Regulatory Guide 1.232
| ML18058B961 | |
| Person / Time | |
|---|---|
| Issue date: | 02/27/2018 |
| From: | Office of Nuclear Regulatory Research |
| To: | |
| Mazza J | |
| References | |
| DG-1330 RG-1.232 | |
| Download: ML18058B961 (134) | |
U.S. NUCLEAR REGULATORY COMMISSION
REGULATORY GUIDE RG 1.232, REVISION 0
Issue Date: March 2018 Technical Lead: Jan Mazza REGULATORY GUIDE 1.232 (Draft was issued as DG-1330 dated February 2017))
GUIDANCE FOR DEVELOPING PRINCIPAL DESIGN
CRITERIA FOR NON-LIGHT-WATER REACTORS
A. INTRODUCTION
Purpose This regulatory guide (RG) describes the Nuclear Regulatory Commissions (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) may be adapted for 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, for nuclear power plants. 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 nuclear power 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)1.
1 While the design criteria described in this RG were developed for nuclear power reactor applicants developing non- LWR designs, the design criteria described in this RG may be applied, as appropriate, to non-light-water non-power reactors.
Written suggestions regarding this guide or development of new guides may be submitted through the NRCs public Web site in the NRC Library at http://www.nrc.gov/reading-rm/doc-collections/, under Document Collections, in Regulatory Guides, at http://www.nrc.gov/reading-rm/doc-collections/reg-guides/contactus.html, Electronic copies of this RG, previous versions of this guide, and other recently issued guides are also available through the NRCs public Web site in the NRC Library at http://www.nrc.gov/reading-rm/doc-collections/, under Document Collections, in Regulatory Guides. This RG is also available through the NRCs Agencywide Documents Access and Management System (ADAMS) at http://www.nrc.gov/reading-rm/adams.html, under ADAMS Accession Number (No.) ML17325A611. The regulatory analysis may be found in ADAMS under Accession No. ML16330A179. The associated draft guide DG-1330 may be found in ADAMS under Accession No. ML16301A307, and the staff responses to the public comments on DG-1330 may be found under ADAMS Accession No. ML17325A616.
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.
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 RG 1.232, Rev. 0, Page 2
April 8, 1993, provides staff insights on issues pertaining to advanced designs and proposes resolutions (Ref. 7).
- 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 30, 1993, provides the Commission position on topics discussed in SECY-93-092 (Ref. 8).
- 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 PlantAssessment of Key Licensing Issues, dated July 17, 2014, provides the NRC staffs review and insights on the Next Generation Nuclear Plant MHTGR proposed licensing approach (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 provides guidance for implementing the mandatory information collections in 10 CFR Parts 50
and 52 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), under control numbers 3150-0011 and 3150-0151. Send comments regarding this information collection to the Information Services Branch, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by e-mail to Infocollects.Resource@nrc.gov, and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202, (3150-0011, 3150-0151) Office of Management and Budget, Washington, DC
20503.
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.
RG 1.232, Rev. 0, Page 3
CONTENTS
A. Introduction ............................................................................................................................................. 1 B. Discussion ............................................................................................................................................... 5 C. Staff Regulatory Guidance .................................................................................................................... 11 D. Implementation ..................................................................................................................................... 20
Acronyms .................................................................................................................................................... 22 References ................................................................................................................................................... 24 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 RG 1.232, Rev. 0, Page 4
B. DISCUSSION
Reason for Issuance This revision (Revision 0) provides guidance for developing PDC for non-LWRs. Applications for a construction permit, design certification, combined license, standard design approval, or manufacturing license are required by 10 CFR 50.34(a)(3)(i), 10 CFR 52.47(a)(3)(i),
10 CFR 52.79(a)(4)(i), 10 CFR 52.137(a)(3)(i), and 10 CFR 52.157(a), respectively, to include the PDC
for the facility in their applications.
Background The NRC Regulatory Framework In accordance with its mission, the NRC protects public health and safety and the environment by regulating the design, siting, construction, and operation of commercial nuclear power facilities. The NRC conducts its reactor licensing activities through a combination of regulatory requirements and guidance. The applicable regulatory requirements are found in Chapter I of Title 10, Energy, of the Code of Federal Regulations, Parts 1 through 199. Regulatory guidance is additional detailed information on specific acceptable means to meet the requirements in regulation. Guidance is provided in several forms, such as in RGs, interim staff guidance, standard review plans, NUREGs, review standards, and Commission policy statements. These regulatory requirements and guidance represent the entirety of the regulatory framework that an applicant should consider when preparing an application for review by the NRC. A key part of the regulatory requirements is in the general design criteria (GDC) in Appendix A to
10 CFR Part 50. These high-level GDC requirements support the design of the current nuclear power plants and are addressed in 10 CFR 50.34, Contents of Applications; Technical Information. Because the current GDC are based on LWR technology, the NRC developed the non-LWR design criteria, included as appendices to this RG, to provide guidance for developing PDC for non-LWR technology.
The nuclear power plants presently operating in the United States were licensed under the process described in 10 CFR Part 50. The NRC and its predecessor, the U. S. Atomic Energy Commission (AEC),
approved construction permits for these plants between 1964 and 1978 and granted the most recent operating license under 10 CFR Part 50 in 2015. The regulations in 10 CFR Part 50 evolved over the years to address specific safety issues discovered as a result of operating experience and industry events.
Some examples include fire protection in 10 CFR 50.48, emergency plans in 10 CFR 50.47, and aircraft impact assessment in 10 CFR 50.150. The NRC applied some of these new regulations retroactively to operating reactors while applying others only to new reactors.
The NRC used its experience in licensing 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 10 CFR Part 50 nuclear power RG 1.232, Rev. 0, Page 5
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 staff requirements memorandum (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. This RG is not intended to be an accompaniment to the aforementioned rulemaking.
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 intended safety functions in design basis events postulated to occur during normal operations, anticipated operational occurrences (AOOs), and postulated accidents. All production and utilization facilities licensed under 10 CFR Part 50, including both LWRs and non-LWRs, are required to describe PDC in their preliminary safety analysis report supporting a construction permit application as described in
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. On October 14, 2008, the Commission issued its most recent policy statement on advanced nuclear power 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 nuclear power 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.
RG 1.232, Rev. 0, Page 6
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 for light water nuclear power reactors are derived from the GDC in Appendix A
to 10 CFR Part 50.
Title 10 CFR 50.342 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 need to be adapted for non-LWR designs and that the GDC in 10 CFR 50 Appendix A are not regulatory requirements for non- LWR designs but provide guidance in establishing the PDC 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 future benefits to risk informing the non-LWR design criteria to the extent possible, depending on the design information and data available. The NRCs 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.
2 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).
RG 1.232, Rev. 0, Page 7
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 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 to 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 inclusive to align with the six technologies above. In addition to the technology-inclusive 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. 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.3 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:
3 The technology-specific design criteria were developed using available design information, previous NRC pre- application reviews of the design types, and more recent industry and DOE national laboratory initiatives in these technology areas (see Reference 17). It is the responsibility of the designer or applicant to provide and justify the PDC
for a specific design.
RG 1.232, Rev. 0, Page 8
- 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.
The NRC issued a draft version of design criteria for informal public comment titled, Public Comment Sought - Advanced Non-Light Water Reactor Design Criteria, 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). NRC used the informal public comments and discussions during the public meeting held on October
11, 2016 (Ref. 23), to develop DG-1330, Guidance for Developing Principal Design Criteria for Non- Light Water Reactors, NRC staff issued the draft RG on February, 2017 (Ref. 24), for a 60 day comment period. NRC staff received over 120 comments on DG-1330 (Ref. 25), and held a public meeting on August 24, 2017 to discuss topics that warranted additional public interaction (Ref. 26). The tables in Appendices A, B, and C of this RG represent the staffs final version of the design criteria that incorporates many of the 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 how 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 RG 1.232, Rev. 0, Page 9
accident would result in a loss of a coolable geometry such that multiple non-LWR design criteria would be violated.
- Safety design approach 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, one or more of the criteria from the SFR-DC or MHTGR-DC may be more applicable to a design or technology than the ARDC.
- MHTGR refers to the category of HTGRs that use the inherent high temperature characteristics of tristructural isotropic (TRISO) coated fuel particles, graphite moderator, and helium coolant, as well as passive heat removal from a low power density core with a relatively large height-to- diameter ratio within an uninsulated steel reactor vessel. The MHTGR is designed in such a way to ensure that during design basis events (including loss of forced cooling or loss of helium pressure conditions) radionuclides are retained at their source in the fuel and regulatory requirements for offsite dose are met at the exclusion area boundary.
- The SFR-DC and MHTGR-DC were developed because the designs were mature and the design features diverse for 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.
- Some of the concepts discussed in the RG are policy issues that may require NRC Commission review and approval. Examples are functional containment performance requirements and the use of specified acceptable system radionuclide release design limits in place of specified acceptable fuel design limits. The NRC has not had the opportunity to fully consider these as they are specific to non-LWR designs.
- Non-LWR designs should provide enhanced margins of safety when compared to LWRs. They may use simplified, passive, or other innovative design features to accomplish their safety and security functions.
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 IAEA also has a Coordinated Research Activity on MHTGR safety design criteria.
The NRC will continue to monitor and collaborate on these documents and consider using them to the extent practical in developing SFR design criteria. The NRC will follow its standard procedures for public participation in the development of future NRC documents that reference or endorse international standards.
RG 1.232, Rev. 0, Page 10
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 AC 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 PDC4. Since the GDC in 10 CFR 50 Appendix A are not regulatory requirements for non- LWR designs but provide guidance in establishing the PDC for non-LWR designs, 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 after considering the underlying safety basis for the criterion and evaluating the rationale for the adaptation described in this RG. For example, FHRs are molten salt reactors that use TRISO fuel, which is the same fuel used for MHTGR technologie
s. An FHR
designer could use the MHTGR-DC where appropriate for the design. Another example is the MSRs that use liquid fuel. An MSR designer may need to develop new PDC for liquid fuel and systems to support this design.
In each case, it is the responsibility of the designer or applicant 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, and how the PDC demonstrate adequate assurance of safety. In instances where a GDC or non- LWR design criterion (ARDC, SFR-DFC, and 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 is not intended to imply that they are 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.
4 While the design criteria described in this RG were developed for nuclear power reactor applicants developing non- LWR designs, the design criteria described in this RG may be applied, as appropriate, to non-light-water non-power reactors.
RG 1.232, Rev. 0, Page 11
Non-LWR Crosswalk Table The following table (Table 1) provides a summary and crosswalk between the LWR GDC
contained in 10 CFR Part 50 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 AC to this RG.
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 to or deviation from 10 CFR Part 50,
Appendix A)
Column 4SFR-DC Title/Status (showing conformity to or deviation from 10 CFR Part 50,
Appendix A)
Column 5MHTGR-DC Title/Status (showing conformity to or 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) for ARDCs, and SFR-DC
Section IV Heat Transport Systems (Criteria 30-46) for MHTGR-DC
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)
RG 1.232, Rev. 0, Page 12
TABLE 1: NON-LIGHT-WATER-REACTOR CROSSWALK
I. Overall Requirements Criterion Current GDC Title ARDC Title/Status SFR-DC Title/Status MHTGR-DC Title/Status
1 Quality standards and records. Same as GDC Same as GDC Same as GDC
2 Design bases for protection Same as GDC Same as GDC Same as GDC
against natural phenomena.
3 Fire protection. Fire protection. Same as ARDC Same as ARDC
Modified for ARDC
4 Environmental and dynamic Environmental and dynamic Environmental and dynamic Environmental and dynamic effects design bases. effects design bases. effects design bases. effects design bases.
Modified for ARDC Modified for SFR-DC Modified for MHTGR-DC
5 Sharing of structures, systems, Same as GDC Same as GDC Same as GDC
and components.
II. Multiple Barriers Criterion Current GDC Title ARDC Title/Status SFR-DC Title/Status MHTGR-DC Title/Status
10 Reactor design. Same as GDC Same as GDC Reactor design.
Modified for MHTGR-DC
11 Reactor inherent protection. Reactor inherent protection. Same as ARDC Same as ARDC
Modified for ARDC
12 Suppression of reactor Suppression of reactor power Same as ARDC Suppression of reactor power power oscillations. oscillations. oscillations.
Modified for ARDC Modified for MHTGR-DC
13 Instrumentation and control. Instrumentation and control. Instrumentation and control. Instrumentation and control.
Modified for ARDC Modified for SFR-DC Modified for MHTGR-DC
RG 1.232, Rev. 0, Page 13
TABLE 1: NON-LIGHT-WATER-REACTOR CROSSWALK
II. Multiple Barriers Criterion Current GDC Title ARDC Title/Status SFR-DC Title/Status MHTGR-DC Title/Status
14 Reactor coolant pressure Reactor coolant pressure Primary coolant pressure Reactor helium coolant boundary. boundary. boundary. pressure boundary.
Modified for ARDC Modified for SFR-DC Modified for MHTGR-DC
15 Reactor coolant system Reactor coolant system design. Primary Reactor coolant Reactor helium pressure design. Modified for ARDC system design. boundary coolant design.
Modified for SFR-DC Modified for MHTGR-DC
16 Containment design. Same as GDC Containment design. Containment design.
Modified for SFR-DC Modified for MHTGR-DC
17 Electric power systems. Electric power systems. Same as ARDC Electric power systems.
Modified for ARDC Modified for MHTGR-DC
18 Inspection and testing of Inspection and testing of electric Same as ARDC Same as ARDC
electric power systems. power systems.
Modified for ARDC
19 Control room. Control room. Control roo
m. Same as ARDC
Modified for ARDC Modified for SFR-DC
III. Reactivity Control Criterion Current GDC Title ARDC Title/Status SFR-DC Title/Status MHTGR-DC Title/Status
20 Protection system functions. Same as GDC Same as GDC Protection system functions.
Modified for MHTGR-DC
21 Protection system reliability Same as GDC Same as GDC Same as GDC
and testability.
22 Protection system Same as GDC Same as GDC Same as GDC
independence.
RG 1.232, Rev. 0, Page 14
TABLE 1: NON-LIGHT-WATER-REACTOR CROSSWALK
III. Reactivity Control Criterion Current GDC Title ARDC Title/Status SFR-DC Title/Status MHTGR-DC Title/Status
23 Protection system failure Same as GDC Protection system failure Same as GDC
modes. modes.
Modified for SFR-DC
24 Separation of protection and Same as GDC Same as GDC Same as GDC
control systems.
25 Protection system Protection system requirements for Same as ARDC Protection system requirements for reactivity reactivity control malfunctions. requirements for reactivity control malfunctions. Modified for ARDC control malfunctions.
Modified for MHTGR-DC
26 Reactivity control system Reactivity control systems Same as ARDC Reactivity control systems redundancy and capability. redundancy and capacity Modified for MHTGR-DC
Modified for ARDC
27 Combined reactivity control Combined reactivity control Same as ARDC Same as ARDC
systems capability systems capability DELETED and incorporated into ARDC 26
28 Reactivity limits. Reactivity limits. Reactivity limits. Reactivity limits.
Modified for ARDC Modified for SFR-DC Modified for MHTGR-DC
29 Protection against Same as GDC Same as GDC Same as GDC
anticipated operational occurrences.
RG 1.232, Rev. 0, Page 15
TABLE 1: NON-LIGHT-WATER-REACTOR CROSSWALK
IV. Fluid Systems (Heat Transport Systems for MHTGRs)
Criterion Current GDC Title ARDC Title/Status SFR-DC Title/Status MHTGR-DC Title/Status
30 Quality of reactor coolant Quality of reactor coolant Quality of reactor primary Quality of reactor helium coolant pressure boundary. pressure boundary. coolant pressure boundary. pressure boundary.
Modified for ARDC Modified for SFR-DC Modified for MHTGR-DC
31 Fracture prevention of reactor Fracture prevention of reactor Fracture prevention of reactor Fracture prevention of reactor coolant pressure boundary. coolant pressure boundary. primary coolant pressure helium coolant pressure Modified for ARDC boundary. boundary.
Modified for SFR-DC Modified for MHTGR-DC
32 Inspection of reactor coolant Inspection of reactor coolant Inspection of reactor primary Inspection of reactor helium pressure boundary. pressure boundary. coolant pressure boundary. coolant pressure boundary.
Modified for ARDC Modified for SFR-DC Modified for MHTGR-DC
33 Reactor coolant makeup. Reactor coolant inventory Reactor Primary coolant Not applicable to MHTGR.
maintenance makeup. inventory maintenance Modified for ARDC makeup.
Modified for SFR-DC
34 Residual heat removal. Residual heat removal. Residual heat removal. Residual heat removal. Modified Modified for ARDC Modified for SFR-DC for MHTGR-DC
Emergency core cooling. Same as ARDC Not applicable to MHTGR.
