirements for SSCs important to safety, if applicable; this mHTGR-DC does not address the residual heat removal system required under ARDC 34.

The staff inserted passive based on the system design for residual heat removal. If a specific mHTGR design can demonstrate that the reactor cavity cooling system (RCCS) provides indefinite core cooling capability, then structural and equipment cooling systems would not be needed.

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.

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

APPENDIX C.

MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGN CRITERIA Appendix C to DG-1330, Page C-21 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.

Same as ARDC The structural and equipment cooling systems shall be designed to permit appropriate periodic functional testing to assure (1) the structural and leaktight integrity of their components, (2) the operability and the performance of the system components, and (3) the operability of the systems as a whole and, under conditions as close to design as practical, the performance of the full operational sequences that bring the systems into operation for reactor shutdown and postulated accidents, including operation of associated systems.

This renamed mHTGR-DC accounts for advanced reactor system design differences to include possible cooling requirements for SSCs important to safety. 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) 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.

Not applicable to mHTGR.

This criterion is not applicable to the mHTGR. The 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.

MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGN CRITERIA Appendix C to DG-1330, Page C-22 V.

Reactor Containment Criterion mHTGR-DC Title and Content NRC Rationale for Adaptions to GDC 51 Fracture prevention of containment pressure boundary.

Not applicable to mHTGR.

This criterion is not applicable to the mHTGR. The 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.

Not applicable to mHTGR.

This criterion is not applicable to the mHTGR. The 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.

Not applicable to mHTGR.

This criterion is not applicable to the mHTGR. The 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.

Not applicable to mHTGR.

This criterion is not applicable to the mHTGR. The 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.

Not applicable to mHTGR.

This criterion is not applicable to the mHTGR. The 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.

Not applicable to mHTGR.

This criterion is not applicable to the mHTGR. The 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.

MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGN CRITERIA Appendix C to DG-1330, Page C-23 V.

Reactor Containment Criterion mHTGR-DC Title and Content NRC Rationale for Adaptions to GDC 57 Closed system isolation valves.

Not applicable to mHTGR.

This criterion is not applicable to the mHTGR. The 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.

Same as ARDC The fuel storage and handling, radioactive waste, and other systems which may contain radioactivity shall be designed to assure adequate safety under normal and postulated accident conditions. These systems shall be designed (1) with a capability to permit appropriate periodic inspection and testing of 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.

The underlying concept of establishing functional requirements for radioactivity control in fuel storage and fuel handling systems is independent of the design of non-LWR advanced reactors.

However, some advanced designs may use dry fuel storage that incorporates cooling jackets that can be liquid-cooled or air-cooled to remove heat. This modification to this GDC allows for both liquid and air-cooling of the dry fuel storage containers.

62 Prevention of criticality in fuel storage and handling.

Same as GDC 63 Monitoring fuel and waste storage.

Same as GDC

APPENDIX C.

MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGN CRITERIA Appendix C to DG-1330, Page C-24 VI.

Fuel and Reactivity Control Criterion mHTGR-DC Title and Content NRC Rationale for Adaptions to GDC 64 Monitoring radioactivity releases.

Means shall be provided for monitoring the reactor building atmosphere, effluent discharge paths, and plant environs for radioactivity that may be released from normal operations, including anticipated operational occurrences, and from postulated accidents.

The underlying concept of monitoring radioactivity releases from the mHTGR particle fuel to the reactor building, effluent discharge paths, and plant environs applies. High radioactivity in the reactor building provides input to the plant protection system. In addition, the reactor building atmosphere is monitored for personnel 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.

The design of the reactor vessel and reactor system shall be such that their integrity is maintained during postulated accidents (1) to ensure the geometry for passive removal of residual heat from the reactor core to the ultimate heat sink and (2) to permit sufficient insertion of the neutron absorbers to provide for reactor shutdown.

New mHTGR design-specific GDC are necessary to ensure that the reactor vessel and reactor system (including the fuel, reflector, control rods, core barrel, and structural supports) integrity is preserved for passive heat removal and for the insertion of neutron absorbers.

71 Reactor building design basis.

The design of the reactor building shall be such that, during postulated accidents, it structurally protects the geometry for passive removal of residual heat from the reactor core to the ultimate heat sink and provides a pathway for the release of reactor helium from the building in the event of depressurization accidents.

The reactor building functions are to protect and maintain passive cooling geometry and to provide a pathway for the release of helium from the building in the case of a line break in the reactor helium pressure boundary. This newly established criterion ensures that these safety functions are provided.

It is noted that the reactor building is not relied upon to meet the offsite dose requirements of 10 CFR 50.34 (10 CFR 52.79).

APPENDIX C.

MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGN CRITERIA Appendix C to DG-1330, Page C-25 VII. Additional mHTGR-DC Criterion mHTGR-DC Title and Content NRC Rationale for Adaptions to GDC 72 Provisions for periodic reactor building inspection.

The reactor building shall be designed to permit (1) appropriate periodic inspection of all important structural areas and the depressurization pathway, and (2) an appropriate surveillance program.

This newly established criterion on periodic inspection and surveillance provides assurance that the reactor building will perform its safety functions of protecting and maintaining the configuration needed for passive cooling and providing a discharge pathway for helium depressurization events.