35 Emergency core cooling.
Modified for ARDC
Inspection of passive emergency Inspection of emergency core Inspection of emergency core core cooling residual heat
36 cooling syste
m. Same as ARDC
cooling system. removal system.
Modified for ARDC
Modified for MHTGR-DC
Testing of passive residual heat Testing of emergency core Testing of emergency core removal emergency core cooling
37 cooling syste
m. Same as ARDC
cooling system. system.
Modified for ARDC
Modified for MHTGR-DC
Containment heat removal.
38 Containment heat removal. Same as ARDC Not applicable to MHTGR.
Modified for ARDC
RG 1.232, Rev. 0, Page 16
TABLE 1: NON-LIGHT-WATER-REACTOR CROSSWALK
IV. Fluid Systems (Heat Transport Systems for MHTGRs)
Criterion Current GDC Title ARDC Title/Status SFR-DC Title/Status MHTGR-DC Title/Status Inspection of containment heat Inspection of containment heat
39 removal system. Same as ARDC Not applicable to MHTGR.
removal system.
Modified for ARDC
Testing of containment heat Testing of containment heat
40 removal system. Same as ARDC Not applicable to MHTGR.
removal system.
Modified for ARDC
Containment atmosphere Containment atmosphere
41 cleanup. Same as ARDC Not applicable to MHTGR.
cleanup.
Modified for ARDC
Inspection of containment Same as GDC Same as GDC
42 Not applicable to MHTGR.
atmosphere cleanup systems.
Testing of containment Testing of containment
43 atmosphere cleanup systems. Same as ARDC Not applicable to MHTGR.
atmosphere cleanup systems.
Modified for ARDC
Structural and equipment Structural and equipment
44 Cooling water. cooling. Cooling water Same as ARDC cooling. Cooling water Modified for ARDC Modified for MHTGR-DC
Inspection of structural and Inspection of cooling water equipment cooling water
45 Same as ARDC Same as ARDC
system. systems.
Modified for ARDC
Testing of structural and Testing of cooling water equipment cooling water
46 Same as ARDC Same as ARDC
system. systems.
Modified for ARDC
RG 1.232, Rev. 0, Page 17
TABLE 1: NON-LIGHT-WATER-REACTOR CROSSWALK
V. Reactor Containment Criterion Current GDC Title ARDC Title/Status SFR-DC Title/Status MHTGR-DC Title/Status Containment design basis. Containment design basis.
50 Containment design basis. Not applicable to MHTGR.
Modified for ARDC Modified for SFR-DC
Fracture prevention of Fracture prevention of Fracture prevention of containment pressure
51 containment pressure boundary. Not applicable to MHTGR.
containment pressure boundary. boundary.
Modified for ARDC
Modified for SFR-DC
Capability for containment Capability for containment Capability for containment
52 leakage rate testing. leakage rate testing. Not applicable to MHTGR.
leakage rate testing.
Modified for ARDC Modified for SFR-DC
Provisions for containment Provisions for containment Provisions for containment
53 testing and inspection. testing and inspection. Not applicable to MHTGR.
testing and inspection.
Modified for ARDC Modified for SFR-DC
Piping systems penetrating Piping systems penetrating Piping systems penetrating
54 containment. containment. Not applicable to MHTGR.
containment.
Modified for ARDC Modified for SFR-DC
Reactor coolant pressure Reactor Primary coolant Reactor coolant pressure boundary penetrating pressure boundary penetrating
55 boundary penetrating Not applicable to MHTGR.
containment. containment.
containment.
Modified for ARDC Modified for SFR-DC
Primary Containment Primary Containment isolation.
56 Primary containment isolation. isolation. Not applicable to MHTGR.
Modified for ARDC
Modified for SFR-DC
Closed system isolation valves. Closed system isolation valves.
57 Closed system isolation valves. Not applicable to MHTGR.
Modified for ARDC Modified for SFR-DC
RG 1.232, Rev. 0, Page 18
TABLE 1: NON-LIGHT-WATER-REACTOR CROSSWALK
VI. Fuel and Radioactivity Control Criterion Current GDC Title ARDC Title/Status SFR-DC Title/Status MHTGR-DC Title/Status Control of releases of Same as GDC Same as GDC Same as GDC
60 radioactive materials to the environment.
Fuel storage and handling and Same as ARDC Same as ARDC
Fuel storage and handling and
61 radioactivity control.
radioactivity control.
Modified for ARDC
Prevention of criticality in fuel Same as GDC Same as GDC Same as GDC
62 storage and handling.
Monitoring fuel and waste Same as GDC Same as GDC Same as GDC
63 storage.
Monitoring radioactivity Monitoring radioactivity Monitoring radioactivity Monitoring radioactivity
64 releases. releases. releases.
releases.
Modified for ARDC Modified for SFR-DC Modified for MHTGR-DC
RG 1.232, Rev. 0, Page 19
TABLE 1: NON-LIGHT-WATER-REACTOR CROSSWALK
VII. Additional Technology-Specific Design Criteria Criterion Current GDC Title ARDC Title/Status SFR-DC Title/Status MHTGR-DC Title/Status Reactor vessel and reactor
70 N/A N/A Intermediate coolant system.
system structural design basis.
Primary coolant and cover gas
71 N/A N/A Reactor building design basis.
purity control.
Provisions for periodic reactor
72 N/A N/A Sodium heating systems.
building inspection.
Sodium leakage detection and
73 N/A N/A reaction prevention and N/A
mitigation.
Sodium/water reaction
74 N/A N/A N/A
prevention/mitigation.
Quality of the intermediate
75 N/A N/A N/A
coolant boundary.
Fracture prevention of the
76 N/A N/A N/A
intermediate coolant boundary.
Inspection of the intermediate
77 N/A N/A N/A
coolant boundary.
Primary coolant system
78 N/A N/A N/A
interfaces.
Cover gas inventory
79 N/A N/A N/A
maintenance.
RG 1.232, Rev. 0, Page 20
D. IMPLEMENTATION
The purpose of this section is to provide information on how applicants and licensees5 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 voluntarily6 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
5 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
6 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.
RG 1.232, Rev. 0, Page 20
underlying NRC regulatory requirements. This is not considered backfitting as defined in
10 CFR 50.109(a)(1) or a violation of any of the issue finality provisions in 10 CFR Part 52.
Additionally, an existing applicant may be required to comply with new rules, orders, or guidance if 10 CFR 50.109(a)(3) applies.
If a licensee believes that the NRC is either using this RG or requesting or requiring the licensee to implement the methods or processes in this RG in a manner inconsistent with the discussion in this Implementation section, then the licensee may file a backfit appeal with the NRC in accordance with the guidance in NUREG-1409, Backfitting Guidelines (Ref. 27), and the NRC Management Directive 8.4, Management of Facility-Specific Backfitting and Information Collection (Ref. 28).
RG 1.232, Rev. 0, Page 21
ACRONYMS/ABBREVIATIONS
AEC Atomic Energy Commission 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 ECCS emergency core cooling system FAUNA Forschungsanlage zur Untersuchung nuklearer Aerosole (Research Facility for Investigating Nuclear Aerosols)
FHR fluoride high-temperature reactors GCR gas-cooled fast reactors GDC general design criterion HTGR high-temperature gas-cooled reactor IAEA International Atomic Energy Agency LOCA loss of coolant accident 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 OMB Office of Management and Budget PDC principal design criteria PRISM Power Reactor Innovative Small Module RCCS reactor cavity cooling system RCPB reactor coolant pressure boundary RG regulatory guide SAFR Sodium Advanced Fast Reactor SARRDL specified acceptable system radionuclide release design limit SFR sodium-cooled fast reactor SFR-DC SFR design criteria SRM staff requirements memorandum SSC structure, system, and component RG 1.232, Rev. 0, Page 22
TRISO tristructural isotropic fuel RG 1.232, Rev. 0, Page 23
REFERENCES7
1. U.S. Code of Federal Regulations, Domestic Licensing of Production and Utilization Facilities, Part 50, Chapter 1, Title 10, Energy. (10 CFR Part 50)
2. U.S. Code of Federal Regulations, Licenses, Certifications, and Approvals for Nuclear Power Plants, Part 52, Title 10, Energy. (10 CFR Part 52)
3. U.S. Nuclear Regulatory Commission, 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. U.S. Nuclear Regulatory Commission, NUREG-1368, Preapplication Safety Evaluation Report for the Power Reactor Innovative Small Module (PRISM) Liquid-Metal Reactor, February 1994.
(ADAMS Accession No. ML063410561).
5. U.S. Nuclear Regulatory Commission, NUREG-0968, Safety Evaluation Report Related to the Construction of the Clinch River Breeder Reactor Plant, March 1983. (ADAMS Accession No. ML082381008).
6. U.S. Nuclear Regulatory Commission, NUREG-1369, Preapplication Safety Evaluation Report for the Sodium Advanced Fast Reactor (SAFR) Liquid-Metal Reactor, December 1991.
(ADAMS Accession No. ML063410547).
7. U.S. Nuclear Regulatory Commission, 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. U.S. Nuclear Regulatory Commission, 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. U.S. Nuclear Regulatory Commission, SECY-03-0047, Policy Issues Related to Licensing Non- Light-Water Reactor Designs, March 2003. (ADAMS Accession No. ML030160002).
10. U.S. Nuclear Regulatory Commission, SRM-SECY-03-0047, Policy Issues Related to Licensing Non-Light-Water Reactor Designs, June 2003. (ADAMS Accession No. ML031770124).
11. U.S. Nuclear Regulatory Commission, Next Generation Nuclear Plant Assessment of Key Licensing Issues, July 17, 2014. (ADAMS Accession Nos. ML14174A734, ML14174A774 (Enclosure 1), and ML14174A845 (Enclosure 2)).
7 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.
RG 1.232, Rev. 0, Page 24
12. U.S. Nuclear Regulatory Commission, Policy Statement on the Regulation of Advanced Reactors (73 FR 60612), October 14, 2008. (ADAMS Accession No. ML082750370).
13. U.S. Nuclear Regulatory Commission, Policy Statement on Severe Reactor Accidents Regarding Future Designs and Existing Plants, August 1985. (ADAMS Accession No. ML003711521).
14. U.S. Nuclear Regulatory Commission, 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. U.S. Nuclear Regulatory Commission, 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. U.S. Nuclear Regulatory Commission, NRC Vision and Strategy: Safely Achieving Effective and Efficient Non-Light Water Reactor Mission Readiness, December 2016. (ADAMS
Accession No. ML16356A670).
17. U.S. Department of Energy, Guidance for Developing Principal Design Criteria for Advanced (Non-Light Water) Reactors, December 2014. (ADAMS Accession Nos. ML14353A246, ML14353A248)8.
18. U.S. Nuclear Regulatory Commission, 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. U.S. Nuclear Regulatory Commission, 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. U.S. Nuclear Regulatory Commission, 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. U.S. Nuclear Regulatory Commission, Public Comment Sought - Advanced Non-Light Water Reactor Design Criteria, April 2016. (ADAMS Accession No. ML16096A420).
22. U.S. Nuclear Regulatory Commission, Non-LWR Design Criteria Public Comments, June 2016 (ML17011A116).
8 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.
RG 1.232, Rev. 0, Page 25
23. U.S. Nuclear Regulatory Commission, Summary of October 11, 2016 Public Meeting Regarding Non-Light Water Reactor Design Criteria. (ADAMS Accession No. ML16314B333).
24. U.S. Nuclear Regulatory Commission, DG-1330, Guidance for Developing Principal Design Criteria for Non-Light Water Reactors, February 2017. (ADAMS Accession No.
25. U.S. Nuclear Regulatory Commission, Advanced Reactor Design Criteria - Public Comment Table, August 2017 2017. (ADAMS Accession No. ML17227A146).
26. U.S. Nuclear Regulatory Commission, Summary of August 24, 2017, Public Meeting on Advanced Non-Light Water Reactor Design Criteria, October, 2017. (ADAMS Accession No.
27. U.S. Nuclear Regulatory Commission, NUREG-1409, Backfitting Guidelines, July 1990.
(ADAMS Accession No. ML032230247).
28. U.S. Nuclear Regulatory Commission, Management Directive 8.4, Management of Facility- Specific Backfitting and Information Collection, October 2013. (ADAMS Accession No. ML12059A460).
29. U.S. Nuclear Regulatory Commission, Response to Gap Analysis Summary Report for Reactor System Issues, September 2016. (ADAMS Accession No. ML16116A083).
30. U.S. Nuclear Regulatory Commission, Response to NuScale Gap Analysis Summary Report for Reactivity Control Systems, Addressing Gap 11, General Design Criteria 26, December 2016.
(ADAMS Accession No. ML16292A589).
31. U.S. Atomic Energy Commission, Proposed General Design Criteria of 1965, AEC-R 2/49, November 5, 1967. (32 FR 10216).
32. U.S. Nuclear Regulatory Commission, 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).
33. Nuclear Energy Agency, Experimental Facilities for Sodium Fast Reactor Safety Studies, Task Group on Advanced Reactors Experimental Facilities (TAREF), 2011, pp. 22 and 54. Available on-line: https://www.oecd-nea.org/globalsearch/download.php?doc=77089 9
34. International Atomic Energy Agency, 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
9 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.
RG 1.232, Rev. 0, Page 26
May 1, 2013, p. 57. Available online: https://www.gen-
4.org/gif/upload/docs/application/pdf/2017-07/sdc_report_2may2013.pdf 10.
35. 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 online:
https://www.iaea.org/NuclearPower/Downloadable/Meetings/2015/2015-06-29-07-03-NPTDS-
mexico/2-3-_IAEAseminarMexicoCity_TSofu_SFRTechnologyOverview.pdf 11.
36. 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 online: http://www.sciencedirect.com/science/article/pii/S030645491500580012.
37. DOE, Idaho National Laboratories (INL), Mechanistic Source Terms White Paper, INL/EXT-
10-17997, Rev.0, July 2010, (ADAMS Accession No. ML102040260).
38. DOE. INL, Modular HTGR Safety Basis and Approach, Idaho National Laboratory, INL/EXT-
11-22708, Rev.0, August 2011, (ADAMS Accession No. ML11251A169).
39. U.S. Nuclear Regulatory Commission, RG 1.26, Quality Group Classifications and Standards for Water-, Steam-, and Radioactive-Waste-Containing Components of Nuclear Power Plants, February 2017. (ADAMS Accession No. ML16082A501).
10 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.
11 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.
12 Copies of Annals of Nuclear Energy Articles may be obtained through the Science Direct Web site:
WWW.ScienceDirect.com.
RG 1.232, Rev. 0, Page 27
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 (e.g., 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 results of this review are presented in the table below, which has three columns:
Column 1Criterion Number Column 2ARDC Title and Content Column 3NRC Rationale for Adaptations to GDC
The table is further divided into six sections similar to 10 CFR Part 50, Appendix A:
Section IOverall Requirements (Criterion 1 - 15)
Section II Multiple Barriers (Criterion 10 - 20)
Section III Reactivity Control (Criterion 21 - 29)
Section IV Fluid Systems (Criterion 30 - 46)
Section V Reactor Containment (Criterion 50 - 57)
Section VI Fuel and Radioactivity Control (Criterion 60 - 64)
Appendix A to RG 1.232, Rev. 0, Page A-1
APPENDIX A. ADVANCED REACTOR DESIGN CRITERIA
I. Overall Requirements Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC
1 Quality standards and records.
Same as GDC
Structures, systems, and components important to safety shall be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed. Where generally recognized codes and standards are used, they shall be identified and evaluated to determine their applicability, adequacy, and sufficiency and shall be supplemented or modified as necessary to assure a quality product in keeping with the required safety function. A quality assurance program shall be established and implemented in order to provide adequate assurance that these structures, systems, and components will satisfactorily perform their safety functions. Appropriate records of the design, fabrication, erection, and testing of structures, systems, and components important to safety shall be maintained by or under the control of the nuclear power unit licensee throughout the life of the unit.
2 Design bases for protection against natural phenomena.
Same as GDC
Structures, systems, and components important to safety shall be designed to withstand the effects of natural phenomena such as earthquakes, tornadoes, hurricanes, floods, tsunami, and seiches without loss of capability to perform their safety functions. The design bases for these structures, systems, and components shall reflect: (1) Appropriate consideration of the most severe of the natural phenomena that have been historically reported for the site and surrounding area, with sufficient margin for the limited accuracy, quantity, and period of time in which the historical data have been accumulated, (2)
appropriate combinations of the effects of normal and accident conditions with the effects of the natural phenomena and (3) the importance of the safety functions to be performed.
Appendix A to RG 1.232, Rev. 0, Page A-2
APPENDIX A. ADVANCED REACTOR DESIGN CRITERIA
I. Overall Requirements Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC
3 Fire protection. The phrase containing examples where noncombustible and heat- Structures, systems, and components important to safety shall resistant materials must be used has been broadened to apply to all be designed and located to minimize, consistent with other advanced reactor designs.
safety requirements, the probability and effect of fires and explosions. Noncombustible and fire- resistant materials shall Instead of and, the phrase locations with structures, systems, and be used wherever practical throughout the unit, particularly in components (SSCs) important to safety uses or, which is locations with structures, systems, or components important to logically correct in this case.
safety. Fire detection and fighting systems of appropriate capacity and capability shall be provided and designed to minimize the adverse effects of fires on structures, systems, and components important to safety. Firefighting systems shall be designed to ensure that their rupture or inadvertent operation does not significantly impair the safety capability of these structures, systems, and components.
4 Environmental and dynamic effects design bases. This change removes the light-water reactor (LWR) emphasis on Structures, systems, and components important to safety shall loss-of-coolant accidents (LOCAs) that may not apply to every be designed to accommodate the effects of and to be compatible design. For example, helium is not needed in a MHTGR to remove with the environmental conditions associated with normal heat from the core during postulated accidents and does not have the operation, maintenance, testing, and postulated accidents. These same importance as water does to LWR designs to ensure that fuel structures, systems, and components, shall be appropriately integrity is maintained. Therefore, a specific reference to LOCAs is protected against dynamic effects, including the effects of not applicable to all designs. LOCAs may still require analysis in missiles, pipe whipping, and discharging fluids, that may result conjunction with postulated accidents if relevant to the design.
from equipment failures and from events and conditions outside the nuclear power unit. However, dynamic effects associated Reference to pipe whip may not be applicable to designs that with postulated pipe ruptures in nuclear power units may be operate at low pressure.
excluded from the design basis when analyses reviewed and approved by the Commission demonstrate that the probability of 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
Appendix A to RG 1.232, Rev. 0, Page A-3
APPENDIX A. ADVANCED REACTOR DESIGN CRITERIA
I. Overall Requirements Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC
Structures, systems, and components important to safety shall not be shared among nuclear power units unless it can be shown that such sharing will not significantly impair their ability to perform their safety functions, including, in the event of an accident in one unit, an orderly shutdown and cooldown of the remaining units.
II. Multiple Barriers Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC
10 Reactor design.
Same as GDC
The reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.
11 Reactor inherent protection. The wording has been changed to broaden the applicability from The reactor core and associated systems that contribute to coolant systems to additional factors (including structures or other reactivity feedback shall be designed so that, in the power fluids) that may contribute to reactivity feedback. These systems are operating range, the net effect of the prompt inherent nuclear to be designed to compensate for rapid reactivity increase.
feedback characteristics tends to compensate for a rapid increase in reactivity.
12 Suppression of reactor power oscillations. The word structures was added because items such as reflectors, The reactor core; associated structures; and associated coolant, which could be considered either outside or not part of the reactor control, and protection systems shall be designed to ensure that core, may affect susceptibility of the core to power oscillations.
power oscillations that can result in conditions exceeding specified acceptable fuel design limits are not possible or can be reliably and readily detected and suppressed.
Appendix A to RG 1.232, Rev. 0, Page A-4
APPENDIX A. ADVANCED REACTOR DESIGN CRITERIA
II. Multiple Barriers Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC
13 Instrumentation and control. Reactor coolant pressure boundary has been relabeled as reactor Instrumentation shall be provided to monitor variables and coolant boundary to create a more broadly applicable non-LWR
systems over their anticipated ranges for normal operation, for term that defines the boundary without giving any implication of anticipated operational occurrences, and for accident conditions, system operating pressure. As such, the term "reactor coolant as appropriate to ensure adequate safety, including those boundary" is applicable to non-LWRs that operate at either low or variables and systems that can affect the fission process, the high pressure.
integrity of the reactor core, the reactor coolant boundary, and the containment and its associated systems. Appropriate controls shall be provided to maintain these variables and systems within prescribed operating ranges.
14 Reactor coolant boundary. Reactor coolant pressure boundary has been relabeled as reactor The reactor coolant boundary shall be designed, fabricated, coolant boundary to create a more broadly applicable non-LWR
erected, and tested so as to have an extremely low probability of term that defines the boundary without giving any implication of abnormal leakage, of rapidly propagating failure, and of gross system operating pressure. As such, the term reactor coolant rupture. boundary is applicable to non-LWRs that operate at either low or high pressure.
15 Reactor coolant system design. Reactor coolant pressure boundary has been relabeled as reactor The reactor coolant system and associated auxiliary, control, coolant boundary to create a more broadly applicable non-LWR
and protection systems shall be designed with sufficient margin term that defines the boundary without giving any implication of to ensure that the design conditions of the reactor coolant system operating pressure. As such, the term "reactor coolant boundary are not exceeded during any condition of normal boundary" is applicable to non-LWRs that operate at either low or operation, including anticipated operational occurrences. high pressure.
16 Containment design. For non-LWR technologies other than SFRs and MHTGRs, Same as GDC designers may use the current GDC to develop applicable principal Reactor containment and associated systems shall be provided design criteria. The assumed degree of leak tightness for a to establish an essentially leak-tight barrier against the containment is used within safety analyses and plant performance uncontrolled release of radioactivity to the environment and to requirements to confirm onsite and offsite doses are below limits as assure that the containment design conditions important to specified in 10 CFR 50.34. It is also recognized that characteristics safety are not exceeded for as long as postulated accident of the coolants, fuels, and containments to be used in non-LWR
conditions require. designs could share common features with SFRs and MHTGRs.
Hence designers may propose using the SFR-DC-16 or Appendix A to RG 1.232, Rev. 0, Page A-5
APPENDIX A. ADVANCED REACTOR DESIGN CRITERIA
II. Multiple Barriers Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC
MHTGR-DC 16 as appropriate. Use of the MHTGR-DC 16 will be subject to a policy decision by the Commission.
17 Electric power systems. The electric power systems are required to provide reliable power Electric power systems shall be provided when required to for SSCs during anticipated operational occurrences or postulated permit functioning of structures, systems, and components. The accident conditions when those SSCs safety functions require safety function for each power system shall be to provide electric power. The safety functions are established by the safety sufficient capacity and capability to ensure that (1) that the analyses (i.e. design basis accidents). Where electric power is design limits for the fission product barriers are not exceeded as needed for anticipated operational occurrences or postulated a result of anticipated operational occurrences and (2) safety accidents, the electric power systems shall be sufficient in capacity functions that rely on electric power are maintained in the event and capability to ensure that safety functions as well as important to of postulated accidents. safety functions are maintained. The electric power systems provide redundancy and defense-in-depth since there would be a minimum The electric power systems shall include an onsite power system of two power systems.
and an additional power system. The onsite electric power system shall have sufficient independence, redundancy, and Compared to GDC 17, more emphasis is placed herein on requiring testability to perform its safety functions, assuming a single reliability of the overall power supply scheme rather than fully failure. An additional power system shall have sufficient prescribing how such reliability can be attained. For example, independence and testability to perform its safety function. reference to offsite electric power systems was deleted to provide for those reactor designs that do not depend on offsite power for the If electric power is not needed for anticipated operational functioning of SSCs important to safety or do not connect to a occurrences or postulated accidents, the design shall power grid.
demonstrate that power for important to safety functions is provided. The onsite power system is envisioned as a fully Class 1E power system and the additional power system is left to the discretion of the designer as long as it meets the performance criteria in paragraph one and the design criteria of paragraph two. For example, the additional independent power source could be from the electrical grid, a diesel generator, a combustion gas turbine or some other alternative, again, at the discretion of the designer.
In this context, important to safety functions refer to the broader, potentially non-safety related functions such as post-accident Appendix A to RG 1.232, Rev. 0, Page A-6
APPENDIX A. ADVANCED REACTOR DESIGN CRITERIA
II. Multiple Barriers Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC
monitoring, control room habitability, emergency lighting, radiation monitoring, communications and/or any others that may be deemed appropriate for the given design. The electric power system for important to safety functions could be non-Class 1E and would not be required to have redundant power sources.
18 Inspection and testing of electric power systems. ARDC 18 is a design-independent companion criterion to Electric power systems important to safety shall be designed to ARDC 17.
permit appropriate periodic inspection and testing of important areas and features, such as wiring, insulation, connections, and Wording pertaining to additional system examples has been deleted switchboards, to assess the continuity of the systems and the to allow increased flexibility associated with various designs.
condition of their components. The systems shall be designed Specifically, the text related to the nuclear power unit, offsite power with a capability to test periodically (1) the operability and system, and onsite power system was deleted to be consistent with functional performance of the components of the systems, such ARDC 17.
as onsite power sources, relays, switches, and buses, and (2) the operability of the systems as a whole and, under conditions as close to design as practical, the full operation sequence that brings the systems into operation, including operation of applicable portions of the protection system, and the transfer of power among systems.
19 Control room. ARDC 19 preserves the language of GDC 19 which states (with A control room shall be provided from which actions can be emphasis added) A control room shall be provided from which taken to operate the nuclear power unit safely under normal actions can be taken to operate the nuclear power unit safely conditions and to maintain it in a safe condition under accident However some clarification of this language is warranted.
conditions. Adequate radiation protection shall be provided to permit access and occupancy of the control room under accident It is clear from this language that there is a need to for operators to conditions without personnel receiving radiation exposures in be able to take actions to control the plant. Therefore, designers excess of 5 rem total effective dose equivalent as defined in must consider how the design of controls support safe operator
§ 50.2 for the duration of the accident. actions. In addition, NRC staff recognizes that in order for operators to act safely as stated in ARDC, that operators must Adequate habitability measures shall be provided to permit have certain knowledge about the status of the plant and be able to access and occupancy of the control room during normal make decisions about the appropriate course of action given a operations and under accident conditions. Equipment at particular operating circumstance. Therefore, these cognitive needs Appendix A to RG 1.232, Rev. 0, Page A-7
APPENDIX A. ADVANCED REACTOR DESIGN CRITERIA
II. Multiple Barriers Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC
appropriate locations outside the control room shall be provided of operators should also be considered by designers when
(1) with a design capability for prompt hot shutdown of the interpreting ARDC 19.
reactor, including necessary instrumentation and controls to maintain the unit in a safe condition during hot shutdown, and This consideration should be reflected in the design of indications,
(2) with a potential capability for subsequent cold shutdown of displays, alarms, controls or other future technologies which are the reactor through the use of suitable procedures. used to inform operators of plant status and may be used to support the decision making process (such as computer based procedures).
This position is consistent with 10 CFR 50.34(f)(2)(iii) which describes the contents required in applications for construction permits. Amongst many other requirements, this rule indicates that the control room design must reflect state-of-the-art human factors principles. These state-of-the-art principles inherently consider both the cognitive and physical aspects of operator action as described above.
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.
An additional control room habitability requirement has been proposed. It addresses a new concern: accidents that are not radiological in nature 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 to RG 1.232, Rev. 0, Page A-8
APPENDIX A. ADVANCED REACTOR DESIGN CRITERIA
III. Reactivity Control Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC
20 Protection system functions.
Same as GDC
The protection system shall be designed (1) to initiate automatically the operation of appropriate systems including the reactivity control systems, to assure that specified acceptable fuel design limits are 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.
21 Protection system reliability and testability.
Same as GDC
The protection system shall be designed for high functional reliability and inservice testability commensurate with the safety functions to be performed. Redundancy and independence designed into the protection system shall be sufficient to assure that (1) no single failure results in loss of the protection function and (2) removal from service of any component or channel does not result in loss of the required minimum redundancy unless the acceptable reliability of operation of the protection system can be otherwise demonstrated. The protection system shall be designed to permit periodic testing of its functioning when the reactor is in operation, including a capability to test channels independently to determine failures and losses of redundancy that may have occurred.
22 Protection system independence.
Same as GDC
The protection system shall be designed to assure that the effects of natural phenomena, and of normal operating, maintenance, testing, and postulated accident conditions on redundant channels do not result in loss of the protection function, or shall be demonstrated to be acceptable on some other defined basis. Design techniques, such as functional Appendix A to RG 1.232, Rev. 0, Page A-9
APPENDIX A. ADVANCED REACTOR DESIGN CRITERIA
III. Reactivity Control Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC
diversity or diversity in component design and principles of operation, shall be used to the extent practical to prevent loss of the protection function.
23 Protection system failure modes.
Same as GDC
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, pressure, steam, water, and radiation) are experienced.
24 Separation of protection and control systems.
Same as GDC
The protection system shall be separated from control systems to the extent that failure of any single control system component or channel, or failure or removal from service of any single protection system component or channel which is common to the control and protection systems leaves intact a system satisfying all reliability, redundancy, and independence requirements of the protection system. Interconnection of the protection and control systems shall be limited so as to assure that safety is not significantly impaired.
25 Protection system requirements for reactivity control Text has been added to clarify that the protection system is designed malfunctions. to protect the specified acceptable fuel design limits for anticipated The protection system shall be designed to ensure that specified operational occurrences (AOOs) in combination with a single acceptable fuel design limits are not exceeded during any failure; the protection system does not have to protect the specified anticipated operational occurrence accounting for a single acceptable fuel design limits during a postulated accident in malfunction of the reactivity control systems. combination with a single failure. The example was deleted to make the ARDC technology inclusive.
Appendix A to RG 1.232, Rev. 0, Page A-10
APPENDIX A. ADVANCED REACTOR DESIGN CRITERIA
III. Reactivity Control Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC
26 Reactivity control systems. Recent licensing activity, associated with the application of GDC 26 A minimum of two reactivity control systems or means shall and GDC 27 to new reactor designs (ADAMS Accession Nos.
provide: ML16116A083 (Ref. 29) and ML16292A589) (Ref. 30), revealed that additional clarity could be provided in the area of reactivity
(1) A means of inserting negative reactivity at a sufficient rate control requirements. ARDC 26 combines the scope of GDC 26 and amount to assure, with appropriate margin for malfunctions, and GDC 27. The development of ARDC 26 is informed by the that the design limits for the fission product barriers are not proposed general design criteria of 1965 (AEC-R 2/49, November exceeded and safe shutdown is achieved and maintained 5), 1967 (32 FR 10216) (Ref. 31), current GDC 26 and 27, the during normal operation, including anticipated operational definition of safety-related SSC in 10 CFR 50.2, SECY-94-084, occurrences. Policy and Technical Issues Associated with the Regulatory Treatment of Non-Safety Systems in Passive Plant Designs (Ref.
(2) A means which is independent and diverse from the 32), and the prior application of reactivity control requirements.
other(s), shall be capable of controlling the rate of reactivity changes resulting from planned, normal power changes to (1) Currently the second sentence of GDC 26 states, that one of the assure that the design limits for the fission product barriers are reactivity control systems shall use control rods and shall be capable not exceeded. of reliably controlling reactivity changes to ensure that, under
(3) A means of inserting negative reactivity at a sufficient rate conditions of normal operation, including AOOs, and with and amount to assure, with appropriate margin for malfunctions, appropriate margin for malfunctions such as stuck rods, specified that the capability to cool the core is maintained and a means of acceptable fuel design limits are not exceeded. The staff recognizes shutting down the reactor and maintaining, at a minimum, a safe that specifying control rods may not be suitable for advanced shutdown condition following a postulated accident. reactors. Additionally, reliably controlling reactivity, as applied to GDC 26, has been interpreted as ensuring the control rods are
(4) A means for holding the reactor shutdown under conditions capable of rapidly (i.e., within a few seconds) shutting down the which allow for interventions such as fuel loading, inspection reactor (ADAMS Accession No. ML16292A589) (Ref. 30).
and repair shall be provided.
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).
The wording of reliably controlling reactivity in GDC 26 has been replaced with inserting negative reactivity at a sufficient rate and amount to more clearly define the requirement. For a non- LWR design the rate of negative reactivity insertion may not Appendix A to RG 1.232, Rev. 0, Page A-11
APPENDIX A. ADVANCED REACTOR DESIGN CRITERIA
III. Reactivity Control Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC
necessitate rapid (within seconds) insertion but should occur in a time frame such that the fission product barrier design limits are not exceeded.
The term specified acceptable fuel design limits is replaced with design limits for fission product barriers to be consistent with the AOO acceptance criteria while also addressing liquid fueled reactors which may not have SAFDLs. ARDC 10 and ARDC 15 provide the appropriate design limits for the fuel and reactor coolant boundary, respectively.
The wording safe shutdown is achieved and maintained has been added again to more clearly define the requirements associated with reliably controlling reactivity in GDC 26. SECY-94-084, Policy and Technical Issues Associated with the Regulatory Treatment of Non-Safety Systems in Passive Plant Designs (Ref.
32), describes the characteristics of a safe shutdown condition as reactor subcriticality, decay heat removal, and radioactive materials containment. ARDC 26 (1) clearly defines that reactor shutdown at any time during the transient is the performance requirement. The distinction between during and following the transient is discussed in (2) below.
In regards to safety class, the capability to insert negative reactivity at a rate and amount to preserve the fission product barrier(s) and to shut down the reactor during an AOO is identified as a function performed by safety-related SSCs in the 10 CFR 50.2 definition of safety-related SSCs.
(2) The first sentence of GDC 26, states that two independent reactivity control systems of different design principles shall be provided. The third sentence of GDC 26, states that the second reactivity control system shall be capable of reliably controlling the Appendix A to RG 1.232, Rev. 0, Page A-12
APPENDIX A. ADVANCED REACTOR DESIGN CRITERIA
III. Reactivity Control Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC
rate of changes resulting from planned, normal power changes (including xenon burnout) to assure specified acceptable fuel design limits are not exceeded. ARDC 26 (2) is consistent with the current requirements of the second reactivity control system specified in GDC 26. The words including xenon burnout have been deleted as this may not be as important for some non-LWR reactor designs.
Also, of different design principles from the first sentence of GDC 26 has been replaced with independent and diverse to clarify the requirement. The reactivity means given by ARDC 26
(2) is a system important to safety but not necessarily safety-related as it does not mitigate an AOO or accident but is used to control planned, normal reactivity changes such that the design limits for the fission product barriers are preserved thereby minimizing challenges to the safety-related reactivity control means or protection system.
The term independent and diverse indicates no shared systems or components and a design which is different enough such that no common failure modes exist between the system or means in ARDC
26 (2) and safety-related systems in ARDC 26 (1) and (3).
(3) 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 applied to GDC 27 requires that the reactor achieve and maintain a safe, stable condition, including subcriticality, using only safety related equipment with margin for stuck rods (ADAMS Accession No. ML16116A083) (Ref. 29).
ARDC 26 (3) is written to clarify that shut down following a postulated accident using safety-related equipment or means is Appendix A to RG 1.232, Rev. 0, Page A-13
APPENDIX A. ADVANCED REACTOR DESIGN CRITERIA
III. Reactivity Control Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC
required. The term following a postulated accident refers to the time when plant parameters are relatively stable, no additional means of mitigation are needed and margins to acceptance criteria are constant or increasing. ARDC 26 allows for a return to power during a postulated accident consistent with the current licensing basis of some existing PWRs if sufficient heat removal capability exists (e.g., PWR main steam line break accident), but ARDC 26
(1) precludes a return to power during an AOO.
(4) The fourth sentence of GDC 26 regarding the capability to reach cold shutdown has been generalized in ARDC 26 (4) to refer to activities which are performed at conditions below (less limiting than) those normally associated with safe shutdown. SECY-94-084 (Ref. 32) describes staff positions on obtaining a cold shutdown and explains that the requirement to bring the plant to cold shutdown is driven by the need to inspect and repair a plant following an accident. In regards to safety class, the capability to bring the plant to a cold shutdown is not covered by the definition of safety-related SSCs in 10 CFR 50.2, and most operating pressurized-water reactors have not credited safety-related SSCs to satisfy this requirement of GDC 26. Based on the information provided above, the system credited for holding the reactor subcritical under conditions necessary for activities such as refueling, inspection and repair is identified as an important to safety system.
27 Combined reactivity control systems capability.
DELETEDInformation incorporated into ARDC 26
28 Reactivity limits. Reactor coolant pressure boundary has been relabeled as reactor The reactivity control systems shall be designed with coolant boundary to create a more broadly applicable non-LWR
appropriate limits on the potential amount and rate of reactivity term that defines the boundary without giving any implication of increase to ensure that the effects of postulated reactivity system operating pressure. As such, the term reactor coolant accidents can neither (1) result in damage to the reactor coolant boundary is applicable to non-LWRs that operate at either low or boundary greater than limited local yielding nor (2) sufficiently high pressure.
Appendix A to RG 1.232, Rev. 0, Page A-14
APPENDIX A. ADVANCED REACTOR DESIGN CRITERIA
III. Reactivity Control Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC
disturb the core, its support structures, or other reactor vessel internals to impair significantly the capability to cool the core. The list of postulated reactivity accidents has been deleted to make the ARDC technology inclusive.
29 Protection against anticipated operational occurrences.
Same as GDC
The protection and reactivity control systems shall be designed to assure an extremely high probability of accomplishing their safety functions in the event of anticipated operational occurrences.
IV. Fluid Systems Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC
30 Quality of reactor coolant boundary. Reactor coolant pressure boundary has been relabeled as reactor Components that are part of the reactor coolant boundary shall coolant boundary to create a more broadly applicable non-LWR
be designed, fabricated, erected, and tested to the highest quality term that defines the boundary without giving any implication of standards practical. Means shall be provided for detecting and, system operating pressure. As such, the term "reactor coolant to the extent practical, identifying the location of the source of boundary" is applicable to non-LWRs that operate at either low or reactor coolant leakage. high pressure.
31 Fracture prevention of reactor coolant boundary. Reactor coolant pressure boundary has been relabeled as reactor The reactor coolant boundary shall be designed with sufficient coolant boundary to create a more broadly applicable non-LWR
margin to ensure that when stressed under operating, term that defines the boundary without giving any implication of maintenance, testing, and postulated accident conditions, (1) the system operating pressure. As such, the term "reactor coolant boundary behaves in a nonbrittle manner and (2) the probability boundary" is applicable to non-LWRs that operate at either low or of rapidly propagating fracture is minimized. The design shall high pressure.
reflect consideration of service temperatures, service degradation of material properties, creep, fatigue, stress rupture, Specific examples are added to the ARDC to account for the high and other conditions of the boundary material under operating, design and operating temperatures, coolant composition, maintenance, testing, and postulated accident conditions and the contaminants, and reaction products Appendix A to RG 1.232, Rev. 0, Page A-15
APPENDIX A. ADVANCED REACTOR DESIGN CRITERIA
IV. Fluid Systems Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC
uncertainties in determining (1) material properties, (2) the effects of irradiation and coolant composition, including contaminants and reaction products, on material properties,,
(3) residual, steady-state, and transient stresses, and (4) size of flaws.
32 Inspection of reactor coolant boundary. Reactor coolant pressure boundary has been relabeled as reactor Components that are part of the reactor coolant boundary shall coolant boundary to create a more broadly applicable non-LWR
be designed to permit (1) periodic inspection and functional term that defines the boundary without giving any implication of testing of important areas and features to assess their structural system operating pressure. As such, the term "reactor coolant and leaktight integrity, and (2) an appropriate material boundary" is applicable to non-LWRs that operate at either low or surveillance program for the reactor vessel. high pressure.
The staff modified the LWR GDC by replacing the term reactor pressure vessel with reactor vessel, which the staff believes is a more generically applicable term.
Functional testing is testing that assesses component and system operational readiness such as required in the ASME OM Code as incorporated by reference in 10 CFR 50.55a and in Plant Technical Specifications.
33 Reactor coolant inventory maintenance. ARDC 33 was relabeled as inventory maintenance to provide A system to maintain reactor coolant inventory for protection more flexibility for advanced reactor designs. The first sentence is against small breaks in the reactor coolant boundary shall be modified so that it ends with ...shall be provided as necessary provided as necessary to ensure that specified acceptable fuel and is combined with the second sentence as necessary to design limits are not exceeded as a result of reactor coolant ensure (without the opening phrase The system safety function inventory loss due to leakage from the reactor coolant boundary shall be) to recognize that the inventory control system may be and rupture of small piping or other small components that are unnecessary for some designs to maintain safety functions that part of the boundary. The system shall be designed to ensure ensure fuel design limits are not exceeded.
that the system safety function can be accomplished using the piping, pumps, and valves used to maintain reactor coolant Reactor coolant pressure boundary has been relabeled as reactor inventory during normal reactor operation. coolant boundary to create a more broadly applicable non-LWR
term that defines the boundary without giving any implication of Appendix A to RG 1.232, Rev. 0, Page A-16
APPENDIX A. ADVANCED REACTOR DESIGN CRITERIA
IV. Fluid Systems Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC
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 in reactor coolant inventory levels such that the residual heat removal function would still be performed, and the fuel design limits met, no safety function would be associated with the inventory maintenance system.
The GDC reference to electric power was removed. Refer to ARDC 17 concerning those systems that require electric power.
34 Residual heat removal. In most advanced reactor designs, a single system (i.e., the residual A system to remove residual heat shall be provided. For normal heat removal system) is provided to perform both the residual heat operations and anticipated operational occurrences, the system removal and emergency core cooling functions. In this case, the safety function shall be to transfer fission product decay heat single system would be designed to meet the requirements of and other residual heat from the reactor core at a rate such that ARDC 34 and ARDC 35. (for more discussion see NUREG-0968 specified acceptable fuel design limits and the design conditions (Ref. 5) and NUREG-1368 (Ref.4)) However, the staff of the reactor coolant boundary are not exceeded. acknowledges that this may not be the case for every advanced reactor design. Therefore, to allow current and future non-LWR
Suitable redundancy in components and features and suitable designers the flexibility to provide a single system or multiple interconnections, leak detection, and isolation capabilities shall systems to perform residual heat removal and emergency core be provided to ensure that the system safety function can be cooling, the staff decided to keep the ARDC 34 and ARDC 35 accomplished, assuming a single failure. separate in lieu of combining them into a single criterion. The staffs approach to provide two separate criteria is consistent with the approach taken in the LWR GDCs.
Appendix A to RG 1.232, Rev. 0, Page A-17
APPENDIX A. ADVANCED REACTOR DESIGN CRITERIA
IV. Fluid Systems Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC
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.
The GDC reference to electric power was removed. Refer to ARDC 17 concerning those systems that require electric power.
35 Emergency core cooling system. In most advanced reactor designs, a single system (i.e, the residual A system to assure sufficient core cooling during postulated heat removal system) is provided to perform both the residual heat accidents and to remove residual heat following postulated removal and emergency core cooling functions. In this case, the accidents shall be provided. The system safety function shall be single system would be designed to meet the requirements of to transfer heat from the reactor core during and following ARDC 34 and ARDC 35. (for more discussion see NUREG-0968 postulated accidents such that fuel and clad damage that could (Ref. 5) and NUREG-1368 (Ref. 4)) However, the staff interfere with continued effective core cooling is prevented. 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 Appendix A to RG 1.232, Rev. 0, Page A-18
APPENDIX A. ADVANCED REACTOR DESIGN CRITERIA
IV. Fluid Systems Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC
coolant accidents may still require analysis in conjunction with postulated accidents if they are relevant to the design.
The GDC reference to electric power was removed. Refer to ARDC 17 concerning those systems that require electric power.
36 Inspection of emergency core cooling system. In most advanced reactor designs, a single system (i.e., the residual A system that provides emergency core cooling shall be heat removal system) is provided to perform both the residual heat designed to permit appropriate periodic inspection of important removal and emergency core cooling functions. In this case, the components to ensure the integrity and capability of the system. single system would be designed to meet the requirements of ARDC 34 and ARDC 35. (for more discussion see NUREG-0968 (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-inclusive ARDC.
Appendix A to RG 1.232, Rev. 0, Page A-19
APPENDIX A. ADVANCED REACTOR DESIGN CRITERIA
IV. Fluid Systems Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC
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 emergency core cooling system. In most advanced reactor designs, a single system (i.e., the residual A system that provides emergency core cooling shall be heat removal system) is provided to perform both the residual heat designed to permit appropriate periodic functional testing to removal and emergency core cooling functions. In this case, the ensure (1) the structural and leaktight integrity of its single system would be designed to meet the requirements of components, (2) the operability and performance of the system ARDC 34 and ARDC 35. (for more discussion see NUREG-0968 components, and (3) the operability of the system as a whole (Ref. 5) and NUREG-1368 (Ref. 4)) However, the staff and, under conditions as close to design as practical, the acknowledges that this may not be the case for every advanced performance of the full operational sequence that brings the reactor design. Therefore, to allow current and future non-LWR
system into operation, including operation of any associated designers the flexibility to provide a single system or multiple systems and interfaces necessary to transfer decay heat to the systems to perform residual heat removal and emergency core ultimate heat sink. cooling, the staff decided to keep the ARDC 34 and ARDC 35 separate in lieu of combining them into a single criterion. The staffs approach to provide two separate criteria is consistent with the approach taken in the LWR GDCs.
The ARDC has slightly different wording than the GDC to clarify the scope of the criterion. Any system, or portions of a system, credited with an emergency core cooling function during postulated accidents (for example, a system that performs both the residual heat removal function and the emergency core cooling function) would need to meet ARDC 37.
Specific mention of pressure testing has been removed yet remains a potential requirement should it be necessary as a component of appropriate periodic functional testing... of cooling systems.
Appendix A to RG 1.232, Rev. 0, Page A-20
APPENDIX A. ADVANCED REACTOR DESIGN CRITERIA
IV. Fluid Systems Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC
Functional testing is testing that assesses component and system operational readiness such as required in the ASME OM Code as incorporated by reference in 10 CFR 50.55a and in Plant Technical Specifications.
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, structural and equipment cooling 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 ARDC 17 concerning those systems that require electric power.
38 Containment heat removal. as necessary is meant to condition an ARDC 38 application A system to remove heat from the reactor containment shall be to designs requiring heat removal for conventional containments provided as necessary to maintain the containment pressure and that are found to require heat removal measures.
temperature within acceptable limits following postulated accidents. The LOCA reference has been removed to provide for any postulated accident that might affect the containment structure.
Suitable redundancy in components and features and suitable interconnections, leak detection, isolation, and containment Containment structure safety system redundancy is addressed in capabilities shall be provided to ensure that the system safety the second paragraph.
function can be accomplished, assuming a single failure.
39 Inspection of containment heat removal system. Examples were deleted to make the ARDC technology inclusive.
Appendix A to RG 1.232, Rev. 0, Page A-21
APPENDIX A. ADVANCED REACTOR DESIGN CRITERIA
IV. Fluid Systems Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC
The containment heat removal system shall be designed to permit appropriate periodic inspection of important components to ensure the integrity and capability of the system.
40 Testing of containment heat removal system. Specific mention of pressure testing has been removed yet The containment heat removal system shall be designed to remains a potential requirement should it be necessary as a permit appropriate periodic functional testing to ensure (1) the component of appropriate periodic functional testing... of structural and leaktight integrity of its components, (2) the containment heat removal.
operability and performance of the system components, and
(3) the operability of the system as a whole, and under Functional testing is testing that assesses component and system conditions as close to the design as practical, the performance of operational readiness such as required in the ASME OM Code as the full operational sequence that brings the system into incorporated by reference in 10 CFR 50.55a and in Plant Technical operation, including the operation of associated systems. Specifications.
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 system, 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 ARDC 17 concerning those systems that require electric power.
41 Containment atmosphere cleanup. Advanced reactors offer potential for reaction product generation Systems to control fission products and other substances that that is different from that associated with clad metal-water may be released into the reactor containment shall be provided interactions. Therefore, the terms hydrogen and oxygen are as necessary to reduce, consistent with the functioning of other removed while other substances is retained to allow for associated systems, the concentration and quality of fission exceptions.
products released to the environment following postulated accidents and to control the concentration of other substances in Considering that a passive containment cooling system may be the containment atmosphere following postulated accidents to used or that the containment may have an additional safety Appendix A to RG 1.232, Rev. 0, Page A-22
APPENDIX A. ADVANCED REACTOR DESIGN CRITERIA
IV. Fluid Systems Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC
ensure that containment integrity and other safety functions are function other than radionuclide retention, additional wording for maintained. maintaining safety functions is added.
Each system shall have suitable redundancy in components and The GDC reference to electric power was removed. Refer to features and suitable interconnections, leak detection, isolation, ARDC 17 concerning those systems that require electric power.
and containment capabilities to ensure that its safety function can be accomplished, assuming a single failure.
42 Inspection of containment atmosphere cleanup systems.
Same as GDC
The containment atmosphere cleanup systems shall be designed to permit appropriate periodic inspection of important components, such as filter frames, ducts, and piping to assure the integrity and capability of the systems.
43 Testing of containment atmosphere cleanup systems. Active has been deleted in item (2), as appropriate operability The containment atmosphere cleanup systems shall be designed and performance testing of system components is required to permit appropriate periodic functional testing to ensure regardless of an active or passive nature, as are cited examples of
(1) the structural and leaktight integrity of its components, active system components.
(2) the operability and performance of the system components, and (3) the operability of the systems as a whole and, under Examples of active systems under item (2) have been deleted, both conditions as close to design as practical, the performance of the to conform to similar wording in ARDC 37 and 40 and ensure that full operational sequence that brings the systems into operation, passive as well as active system components are considered.
including the operation of associated systems.
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.
Functional testing is testing that assesses component and system operational readiness such as required in the ASME OM Code as Appendix A to RG 1.232, Rev. 0, Page A-23
APPENDIX A. ADVANCED REACTOR DESIGN CRITERIA
IV. Fluid Systems Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC
incorporated by reference in 10 CFR 50.55a and in Plant Technical Specifications.
The GDC reference to electric power was removed. Refer to ARDC 17 concerning those systems that require electric power.
44 Structural and equipment cooling. This renamed ARDC accounts for advanced reactor design system A system to transfer heat from structures, systems, and differences to include cooling requirements for SSCs, if applicable;
components important to safety to an ultimate heat sink shall be this ARDC does not address the residual heat removal system provided, as necessary, to transfer the combined heat load of required under ARDC 34, and emergency core cooling system these structures, systems, and components under normal (ECCS) system under ARDC 35 operating and accident conditions.
The GDC reference to electric power was removed. Refer to Suitable redundancy in components and features and suitable ARDC 17 concerning those systems that require electric power.
interconnections, leak detection, and isolation capabilities shall be provided to ensure that the system safety function can be accomplished, assuming a single failure.
45 Inspection of structural and equipment cooling systems. This renamed ARDC accounts for advanced reactor system design The structural and equipment cooling systems shall be designed differences to include possible cooling requirements for SSCs to permit appropriate periodic inspection of important important to safety.
components, such as heat exchangers and piping, to ensure the integrity and capability of the systems.
46 Testing of structural and equipment cooling systems. This renamed ARDC accounts for advanced reactor system design The structural and equipment cooling systems shall be designed differences to include possible cooling requirements for SSCs to permit appropriate periodic functional testing to ensure important to safety. Specific mention of pressure testing has
(1) the structural and leaktight integrity of their components, been removed yet remains a potential requirement should it be
(2) the operability and performance of the system components, necessary as a component of appropriate periodic functional and (3) the operability of the systems as a whole and, under testing... of cooling systems. A non-leaktight system may be conditions as close to design as practical, the performance of the acceptable for some designs provided that (1) the system leakage full operational sequences that bring the systems into operation does not impact safety functions under all conditions, and (2)
for reactor shutdown and postulated accidents, including the defense in depth is not impacted by system leakage.
operation of associated systems.
Appendix A to RG 1.232, Rev. 0, Page A-24
APPENDIX A. ADVANCED REACTOR DESIGN CRITERIA
IV. Fluid Systems Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC
Functional testing is testing that assesses component and system operational readiness such as required in the ASME OM Code as incorporated by reference in 10 CFR 50.55a and in Plant Technical Specifications.
Active has been deleted in item (2) because appropriate operability and performance tests of system components are required regardless of their active or passive natur
e. 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.
Appendix A to RG 1.232, Rev. 0, Page A-25
APPENDIX A. ADVANCED REACTOR DESIGN CRITERIA
V. Reactor Containment Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC
50 Containment design basis. ARDC 50 specifically addresses a containment structure in the The containment structure, including access openings, opening sentence and ARDC 51-57 support the containment penetrations, and the containment heat removal system shall be structures design basis. Therefore, ARDC 51-57 are modified by designed so that the containment structure and its internal adding the word structure to highlight the containment structure- compartments can accommodate, without exceeding the design specific criteria.
leakage rate and with sufficient margin, the calculated pressure and temperature conditions resulting from postulated accidents. The word reactor was removed because the containment is a This margin shall reflect consideration of (1) the effects of barrier between the fission products and the environment. There are potential energy sources that have not been included in the diverse advanced reactor designs and, hence, there is no single determination of the peak conditions, (2) the limited experience containment concept.
and experimental data available for defining accident phenomena and containment responses, and (3) the The phrase loss-of-coolant accident is LWR specific because this conservatism of the calculational model and input parameters. 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 GDC is LWR specific and therefore deleted.
51 Fracture prevention of containment pressure boundary. ARDC 51-57 support ARDC 50, which specifically applies to non- The boundary of the containment structure shall be designed LWR designs that use a fixed containment structure. Therefore, the with sufficient margin to ensure that, under operating, word structure is added to each of these ARDC to clearly convey maintenance, testing, and postulated accident conditions, (1) its the understanding that this criterion applies to designs employing materials behave in a nonbrittle manner and (2) the probability containment structures.
of rapidly propagating fracture is minimized. The design shall reflect consideration of service temperatures and other The word reactor was removed because the containment is a conditions of the containment boundary materials during barrier between the fission products and the environment. There are operation, maintenance, testing, and postulated accident diverse advanced reactor designs and, hence, there is no single conditions, and the uncertainties in determining (1) material containment concept.
properties, (2) residual, steady-state, and transient stresses, and
(3) size of flaws. The term ferritic was removed to avoid limiting the scope of the criterion to ferritic materials. With this revision, the staff believes Appendix A to RG 1.232, Rev. 0, Page A-26
APPENDIX A. ADVANCED REACTOR DESIGN CRITERIA
V. Reactor Containment Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC
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. ARDC 51-57 support ARDC 50, which specifically applies to non- The containment structure and other equipment that may be LWR designs that use a fixed containment structure. Therefore, the subjected to containment test conditions shall be designed so word structure is added to each of these ARDC to clearly convey that periodic integrated leakage rate testing can be conducted at the understanding that this criterion applies to designs employing containment design pressure. containment structures.
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.
53 Provisions for containment testing and inspection. ARDC 51-57 support ARDC 50, which specifically applies to non- The containment structure shall be designed to permit LWR designs that use a fixed containment structure. Therefore, the
(1) appropriate periodic inspection of all important areas, such word structure is added to each of these ARDC to clearly convey as penetrations, (2) an appropriate surveillance program, and the understanding that this criterion only applies to designs
(3) periodic testing at containment design pressure of the leak- employing containment structures.
tightness of penetrations that have resilient seals and expansion bellows. 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.
54 Piping systems penetrating containment. ARDC 51-57 support ARDC 50, which specifically applies to non- Piping systems penetrating the containment structure shall be LWR designs that use a fixed containment structure. Therefore, the provided with leak detection, isolation, and containment word structure is added to each of these ARDC to clearly convey capabilities having redundancy, reliability, and performance the understanding that this ARDC only applies to designs Appendix A to RG 1.232, Rev. 0, Page A-27
APPENDIX A. ADVANCED REACTOR DESIGN CRITERIA
V. Reactor Containment Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC
capabilities that reflect the importance to safety of isolating employing containment structures. The word reactor was these piping systems. Such piping systems shall be designed removed because the containment is a barrier between the fission with the capability to verify, by testing, the operational products and the environment. There are diverse advanced reactor readiness of any isolation valves and associated apparatus designs and, hence, there is no single containment concept. In all periodically and to confirm that valve leakage is within cases, the rules for containment penetrations to fulfill containment acceptable limits. isolation would apply. How this is accomplished should be left to the designer of the particular advanced reactor design, without being too prescriptive as to whether it is a primary or secondary or reactor containment. There may be a need for a containment structure outside the reactor region. For example, in the MSR
design, some of the liquid fuel salt is drawn off to a processing system to clean it up and remove fission products before returning it to the reactor. The liquid 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 must also meet the containment isolation requirements.
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.
ASME, International Operation and Maintenance of Nuclear Power Plants, Division 1: OM Code: Section IST (ASME OM Code)
defines operational readiness as the ability of a component to perform its specified functions. The ASME OM Code is incorporated by reference in the NRC regulations in 10 CFR 50.55a, including the definition of operational readiness for pumps, valves, and dynamic restraints.
55 Reactor coolant boundary penetrating containment. ARDC 51-57 support ARDC 50, which specifically applies to non- Each line that is part of the reactor coolant boundary and that LWR designs that use a fixed containment structure. Therefore, the penetrates the containment structure shall be provided with word structure is added to each of these ARDC to clearly convey Appendix A to RG 1.232, Rev. 0, Page A-28
APPENDIX A. ADVANCED REACTOR DESIGN CRITERIA
V. Reactor Containment Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC
containment isolation valves, as follows, unless it can be the understanding that this ARDC only applies to designs demonstrated that the containment isolation provisions for a employing containment structures. The word reactor was specific class of lines, such as instrument lines, are acceptable removed because the containment is a barrier between the fission on some other defined basis: products and the environment. There are diverse advanced reactor designs and, hence, there is no single containment concep
t. In all
(1) One locked closed isolation valve inside and one locked cases, the rules for containment penetrations to fulfill containment closed isolation valve outside containment; or isolation would apply. How this is accomplished should be left to
(2) One automatic isolation valve inside and one locked closed the designer of the particular advanced reactor design, without isolation valve outside containment; or being too prescriptive as to whether it is a primary or secondary or
(3) One locked closed isolation valve inside and one automatic reactor containment. There may be a need for a containment isolation valve outside containment. A simple check valve may structure outside the reactor region. For example, in the MSR
not be used as the automatic isolation valve outside design, some of the liquid fuel salt is drawn off to a processing containment; or system to clean it up and remove fission products before returning it
(4) One automatic isolation valve inside and one automatic to the reactor. The liquid fuel salt is highly radioactive and would isolation valve outside containment. A simple check valve may need a containment around the entire system. Alternatively, in an not be used as the automatic isolation valve outside SFR, the guard vessel would be the primary containment and, in the containment. case of the PRISM design, a dome-shaped structure above it that would be the secondary containment. The secondary containment Isolation valves outside containment shall be located as close to must also meet the containment isolation requirements.
containment as practical and upon loss of actuating power, automatic isolation valves shall be designed to take the position Reactor coolant pressure boundary has been relabeled as reactor that provides greater safety. coolant boundary to create a more broadly applicable non-LWR
term that defines the boundary without giving any implication of Other appropriate requirements to minimize the probability or system operating pressure. As such, the term reactor coolant consequences of an accidental rupture of these lines or of lines boundary is applicable to non-LWRs that operate at either low or connected to them shall be provided as necessary to ensure high pressure.
adequate safety. Determination of the appropriateness of these requirements, such as higher quality in design, fabrication, and testing; additional provisions for inservice inspection; protection against more severe natural phenomena; and additional isolation valves and containment, shall include consideration of the population density, use characteristics, and physical characteristics of the site environs.
Appendix A to RG 1.232, Rev. 0, Page A-29
APPENDIX A. ADVANCED REACTOR DESIGN CRITERIA
V. Reactor Containment Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC
56 Containment isolation. ARDC 51-57 support ARDC 50, which specifically applies to non- Each line that connects directly to the containment atmosphere LWR designs that use a fixed containment structure. Therefore, the and penetrates the containment structure shall be provided with word structure is added to each of these ARDC to clearly convey containment isolation valves as follows, unless it can be the understanding that this criterion only applies to designs demonstrated that the containment isolation provisions for a employing containment structures.
specific class of lines, such as instrument lines, are acceptable on some other defined basis: The word primary in the title and the text was removed, and the
(1) One locked closed isolation valve inside and one locked word reactor was also removed because the containment is a closed isolation valve outside containment; or barrier between the fission products and the environmen
t. There are
(2) One automatic isolation valve inside and one locked closed diverse advanced reactor designs and, hence, there is no single isolation valve outside containment; or containment concept. In all cases, the rules for containment
(3) One locked closed isolation valve inside and one automatic penetrations to fulfill containment isolation would apply. How this isolation valve outside containment. A simple check valve may is accomplished should be left to the designer of the particular not be used as the automatic isolation valve outside advanced reactor design, without being too prescriptive as to containment; or whether it is a primary or secondary or reactor containmen
t. There
(4) One automatic isolation valve inside and one automatic may be a need for a containment structure outside the reactor isolation valve outside containment. A simple check valve may region. For example, in the MSR design, some of the liquid fuel salt not be used as the automatic isolation valve outside is drawn off to a processing system to clean it up and remove containment. fission products before returning it to the reactor. The liquid fuel salt is highly radioactive and would need a containment around the Isolation valves outside containment shall be located as close to entire system. Alternatively, in an SFR, the guard vessel would be the containment as practical and upon loss of actuating power, the primary containment and, in the case of the PRISM design, a automatic isolation valves shall be designed to take the position dome-shaped structure above it that would be the secondary that provides greater safety. containment. The secondary containment must also meet the containment isolation requirements.
57 Closed system isolation valves. ARDC 51-57 support ARDC 50, which specifically applies to non- Each line that penetrates the containment structure and is neither LWR designs that use a fixed containment structure. Therefore, the part of the reactor coolant boundary nor connected directly to word structure is added to each of these ARDC to clearly convey the containment atmosphere shall have at least one containment the understanding that this criterion only applies to designs isolation valve, unless it can be demonstrated that the employing containment structures. The word reactor was containment safety function can be met without an isolation removed because the containment is a barrier between the fission valve and assuming failure of a single active component. The products and the environment. There are diverse advanced reactor Appendix A to RG 1.232, Rev. 0, Page A-30
APPENDIX A. ADVANCED REACTOR DESIGN CRITERIA
V. Reactor Containment Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC
isolation valve, if required, shall be either automatic, or locked designs and, hence, there is no single containment concept. In all closed, or capable of remote manual operation. This valve shall cases, the rules for containment penetrations to fulfill containment be outside containment and located as close to the containment isolation would apply. How this is accomplished should be left to as practical. A simple check valve may not be used as the the designer of the particular advanced reactor design, without automatic isolation valve. being too prescriptive as to whether it is a primary or secondary or reactor containment. There may be a need for a containment structure outside the reactor region. For example, in the MSR
design, some of the liquid fuel salt is drawn off to a processing system to clean it up and remove fission products before returning it to the reactor. The liquid fuel salt is highly radioactive and would need a containment around the entire system. Alternatively, in an SFR, the guard vessel would be the primary containment and, in the case of the PRISM design, a dome-shaped structure above it that would be the secondary containment. The secondary containment also has penetrations and needs containment isolation requirements to be fulfilled.
Reactor coolant pressure boundary is relabeled as reactor coolant boundary to create a more broadly applicable non-LWR term that defines the boundary without giving any implication of system operating pressure. As such, the term reactor coolant boundary is applicable to non-LWRs that operate at either low or high pressure.
Appendix A to RG 1.232, Rev. 0, Page A-31
APPENDIX A. ADVANCED REACTOR DESIGN CRITERIA
VI. Fuel and Radioactivity Control Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC
60 Control of releases of radioactive materials to the environment.
Same as GDC
The nuclear power unit design shall include means to control suitably the release of radioactive materials in gaseous and liquid effluents and to handle radioactive solid wastes produced during normal reactor operation, including anticipated operational occurrences. Sufficient holdup capacity shall be provided for retention of gaseous and liquid effluents containing radioactive materials, particularly where unfavorable site environmental conditions can be expected to impose unusual operational limitations upon the release of such effluents to the environment.
61 Fuel storage and handling and radioactivity control. The underlying concept of establishing functional requirements for The fuel storage and handling, radioactive waste, and other radioactivity control in fuel storage and fuel handling systems is systems that may contain radioactivity shall be designed to independent of the design of non-LWR advanced reactors.
ensure adequate safety under normal and postulated accident However, some advanced designs may use dry fuel storage that conditions. These systems shall be designed (1) with a incorporates cooling jackets that can be liquid cooled or air cooled capability to permit appropriate periodic inspection and testing to remove heat. This modification to this GDC allows for both of components important to safety, (2) with suitable shielding liquid and air cooling of the dry fuel storage containers.
for radiation protection, (3) with appropriate containment, confinement, and filtering systems, (4) with a residual heat removal capability having reliability and testability that reflects the importance to safety of decay heat and other residual heat removal, and (5) to prevent significant reduction in fuel storage cooling under accident conditions.
62 Prevention of criticality in fuel storage and handling.
Same as GDC
Criticality in the fuel storage and handling system shall be prevented by physical systems or processes, preferably by use of geometrically safe configurations.
Appendix A to RG 1.232, Rev. 0, Page A-32
APPENDIX A. ADVANCED REACTOR DESIGN CRITERIA
VI. Fuel and Radioactivity Control Criterion ARDC Title and Content NRC Rationale for Adaptions to GDC
63 Monitoring fuel and waste storage.
Same as GDC
Appropriate systems shall be provided in fuel storage and radioactive waste systems and associated handling areas (1) to detect conditions that may result in loss of residual heat removal capability and excessive radiation levels and (2) to initiate appropriate safety actions.
64 Monitoring radioactivity releases. The phrase spaces containing components for recirculation of loss- Means shall be provided for monitoring the reactor containment of-coolant accident fluids was removed to allow for plant designs atmosphere, effluent discharge paths, and plant environs for that do not have LOCA fluids but may have other similar equipment radioactivity that may be released from normal operations, in spaces where radioactivity should be monitored.
including anticipated operational occurrences, and from postulated accidents.
Appendix A to RG 1.232, Rev. 0, Page A-33
APPENDIX B
SODIUM-COOLED FAST REACTOR DESIGN CRITERIA
The table below contains the sodium-cooled fast reactor design criteria (SFR-DC). These criteria are applicable to SFRs of both pool- and loop-type designs.13 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)
13 The technology-specific design criteria were developed using available design information, previous NRC pre- application reviews of the design types, and more recent industry and DOE national laboratory initiatives in these technology areas (see Reference 17). It is the responsibility of the designer or applicant to provide and justify the PDC
for a specific design.
Appendix B to RG 1.232, Rev. 0, Page B-1
APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA
I. Overall Requirements Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC
1 Quality standards and records.
Same as GDC
Structures, systems, and components important to safety shall be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed. Where generally recognized codes and standards are used, they shall be identified and evaluated to determine their applicability, adequacy, and sufficiency and shall be supplemented or modified as necessary to assure a quality product in keeping with the required safety function. A quality assurance program shall be established and implemented in order to provide adequate assurance that these structures, systems, and components will satisfactorily perform their safety functions. Appropriate records of the design, fabrication, erection, and testing of structures, systems, and components important to safety shall be maintained by or under the control of the nuclear power unit licensee throughout the life of the unit.
2 Design bases for protection against natural phenomena.
Same as GDC
Structures, systems, and components important to safety shall be designed to withstand the effects of natural phenomena such as earthquakes, tornadoes, hurricanes, floods, tsunami, and seiches without loss of capability to perform their safety functions. The design bases for these structures, systems, and components shall reflect: (1) Appropriate consideration of the most severe of the natural phenomena that have been historically reported for the site and surrounding area, with sufficient margin for the limited accuracy, quantity, and period of time in which the historical data have been accumulated, (2)
appropriate combinations of the effects of normal and accident conditions with the effects of the natural phenomena and (3) the importance of the safety functions to be performed.
Appendix B to RG 1.232, Rev. 0, Page B-2
APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA
I. Overall Requirements Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC
3 Fire protection. The phrase containing examples where noncombustible and fire- Same as ARDC resistant materials must be used has been broadened to apply to all Structures, systems, and components important to safety shall advanced reactor designs.
be designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and Instead of and, the phrase locations with structures, systems, and explosions. Noncombustible and fire-resistant materials shall be components (SSCs) important to safety uses or, which is used wherever practical throughout the unit, particularly in logically correct in this case.
locations with structures, systems, or components important to safety. Fire detection and fighting systems of appropriate capacity and capability shall be provided and designed to minimize the adverse effects of fires on structures, systems, and components important to safety. Firefighting systems shall be designed to ensure that their rupture or inadvertent operation does not significantly impair the safety capability of these structures, systems, and components.
4 Environmental and dynamic effects design bases. This change removes the light-water reactor (LWR) emphasis on Structures, systems, and components important to safety shall loss-of-coolant accidents (LOCAs) that may not apply to every be designed to accommodate the effects of, and to be design. For example, helium is not needed in a MHTGR to remove compatible with, the environmental conditions associated with heat from the core during postulated accidents and does not have the normal operation, maintenance, testing, anticipated operational same importance as water does for LWR designs to ensure that fuel occurrences, and postulated accidents, including the effects of integrity is maintained. Therefore, a specific reference to LOCAs is liquid sodium and its aerosols and oxidation products. These not applicable to all designs. LOCAs may still require analysis in structures, systems, and components shall be appropriately conjunction with postulated accidents if relevant to the design.
protected against dynamic effects, including the effects of missiles, pipe whipping, and discharging fluids that may result The phrase the environmental conditions associated with from equipment failures and from events and conditions outside anticipated operational occurrences has been added to ensure that the nuclear power unit. However, dynamic effects associated the criterion would apply to all SFR design-basis events, as with postulated pipe ruptures in nuclear power units may be suggested in NUREG-1368.
excluded from the design basis when analyses reviewed and approved by the Commission demonstrate that the probability of A new sentence is added to ensure the designer considers the effects fluid system piping rupture is extremely low under conditions of sodium leakage and associated chemical reactions with SSCs consistent with the design basis for the piping. important to safety, which must be protected.
Appendix B to RG 1.232, Rev. 0, Page B-3
APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA
I. Overall Requirements Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC
Chemical consequences of accidents, such as sodium leakage, shall be appropriately considered for the design of structures, systems, and components important to safety, which must be protected.
5 Sharing of structures, systems, and components.
Same as GDC
Structures, systems, and components important to safety shall not be shared among nuclear power units unless it can be shown that such sharing will not significantly impair their ability to perform their safety functions, including, in the event of an accident in one unit, an orderly shutdown and cooldown of the remaining units.
II. Multiple Barriers Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC
10 Reactor design.
Same as GDC
The reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.
11 Reactor inherent protection. The wording has been changed to broaden the applicability from Same as ARDC coolant systems to additional factors (including structures or other The reactor core and associated systems that contribute to fluids) that may contribute to reactivity feedback. These systems are reactivity feedback shall be designed so that, in the power to be designed to compensate for rapid reactivity increase.
operating range, the net effect of the prompt inherent nuclear Appendix B to RG 1.232, Rev. 0, Page B-4
APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA
II. Multiple Barriers Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC
feedback characteristics tends to compensate for a rapid increase in reactivity.
12 Suppression of reactor power oscillations. The word structures was added because items such as reflectors, Same as ARDC which could be considered either outside or not part of the reactor The reactor core; associated structures; and associated coolant, core, may affect susceptibility of the core to power oscillations.
control, and protection systems shall be designed to ensure that power oscillations that can result in conditions exceeding specified acceptable fuel design limits are not possible or can be reliably and readily detected and suppressed.
13 Instrumentation and control. Reactor coolant pressure boundary has been relabeled as primary Instrumentation shall be provided to monitor variables and coolant boundary to conform to standard terms used in the liquid- systems over their anticipated ranges for normal operation, for metal reactor (LMR) industry.
anticipated operational occurrences, and for accident conditions, as appropriate to ensure adequate safety, including those The use of the term primary indicates that the SFR-DC are variables and systems that can affect the fission process, the applicable to the primary cooling system, not the intermediate integrity of the reactor core, the primary coolant boundary, and cooling system.
the containment and its associated systems. Appropriate controls shall be provided to maintain these variables and systems within prescribed operating ranges.
14 Primary coolant boundary. Reactor coolant pressure boundary (RCPB) has been relabeled as The primary coolant boundary shall be designed, fabricated, primary coolant boundary to conform to standard terms used in erected, and tested so as to have an extremely low probability of the LMR industry.
abnormal leakage, of rapidly propagating failure, and of gross rupture. The use of the term primary indicates that the SFR-DC are applicable only to the primary cooling system, not the intermediate cooling system.
The cover gas boundary is included as part of the primary coolant boundary (referred to as RCPB by PRISM) per NUREG-1368 (page
3-38) (Ref. 4).
Appendix B to RG 1.232, Rev. 0, Page B-5
APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA
II. Multiple Barriers Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC
15 Primary coolant system design. Reactor coolant pressure boundary has been relabeled as primary The primary coolant system and associated auxiliary, control, coolant boundary to conform to standard terms used in the LMR
and protection systems shall be designed with sufficient margin industry.
to ensure that the design conditions of the primary coolant boundary are not exceeded during any condition of normal The use of the term primary indicates that the SFR-DC are operation, including anticipated operational occurrences. applicable only to the primary cooling system, not the intermediate cooling system.
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) (Ref. 4).
16 Containment design. The Commission approved the staffs recommendation to restrict A reactor containment consisting of a low-leakage, pressure- the leakage of the containment to be less than that needed to meet retaining structure surrounding the reactor and its primary the acceptable onsite and offsite dose consequence limits in cooling system shall be provided to control the release of SECY-93-092 (Ref. 7). Therefore, the Commission agreed that the radioactivity to the environment and to ensure that the reactor containment leakage for advanced reactors, similar to and including containment design conditions important to safety are not PRISM, NUREG-1368 (Ref. 4) should not be required to meet the exceeded for as long as postulated accident conditions require. essentially leaktight statement in GDC 16.
The containment leakage shall be restricted to be less than that Furthermore, all past, and current, SFR designs use a low-leakage, needed to meet the acceptable onsite and offsite dose pressure-retaining containment concept, which aims to provide a consequence limits, as specified in 10 CFR 50.34 for postulated barrier to contain the fission products and other substances and to accidents. 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 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.
Appendix B to RG 1.232, Rev. 0, Page B-6
APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA
II. Multiple Barriers Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC
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. 33), indicates that it is necessary for structures to 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. 34) 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. 35), 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. 36) notes that Beschreibung der Forschungsanlage zur Untersuchung nuklearer Aerosole (FAUNA)
spray fire experiments show peak pressures in containment over
3.5 bar within the first 5 seconds, gradually tapering downwards to less than 3.5 bar at 25 seconds.
17 Electric power systems. The electric power systems are required to provide reliable power Same as ARDC for SSCs during anticipated operational occurrences or postulated Electric power systems shall be provided when required to accident conditions when those SSCs safety functions require permit functioning of structures, systems, and components. The electric power. The safety functions are established by the safety safety function for each power system shall be to provide analyses (i.e. design basis accidents). Where electric power is Appendix B to RG 1.232, Rev. 0, Page B-7
APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA
II. Multiple Barriers Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC
sufficient capacity and capability to ensure that (1) that the needed for anticipated operational occurrences or postulated design limits for the fission product barriers are not exceeded as accidents, the electric power systems shall be sufficient in capacity a result of anticipated operational occurrences and (2) safety and capability to ensure that safety functions as well as important to functions that rely on electric power are maintained in the event safety functions are maintained. The electric power systems provide of postulated accidents. redundancy and defense-in-depth since there would be a minimum of two power systems.
The electric power systems shall include an onsite power system and an additional power system. The onsite electric power Compared to GDC 17, more emphasis is placed herein on requiring system shall have sufficient independence, redundancy, and reliability of the overall power supply scheme rather than fully testability to perform its safety functions, assuming a single prescribing how such reliability can be attained. For example, failure. An additional power system shall have sufficient reference to offsite electric power systems was deleted to provide independence and testability to perform its safety function. for those reactor designs that do not depend on offsite power for the functioning of SSCs important to safety or do not connect to a If electric power is not needed for anticipated operational power grid.
occurrences or postulated accidents, the design shall demonstrate that power for important to safety functions is The onsite power system is envisioned as a fully Class 1E power provided. system and the additional power system is left to the discretion of the designer as long as it meets the performance criteria in paragraph one and the design criteria of paragraph two. For example, the additional independent power source could be from the electrical grid, a diesel generator, a combustion gas turbine or some other alternative, again, at the discretion of the designer.
In this context, important to safety functions refer to the broader, potentially non-safety related functions such as post-accident monitoring, control room habitability, emergency lighting, radiation monitoring, communications and/or any others that may be deemed appropriate for the given design. The electric power system for important to safety functions could be non-Class 1E and would not be required to have redundant power sources.
18 Inspection and testing of electric power systems. ARDC 18 is a design-independent companion criterion to Same as ARDC. ARDC 17.
Appendix B to RG 1.232, Rev. 0, Page B-8
APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA
II. Multiple Barriers Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC
Electric power systems important to safety shall be designed to permit appropriate periodic inspection and testing of important Wording pertaining to additional system examples has been deleted areas and features, such as wiring, insulation, connections, and to allow increased flexibility associated with various designs.
switchboards, to assess the continuity of the systems and the Specifically, the text related to the nuclear power unit, offsite power condition of their components. The systems shall be designed system, and onsite power system was deleted to be consistent with with a capability to test periodically (1) the operability and ARDC 17.
functional performance of the components of the systems, such as onsite power sources, relays, switches, and buses, and (2) the operability of the systems as a whole and, under conditions as close to design as practical, the full operation sequence that brings the systems into operation, including operation of applicable portions of the protection system, and the transfer of power among systems.
19 Control room. ARDC 19 preserves the language of GDC 19 which states (with A control room shall be provided from which actions can be emphasis added) A control room shall be provided from which taken to operate the nuclear power unit safely under normal actions can be taken to operate the nuclear power unit safely conditions and to maintain it in a safe condition under accident However some clarification of this language is warranted.
conditions. Adequate radiation protection shall be provided to permit access and occupancy of the control room under accident It is clear from this language that there is a need to for operators to conditions without personnel receiving radiation exposures in be able to take actions to control the plant. Therefore, designers excess of 5 rem total effective dose equivalent, as defined in must consider how the design of controls support safe operator
§ 50.2 for the duration of the accident. actions. In addition, NRC staff recognize that in order for operators to act safely as stated in ARDC, that operators must have certain Adequate habitability measures shall be provided to permit knowledge about the status of the plant and be able to make access and occupancy of the control room during normal decisions about the appropriate course of action given a particular operations and under accident conditions. operating circumstance. Therefore, these cognitive needs of operators should also be considered by designers when interpreting Adequate protection against sodium aerosols shall be provided ARDC 19.
to permit access and occupancy of the control room under accident conditions. This consideration should be reflected in the design of indications, displays, alarms, controls or other future technologies which are Appendix B to RG 1.232, Rev. 0, Page B-9
APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA
II. Multiple Barriers Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC
Equipment at appropriate locations outside the control room used to inform operators of plant status and may be used to support shall be provided (1) with a design capability for prompt hot the decision making process (such as computer based procedures).
shutdown of the reactor, including necessary instrumentation and controls to maintain the unit in a safe condition during hot This position is consistent with 10 CFR 50.34(f)(2)(iii) which shutdown, and (2) with a potential capability for subsequent describes the contents required in applications for construction cold shutdown of the reactor through the use of suitable permits. Amongst many other requirements, this rule indicates that procedures. the control room design must reflect state-of-the-art human factors principles. These state-of-the-art principles inherently consider both the cognitive and physical aspects of operator action as described above.
The criterion was updated to remove specific emphasis on LOCAs, which may be not appropriate for advanced designs.
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.
An additional control room habitability requirement has been proposed. It addresses a new concern: accidents that are not radiological in nature may also affect control room access and occupancy. This may include accidental sodium leakage and sodium fire, which could release sodium aerosols.
The last paragraph of the GDC has been eliminated for the SFR-DC
because it is not applicable to future applicants.
Appendix B to RG 1.232, Rev. 0, Page B-10
APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA
III. Reactivity Control Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC
20 Protection system functions.
Same as GDC
The protection system shall be designed (1) to initiate automatically the operation of appropriate systems including the reactivity control systems, to assure that specified acceptable fuel design limits are 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.
21 Protection system reliability and testability.
Same as GDC
The protection system shall be designed for high functional reliability and inservice testability commensurate with the safety functions to be performed. Redundancy and independence designed into the protection system shall be sufficient to assure that (1) no single failure results in loss of the protection function and (2) removal from service of any component or channel does not result in loss of the required minimum redundancy unless the acceptable reliability of operation of the protection system can be otherwise demonstrated. The protection system shall be designed to permit periodic testing of its functioning when the reactor is in operation, including a capability to test channels independently to determine failures and losses of redundancy that may have occurred.
Appendix B to RG 1.232, Rev. 0, Page B-11
APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA
III. Reactivity Control Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC
22 Protection system independence.
Same as GDC
The protection system shall be designed to assure that the effects of natural phenomena, and of normal operating, maintenance, testing, and postulated accident conditions on redundant channels do not result in loss of the protection function, or shall be demonstrated to be acceptable on some other defined basis. Design techniques, such as functional diversity or diversity in component design and principles of operation, shall be used to the extent practical to prevent loss of the protection function.
23 Protection system failure modes. In NUREG-1368, Table 3.3 (page 3-21) (Ref. 4), the NRC staff The protection system shall be designed to fail into a safe state recommended adding the phrase sodium and sodium reaction or into a state demonstrated to be acceptable on some other products to the list of postulated adverse environments in the GDC.
defined basis, if conditions such as disconnection of the system, Therefore, sodium and sodium reaction products are added to the loss of energy (e.g., electric power, instrument air), or second list of examples in parentheses in SFR-DC 23.
postulated adverse environments (e.g., extreme heat or cold, fire, sodium and sodium reaction products, pressure, steam, water, and radiation) are experienced.
24 Separation of protection and control systems.
Same as GDC
The protection system shall be separated from control systems to the extent that failure of any single control system component or channel, or failure or removal from service of any single protection system component or channel which is common to the control and protection systems leaves intact a system satisfying all reliability, redundancy, and independence requirements of the protection system. Interconnection of the protection and control systems shall be limited so as to assure that safety is not significantly impaired.
Appendix B to RG 1.232, Rev. 0, Page B-12
APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA
III. Reactivity Control Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC
25 Protection system requirements for reactivity control Text has been added to clarify that the protection system is designed malfunctions. to protect the specified acceptable fuel design limits for AOOs in Same as ARDC combination with a single failure; the protection system does not The protection system shall be designed to ensure that specified have to protect the specified acceptable fuel design limits during a acceptable fuel design limits are not exceeded during any postulated accident in combination with a single failure. The anticipated operational occurrence accounting for a single example was deleted to make the ARDC technology inclusive.
malfunction of the reactivity control systems.
26 Reactivity control systems. Recent licensing activity, associated with the application of GDC
Same as ARDC 26 and GDC 27 to new reactor designs (ADAMS Accession Nos.
A minimum of two reactivity control systems or means shall ML16116A083 (Ref. 29) and ML16292A589) (Ref. 30), revealed provide: that additional clarity could be provided in the area of reactivity control requirements. ARDC 26 combines the scope of GDC 26
(1) A means of inserting negative reactivity at a sufficient rate and GDC 27. The development of ARDC 26 is informed by the and amount to assure, with appropriate margin for malfunctions, proposed general design criteria of 1965 (AEC-R 2/49, November that the design limits for the fission product barriers are not 5), 1967 (32 FR 10216) (Ref. 31), current GDC 26 and 27, the exceeded and safe shutdown is achieved and maintained definition of safety-related SSC in 10 CFR 50.2, SECY-94-084, during normal operation, including anticipated operational Policy and Technical Issues Associated with the Regulatory occurrences. Treatment of Non-Safety Systems in Passive Plant Designs (Ref.
32), and the prior application of reactivity control requirements.
(2) A means which is independent and diverse from the other(s), shall be capable of controlling the rate of reactivity (1) Currently the second sentence of GDC 26 states, that one of the changes resulting from planned, normal power changes to reactivity control systems shall use control rods and shall be capable assure that the design limits for the fission product barriers are of reliably controlling reactivity changes to ensure that, under not exceeded. conditions of normal operation, including AOOs, and with
(3) A means of inserting negative reactivity at a sufficient rate appropriate margin for malfunctions such as stuck rods, specified and amount to assure, with appropriate margin for malfunctions, acceptable fuel design limits are not exceeded. The staff recognizes that the capability to cool the core is maintained and a means of that specifying control rods may not be suitable for advanced shutting down the reactor and maintaining, at a minimum, a safe reactors. Additionally, reliably controlling reactivity, as applied to shutdown condition following a postulated accident. GDC 26, has been interpreted as ensuring the control rods are capable of rapidly (i.e., within a few seconds) shutting down the reactor (ADAMS Accession No. ML16292A589) (Ref. 30).
Appendix B to RG 1.232, Rev. 0, Page B-13
APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA
III. Reactivity Control Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC
(4) A means for holding the reactor shutdown under conditions The staff changed control rods to means in recognition that which allow for interventions such as fuel loading, inspection advanced reactor designs may not rely on control rods to rapidly and repair shall be provided. shut down the reactor (e.g., alternative system designs or inherent feedback mechanisms may be relied upon to perform this function).
The wording of reliably controlling reactivity in GDC 26 has been replaced with inserting negative reactivity at a sufficient rate and amount to more clearly define the requirement. For a non- LWR design the rate of negative reactivity insertion may not necessitate rapid (within seconds) insertion but should occur in a time frame such that the fission product barrier design limits are not exceeded.
The term specified acceptable fuel design limits is replaced with design limits for fission product barriers to be consistent with the AOO acceptance criteria while also addressing liquid fueled reactors which may not have SAFDLs. ARDC 10 and ARDC 15 provide the appropriate design limits for the fuel and reactor coolant boundary, respectively.
The wording safe shutdown is achieved and maintained has been added again to more clearly define the requirements associated with reliably controlling reactivity in GDC 26. SECY-94-084, Policy and Technical Issues Associated with the Regulatory Treatment of Non-Safety Systems in Passive Plant Designs (Ref.
32), describes the characteristics of a safe shutdown condition as reactor subcriticality, decay heat removal, and radioactive materials containment. ARDC 26 (1) clearly defines that reactor shutdown at any time during the transient is the performance requirement. The distinction between during and following the transient is discussed in (2) below.
In regards to safety class, the capability to insert negative reactivity at a rate and amount to preserve the fission product barrier(s) and to Appendix B to RG 1.232, Rev. 0, Page B-14
APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA
III. Reactivity Control Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC
shut down the reactor during an AOO is identified as a function performed by safety-related SSCs in the 10 CFR 50.2 definition of safety-related SSCs.
(2) The first sentence of GDC 26, states that two independent reactivity control systems of different design principles shall be provided. The third sentence of GDC 26, 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 specified acceptable fuel design limits are not exceeded. ARDC 26 (2) is consistent with the current requirements of the second reactivity control system specified in GDC 26. The words including xenon burnout have been deleted as this may not be as important for some non-LWR reactor designs.
Also, of different design principles from the first sentence of GDC 26 has been replaced with independent and diverse to clarify the requirement. The reactivity means given by ARDC 26
(2) is a system important to safety but not necessarily safety-related as it does not mitigate an AOO or accident but is used to control planned, normal reactivity changes such that the design limits for the fission product barriers are preserved thereby minimizing challenges to the safety-related reactivity control means or protection system.
The term independent and diverse indicates no shared systems or components and a design which is different enough such that no common failure modes exist between the system or means in ARDC
26 (2) and safety-related systems in ARDC 26 (1) and (3).
(3) 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 Appendix B to RG 1.232, Rev. 0, Page B-15
APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA
III. Reactivity Control Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC
to cool the core is maintained. Reliably controlling reactivity, as applied to GDC 27 requires that the reactor achieve and maintain a safe, stable condition, including subcriticality, using only safety related equipment with margin for stuck rods (ADAMS Accession No. ML16116A083) (Ref. 29).
ARDC 26 (3) is written to clarify that shut down following a postulated accident using safety-related equipment or means is required. The term following a postulated accident refers to the time when plant parameters are relatively stable, no additional means of mitigation are needed and margins to acceptance criteria are constant or increasing. ARDC 26 allows for a return to power during a postulated accident consistent with the current licensing basis of some existing PWRs if sufficient heat removal capability exists (e.g., PWR main steam line break accident), but ARDC 26 (1)
precludes a return to power during an AOO.
(4) The fourth sentence of GDC 26 regarding the capability to reach cold shutdown has been generalized in ARDC 26 (4) to refer to activities which are performed at conditions below (less limiting than) those normally associated with safe shutdown. SECY-94-084 (Ref. 32) describes staff positions on obtaining a cold shutdown and explains that the requirement to bring the plant to cold shutdown is driven by the need to inspect and repair a plant following an accident. In regards to safety class, the capability to bring the plant to a cold shutdown is not covered by the definition of safety-related SSCs in 10 CFR 50.2, and most operating pressurized-water reactors have not credited safety-related SSCs to satisfy this requirement of GDC 26. Based on the information provided above, the system credited for holding the reactor subcritical under conditions necessary for activities such as refueling, inspection and repair is identified as an important to safety system.
Appendix B to RG 1.232, Rev. 0, Page B-16
APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA
III. Reactivity Control Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC
27 Combined reactivity control systems capability.
Same as ARDC
DELETEDInformation incorporated into SFR-DC 26
28 Reactivity limits. Reactor coolant pressure boundary has been relabeled as primary The reactivity control systems shall be designed with coolant boundary to conform to standard terms used in the LMR
appropriate limits on the potential amount and rate of reactivity industry. The use of the term primary indicates that the SFR-DC
increase to ensure that the effects of postulated reactivity are applicable to the primary cooling system, not the intermediate accidents can neither (1) result in damage to the primary coolant cooling system.
boundary greater than limited local yielding nor (2) sufficiently disturb the core, its support structures or other reactor vessel The list of postulated reactivity accidents has been deleted.
internals to impair significantly the capability to cool the core.
29 Protection against anticipated operational occurrences.
Same as GDC
The protection and reactivity control systems shall be designed to assure an extremely high probability of accomplishing their safety functions in the event of anticipated operational occurrences.
IV. Fluid Systems Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC
30 Quality of primary coolant boundary. Reactor coolant pressure boundary has been relabeled as Components that are part of the primary coolant boundary shall primary coolant boundary to conform to standard terms used in be designed, fabricated, erected, and tested to the highest the LMR industry.
quality standards practical. Means shall be provided for detecting and, to the extent practical, identifying the location of The use of the term primary indicates that the SFR-DC are the source of primary coolant leakage. applicable only to the primary cooling system, not the intermediate cooling system.
Appendix B to RG 1.232, Rev. 0, Page B-17
APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA
IV. Fluid Systems Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC
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) (Ref. 4).
31 Fracture prevention of primary coolant boundary. Reactor coolant pressure boundary has been relabeled as The primary coolant boundary shall be designed with sufficient primary coolant boundary to conform to standard terms used in margin to ensure that, when stressed under operating, the LMR industry.
maintenance, testing, and postulated accident conditions,
(1) the boundary behaves in a nonbrittle manner and (2) the The use of the term primary indicates that the SFR-DC are probability of rapidly propagating fracture is minimized. The applicable only to the primary cooling system, not the intermediate design shall reflect consideration of service temperatures, cooling system.
service degradation of material properties, creep, fatigue, stress rupture, and other conditions of the boundary material under The cover gas boundary is included as part of the primary coolant operating, maintenance, testing, and postulated accident boundary (referred to as RCPB by PRISM) per NUREG-1368 conditions and the uncertainties in determining (1) material (page 3-38) (Ref. 4).
properties, (2) the effects of irradiation and coolant composition, including contaminants and reaction products, on Specific examples are added to the SFR-DC to account for the high material properties, (3) residual, steady-state, and transient design and operating temperatures, coolant composition, stresses, and (4) size of flaws. contaminants, and reaction products
32 Inspection of primary coolant boundary. Reactor coolant pressure boundary has been relabeled as Components that are part of the primary coolant boundary shall primary coolant boundary to conform to standard terms used in be designed to permit (1) periodic inspection and functional the LMR industry.
testing of important areas and features to assess their structural and leaktight integrity, and (2) an appropriate material The use of the term primary indicates that the SFR-DC are surveillance program for the reactor vessel. applicable only to the primary cooling system, not the intermediate cooling system.
The cover gas boundary is included as part of the primary coolant boundary (referred to as RCPB by PRISM) per NUREG-1368 (page 3-38) (Ref.4).
Appendix B to RG 1.232, Rev. 0, Page B-18
APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA
IV. Fluid Systems Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC
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.
Functional testing is testing that assesses component and system operational readiness such as required in the ASME
OM Code as incorporated by reference in 10 CFR 50.55a and in Plant Technical Specifications.
33 Primary coolant inventory maintenance. This SFR-DC was retitled as inventory maintenance to provide A system to maintain primary coolant inventory for protection more flexibility for advanced reactor designs.
against small breaks in the primary coolant boundary shall be provided as necessary to ensure that specified acceptable fuel The first sentence is modified so that it ends with ...shall be design limits are not exceeded as a result of primary coolant provided as necessary and is combined with the second sentence inventory loss due to leakage from the primary coolant as necessary to ensure (without the opening phrase, The boundary and rupture of small piping or other small system safety function shall be) to recognize that the inventory components that are part of the boundary. The system shall be control system may be unnecessary for some designs to maintain designed to ensure that the system safety function can be safety functions that ensure fuel design limits are not exceeded.
accomplished using the piping, pumps, and valves used to maintain primary coolant inventory during normal reactor Reactor coolant pressure boundary has been relabeled as operation. primary coolant boundary to 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 indicates that the SFR-DC is applicable to the primary cooling system, not the intermediate cooling system.
Appendix B to RG 1.232, Rev. 0, Page B-19
APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA
IV. Fluid Systems Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC
Both pool- and loop-type SFR designs limit loss of primary coolant so that an inventory adequate to perform the safety function of the residual heat removal system is maintained under operating, maintenance, testing, and postulated accident conditions.
The GDC reference to electric power was removed. Refer to SFR-
DC 17 concerning those systems that require electric power.
34 Residual heat removal. In most advanced reactor designs the residual heat removal system A system to remove residual heat shall be provided. For normal is designed to meet the requirements of SFR-DC 34 and SFR-DC
operations and anticipated operational occurrences, the system 35 (for more discussion see NUREG-0968 (Ref. 5) and NUREG-
safety function shall be to transfer fission product decay heat 1368 (Ref. 4)).
and other residual heat from the reactor core at a rate such that specified acceptable fuel design limits and the design It is anticipated that the residual heat removal system for non- conditions of the primary coolant boundary are not exceeded. LWRs will have the same regulatory treatment as the current LWR
fleet.
Suitable redundancy in components and features and suitable interconnections leak detection, and isolation capabilities, shall Reactor coolant pressure boundary has been relabeled as be provided to ensure that the system safety function can be primary coolant boundary to reflect that the SFR primary system accomplished, assuming a single failure. operates at low-pressure and 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 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.
Appendix B to RG 1.232, Rev. 0, Page B-20
APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA
IV. Fluid Systems Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC
35 Emergency core cooling. In most advanced reactor designs, a single system (i.e. the residual Same as ARDC heat removal system) is provided to perform both the residual heat A system to assure sufficient core cooling during postulated removal and emergency core cooling functions. In this case, the accidents and to remove residual heat following postulated single system would be designed to meet the requirements of SFR-
accidents shall be provided. The system safety function shall be DC 34 and SFR-DC 35. (for more discussion see NUREG-0968 to transfer heat from the reactor core during and following (Ref. 5) and NUREG-1368 (Ref. 4)) However, the staff postulated accidents such that fuel and clad damage that could acknowledges that this may not be the case for every advanced interfere with continued effective core cooling is prevented. reactor design. Therefore, to allow current and future non-LWR
designers the flexibility to provide a single system or multiple Suitable redundancy in components and features and suitable systems to perform residual heat removal and emergency core interconnections, leak detection, isolation, and containment cooling, the staff decided to keep the SFR-DC 34 and SFR-DC 35 capabilities shall be provided to ensure that the system safety separate in lieu of combining them into a single criterion. Effective function can be accomplished, assuming a single failure. 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 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 emergency core cooling system. In most advanced reactor designs, a single system (i.e. the residual Same as ARDC heat removal system) is provided to perform both the residual heat A system that provides emergency core cooling shall be removal and emergency core cooling functions. In this case, the designed to permit appropriate periodic inspection of important single system would be designed to meet the requirements of SFR-
components to ensure the integrity and capability of the system. DC 34 and SFR-DC 35. (for more discussion see NUREG-0968 Appendix B to RG 1.232, Rev. 0, Page B-21
APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA
IV. Fluid Systems Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC
(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 components associated with residual heat removal. This revision allows for a technology-inclusive ARDC.
Review of the proposed DOE SFR and MHTGR 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 emergency core cooling system. In most advanced reactor designs, a single system (i.e., the residual Same as ARDC heat removal system) is provided to perform both the residual heat A system that provides emergency core cooling shall be removal and emergency core cooling functions. In this case, the designed to permit appropriate periodic functional testing to single system would be designed to meet the requirements of SFR-
ensure (1) the structural and leaktight integrity of its DC 34 and SFR-DC 35. (for more discussion see NUREG-0968 Appendix B to RG 1.232, Rev. 0, Page B-22
APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA
IV. Fluid Systems Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC
components, (2) the operability and performance of the system (Ref. 5) and NUREG-1368 (Ref. 4)) However, the staff components, and (3) the operability of the system as a whole acknowledges that this may not be the case for every advanced and, under conditions as close to design as practical, the reactor design. Therefore, to allow current and future non-LWR
performance of the full operational sequence that brings the designers the flexibility to provide a single system or multiple system into operation, including operation of any associated systems to perform residual heat removal and emergency core systems and interfaces necessary to transfer decay heat to the cooling, the staff decided to keep the SFR-DC 34 and SFR-DC 35 ultimate heat sink. separate in lieu of combining them into a single criterion. The staffs approach to provide two separate criteria is consistent with the approach taken in the LWR GDCs.
The SFR-DC has slightly different wording than the GDC to clarify the scope of the criteria. Any system, or portions of a system, credited with an emergency core cooling function during postulated accidents (for example, a system that performs both the residual heat removal function and the emergency core cooling function)
would need to meet SFR-DC 37.
Specific mention of pressure testing has been removed yet remains a potential requirement should it be necessary as a component of appropriate periodic functional testing... of cooling systems.
Functional testing is testing that assesses component and system operational readiness such as required in the ASME
OM Code as incorporated by reference in 10 CFR 50.55a and in Plant Technical Specifications.
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.
Appendix B to RG 1.232, Rev. 0, Page B-23
APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA
IV. Fluid Systems Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC
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, structural and equipment cooling 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. as necessary is meant to condition an SFR-DC 38 Same as ARDC application to designs requiring heat removal for conventional A system to remove heat from the reactor containment shall be containments that are found to require heat removal measures.
provided as necessary to maintain the containment pressure and temperature within acceptable limits following postulated The LOCA reference has been removed to provide for any accidents. postulated accident that might affect the containment structure.
Suitable redundancy in components and features and suitable Containment structure safety system redundancy is addressed in the interconnections, leak detection, isolation, and containment second paragraph.
capabilities shall be provided to ensure that the system safety function can be accomplished, assuming a single failure.
39 Inspection of containment heat removal system. Examples were deleted to make the SFR-DC technology inclusive.
Same as ARDC
The containment heat removal system shall be designed to permit appropriate periodic inspection of important components to ensure the integrity and capability of the system.
40 Testing of containment heat removal system. Specific mention of pressure testing has been removed yet Same as ARDC remains a potential requirement should it be necessary as a Appendix B to RG 1.232, Rev. 0, Page B-24
APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA
IV. Fluid Systems Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC
The containment heat removal system shall be designed to component of appropriate periodic functional testing... of permit appropriate periodic functional testing to ensure (1) the containment heat removal.
structural and leaktight integrity of its components, (2) the operability and performance of the system components, and Functional testing is testing that assesses component and system
(3) the operability of the system as a whole, and under operational readiness such as required in the ASME
conditions as close to the design as practical, the performance OM Code as incorporated by reference in 10 CFR 50.55a and in of the full operational sequence that brings the system into Plant Technical Specifications.
operation, including the operation of associated 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.
Reference to the operation of applicable portions of the protection system, structural and equipment cooling systems, and power transfers is considered part of the more general associated systems for operability testing of the system as a whole.
The GDC reference to electric power was removed. Refer to SFR-
DC17 concerning those systems that require electric power.
41 Containment atmosphere cleanup. Advanced reactors offer potential for reaction product generation Same as ARDC that is different from that associated with clad metal-water Systems to control fission products and other substances that interactions. Therefore, the terms hydrogen and oxygen are may be released into the reactor containment shall be provided removed while other substances is retained to allow for as necessary to reduce, consistent with the functioning of other exceptions.
associated systems, the concentration and quality of fission products released to the environment following postulated Considering that a passive containment cooling system may be accidents and to control the concentration of other substances in used or that the containment may have an additional safety function the containment atmosphere following postulated accidents to other than radionuclide retention, additional wording for ensure that containment integrity and other safety functions are maintaining safety functions is added.
maintained.
Appendix B to RG 1.232, Rev. 0, Page B-25
APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA
IV. Fluid Systems Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC
Each system shall have suitable redundancy in components and features and suitable interconnections, leak detection, isolation, and containment capabilities to ensure that its safety function can be accomplished, assuming a single failure.
42 Inspection of containment atmosphere cleanup systems.
Same as GDC
The containment atmosphere cleanup systems shall be designed to permit appropriate periodic inspection of important components, such as filter frames, ducts, and piping to assure the integrity and capability of the systems.
43 Testing of containment atmosphere cleanup systems. Active has been deleted in item (2), as appropriate operability Same as ARDC and performance testing of system components is required The containment atmosphere cleanup systems shall be designed regardless of an active or passive nature, as are cited examples of to permit appropriate periodic functional testing to ensure active system components.
(1) the structural and leaktight integrity of its components,
(2) the operability and performance of the system components, Examples of active systems under item (2) have been deleted, both and (3) the operability of the systems as a whole and, under to conform to similar wording in ARDC 37 and 40 and ensure that conditions as close to design as practical, the performance of passive as well as active system components are considered.
the full operational sequence that brings the systems into operation, including the operation of associated systems. Specific mention of pressure testing has been removed yet remains a potential requirement should it be necessary as a component of appropriate periodic functional testing... of cooling systems. A non-leaktight system may be acceptable for some designs provided that (1) the system leakage does not impact safety functions under all conditions and (2) defense in depth is not impacted by system leakage.
Functional testing is testing that assesses component and system operational readiness such as required in the ASME
OM Code as incorporated by reference in 10 CFR 50.55a and in Plant Technical Specifications.
Appendix B to RG 1.232, Rev. 0, Page B-26
APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA
IV. Fluid Systems Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC
The GDC reference to electric power was removed. Refer to SFR-
DC17 concerning those systems that require electric power.
44 Structural and equipment cooling. This renamed SFR-DC accounts for advanced reactor system Same as ARDC design differences to include cooling requirements for SSCs, if A system to transfer heat from structures, systems, and applicable; this SFR-DC does not address the residual heat removal components important to safety to an ultimate heat sink shall be system required under SFR-DC 34, and ECCS system under SFR-
provided, as necessary, to transfer the combined heat load of DC 35.
these structures, systems, and components under normal operating and accident conditions. The GDC reference to electric power was removed. Refer to SFR-
DC17 concerning those systems that require electric power.
Suitable redundancy in components and features and suitable interconnections, leak detection, and isolation capabilities shall be provided to ensure that the system safety function can be accomplished, assuming a single failure.
45 Inspection of structural and equipment cooling systems. This renamed ARDC accounts for advanced reactor system design Same as ARDC differences to include possible cooling requirements for SSCs The structural and equipment cooling systems shall be designed important to safety.
to permit appropriate periodic inspection of important components, such as heat exchangers and piping, to ensure the integrity and capability of the systems.
46 Testing of structural and equipment cooling systems. This renamed ARDC accounts for advanced reactor system design Same as ARDC differences to include possible cooling requirements for SSCs The structural and equipment cooling systems shall be designed important to safety. Specific mention of pressure testing has been to permit appropriate periodic functional testing to ensure removed yet remains a potential requirement should it be necessary
(1) the structural and leaktight integrity of their components, as a component of appropriate periodic functional testing... of
(2) the operability and performance of the system components, cooling systems. A non-leaktight system may be acceptable for and (3) the operability of the systems as a whole and, under some designs provided that (1) the system leakage does not impact conditions as close to design as practical, the performance of safety functions under all conditions and (2) defense in depth is not the full operational sequences that bring the systems into impacted by system leakage.
operation for reactor shutdown and postulated accidents, including the operation of associated systems.
Appendix B to RG 1.232, Rev. 0, Page B-27
APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA
IV. Fluid Systems Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC
Functional testing is testing that assesses component and system operational readiness such as required in the ASME
OM Code as incorporated by reference in 10 CFR 50.55a and in Plant Technical Specifications.
Active has been deleted in item (2) because appropriate operability and performance tests of system components are required regardless of their active or passive natur
e. 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.
V. Reactor Containment Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC 50 Containment design basis. SFR-DC 50 specifically addresses a containment structure in the The reactor containment structure, including access openings, opening sentence and SFR-DC 51-57 support the containment penetrations, and the containment heat removal system shall be structures design basis. Therefore, SFR-DC 51-57 are modified by designed so that the containment structure and its internal adding the word structure to highlight the containment structure- compartments can accommodate, without exceeding the design specific criteria.
leakage rate and with sufficient margin, the calculated pressure and temperature conditions resulting from postulated accidents. The phrase loss-of-coolant accident is LWR specific because this This margin shall reflect consideration of (1) the effects of is understood to be the limiting containment structure accident for potential energy sources that have not been included in the an LWR design. It is replaced by the phrase postulated accident determination of the peak conditions, (2) the limited experience to allow for consideration of the design-specific containment and experimental data available for defining accident structure limiting accident for non-LWR designs.
phenomena and containment responses, and (3) the conservatism of the calculational model and input parameters The example at the end of subpart 1 of the GDC is LWR specific and therefore deleted Appendix B to RG 1.232, Rev. 0, Page B-28
APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA
V. Reactor Containment Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC 51 Fracture prevention of containment pressure boundary. SFR-DC 51-57 support SFR-DC 50, which specifically applies to The boundary of the reactor containment structure shall be non-LWR designs that use a fixed containment structure.
designed with sufficient margin to ensure that, under operating, Therefore, the word structure is added to each of these SFR-DC
maintenance, testing, and postulated accident conditions, (1) its to clearly convey the understanding that this criterion applies to materials behave in a nonbrittle manner and (2) the probability designs employing containment structures.
of rapidly propagating fracture is minimized. The design shall The term ferritic was removed to avoid limiting the scope of the reflect consideration of service temperatures and other criterion to ferritic materials. With this revision, the staff believes conditions of the containment boundary materials during that this criterion is more broadly applicable.
operation, maintenance, testing, and postulated accident conditions, and the uncertainties in determining (1) material The word pressure was left in the title to reflect that, while a properties, (2) residual, steady-state, and transient stresses, and design might not have a high-pressure containment like a
(3) size of flaws. traditional LWR, the containment still serves a pressure-retaining function. Refer to the SFR-DC 16 rationale for additional information related to SFR containment pressure.
52 Capability for containment leakage rate testing. SFR-DC 51-57 support SFR-DC 50, which specifically applies to The reactor containment structure and other equipment that non-LWR designs that use a fixed containment structure.
may be subjected to containment test conditions shall be Therefore, the word structure is added to each of these SFR-DC
designed so that periodic integrated leakage rate testing can be to clearly convey the understanding that this criterion applies to conducted to demonstrate resistance at containment design designs employing containment structures.
pressure.
53 Provisions for containment testing and inspection. SFR-DC 51-57 support SFR-DC 50, which specifically applies to The reactor containment structure shall be designed to permit non-LWR designs that use a fixed containment structure.
(1) appropriate periodic inspection of all important areas, such Therefore, the word structure is added to each of these SFR-DC
as penetrations, (2) an appropriate surveillance program, and to clearly convey the understanding that this criterion only applies
(3) periodic testing at containment design pressure of the leak- to designs employing containment structures.
tightness of penetrations that have resilient seals and expansion bellows.
54 Piping systems penetrating containment. SFR-DC 51-57 support SFR-DC 50, which specifically applies to Piping systems penetrating the reactor containment structure non-LWR designs that use a fixed containment structure.
shall be provided with leak detection, isolation, and Therefore, the word structure is added to each of these SFR-DC
containment capabilities that have redundancy, reliability, and Appendix B to RG 1.232, Rev. 0, Page B-29
APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA
V. Reactor Containment Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC
performance capabilities necessary to perform the containment to clearly convey the understanding that this criterion only applies safety function and that reflect the importance to safety of to designs employing containment structures.
preventing radioactivity releases from containment through these piping systems. Such piping systems shall be designed Not all penetrations will provide a release path to the atmosphere.
with the capability to verify, by testing, the operational Piping that may be of interest in the case of an SFR design is for readiness of any isolation valves and associated apparatus the intermediate heat transport system and the residual heat periodically and to confirm that valve leakage is within removal system A designer may be able to satisfactorily acceptable limits. 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) indicates that GDC 54 is 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, for liquid metal reactors.
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.
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.
Appendix B to RG 1.232, Rev. 0, Page B-30
APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA
V. Reactor Containment Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC 55 Primary coolant boundary penetrating containment. SFR-DCs 51-57 support SFR-DC 50, which specifically applies to Each line that is part of the primary coolant boundary and that advanced non-LWR designs that use a fixed containment structure.
penetrates the reactor containment structure shall be provided Therefore, the word structure is added to each of these SFR-DCs with containment isolation valves as follows, unless it can be to clearly convey the understanding that this criterion only applies demonstrated that the containment isolation provisions for a to designs employing containment structures.
specific class of lines, such as instrument lines, are acceptable on some other defined basis: The title of SFR-DC 55 is the Primary coolant boundary
(1) One locked closed isolation valve inside and one locked penetrating containment. The SFR intermediate coolant system is closed isolation valve outside containment; or a separate closed system that does not allow any direct mixing of
(2) One automatic isolation valve inside and one locked closed intermediate fluid with the primary coolant sodium. The tubing of isolation valve outside containment; or the intermediate heat exchanger and associated intermediate
(3) One locked closed isolation valve inside and one automatic coolant system piping are a part of the primary coolant boundary.
isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside SFR-DC 57, Closed system isolation valves, addresses closed containment; or systems that penetrate containment and would be the appropriate
(4) One automatic isolation valve inside and one automatic place to address a closed system, such as an intermediate coolant isolation valve outside containment. A simple check valve system, that penetrates containment and is not part of the primary may not be used as the automatic isolation valve outside coolant boundary (in its entirety). This is similar to the treatment of containment. the main steam system and the steam generator in a pressurized- water reactor.
Isolation valves outside containment shall be located as close to containment as practical and, upon loss of actuating power, Reactor coolant pressure boundary has been relabeled as automatic isolation valves shall be designed to take the position primary coolant boundary to conform to standard terms used in that provides greater safety. the LMR industry.
Other appropriate requirements to minimize the probability or The use of the term primary indicates that the SFR-DC is consequences of an accidental rupture of these lines or of lines applicable to the primary cooling system, not the intermediate connected to them shall be provided as necessary to ensure cooling system.
adequate safety. Determination of the appropriateness of these requirements, such as higher quality in design, fabrication, and The cover gas boundary is included as part of the primary coolant testing, additional provisions for inservice inspection, boundary (referred to as RCPB by PRISM) per NUREG-1368 protection against more severe natural phenomena, and (page 3-38).
additional isolation valves and containment, shall include Appendix B to RG 1.232, Rev. 0, Page B-31
APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA
V. Reactor Containment Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC
consideration of the population density, use characteristics, and physical characteristics of the site environs.
56 Containment isolation. SFR-DC 51-57 support SFR-DC 50, which specifically applies to Each line that connects directly to the containment atmosphere non-LWR designs that use a fixed containment structure.
and penetrates the reactor containment structure shall be Therefore, the word structure is added to each of these SFR-DC
provided with containment isolation valves as follows, unless it to clearly convey the understanding that this criterion only applies can be demonstrated that the containment isolation provisions to designs employing containment structures.
for a specific class of lines, such as instrument lines, are acceptable on some other defined basis: The word primary in the title and the text was removed, and the
(1) One locked closed isolation valve inside and one locked word reactor was also removed because the containment is a closed isolation valve outside containment; or barrier between the fission products and the environment.
(2) One automatic isolation valve inside and one locked closed isolation valve outside containment; or There are diverse advanced reactor designs and, hence, there is no
(3) One locked closed isolation valve inside and one automatic single containment concept. In all cases, the rules for containment isolation valve outside containment. A simple check valve penetrations to fulfill containment isolation would apply. How this may not be used as the automatic isolation valve outside is accomplished should be left to the designer of the particular containment; or advanced reactor design, without being too prescriptive as to
(4) One automatic isolation valve inside and one automatic whether it is a primary or secondary or reactor containment. There isolation valve outside containment. A simple check valve may be a need for a containment structure outside the reactor may not be used as the automatic isolation valve outside region. For example, in the MSR design, some of the liquid fuel containment. salt is drawn off to a processing system to clean it up and remove Isolation valves outside containment shall be located as close to fission products before returning it to the reactor. The liquid fuel the containment as practical and upon loss of actuating power, salt is highly radioactive and would need a containment around the automatic isolation valves shall be designed to take the position entire system. Alternatively, in an SFR, the guard vessel would be that provides greater safety. 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 must also meet the containment isolation requirements.
57 Closed system isolation valves. SFR-DCs 51-57 support SFR-DC 50, which specifically applies to Each line that penetrates the reactor containment structure and advanced non-LWR designs that use a fixed containment structure.
is neither part of the primary coolant boundary nor connected Therefore, the word structure is added to each of these SFR-DCs directly to the containment atmosphere shall have at least one Appendix B to RG 1.232, Rev. 0, Page B-32
APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA
V. Reactor Containment Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC
containment isolation valve unless it can be demonstrated that to clearly convey the understanding that this criterion only applies the containment safety function can be met without an isolation to designs employing containment structures.
valve and assuming failure of a single active component. The isolation valve, if required, shall be either automatic, or locked Reactor coolant pressure boundary has been relabeled as closed, or capable of remote manual operation. This valve shall primary coolant boundary to conform to standard terms used in be outside containment and located as close to the containment the LMR industry.
as practical. A simple check valve may not be used as the automatic isolation valve. The use of the term primary indicates that the SFR-DC is applicable 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).
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
The nuclear power unit design shall include means to control suitably the release of radioactive materials in gaseous and liquid effluents and to handle radioactive solid wastes produced during normal reactor operation, including anticipated operational occurrences. Sufficient holdup capacity shall be provided for retention of gaseous and liquid effluents containing radioactive materials, particularly where unfavorable site environmental conditions can be expected to impose unusual operational limitations upon the release of such effluents to the environment.
Appendix B to RG 1.232, Rev. 0, Page B-33
APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA
VI. Fuel and Reactivity Control Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC 61 Fuel storage and handling and radioactivity control. The underlying concept of establishing functional requirements for Same as ARDC radioactivity control in fuel storage and fuel handling systems is The fuel storage and handling, radioactive waste, and other independent of the design of non-LWR reactors. However, some systems that may contain radioactivity shall be designed to advanced designs may use dry fuel storage that incorporates ensure adequate safety under normal and postulated accident cooling jackets that can be liquid cooled or air cooled to remove conditions. These systems shall be designed (1) with a heat. This modification to this GDC allows for both liquid and air capability to permit appropriate periodic inspection and testing cooling of the dry fuel storage containers.
of components important to safety, (2) with suitable shielding for radiation protection, (3) with appropriate containment, confinement, and filtering systems, (4) with a residual heat removal capability having reliability and testability that reflects the importance to safety of decay heat and other residual heat removal, and (5) to prevent significant reduction in fuel storage cooling under accident conditions.
62 Prevention of criticality in fuel storage and handling.
Same as GDC
Criticality in the fuel storage and handling system shall be prevented by physical systems or processes, preferably by use of geometrically safe configurations.
63 Monitoring fuel and waste storage.
Same as GDC
Appropriate systems shall be provided in fuel storage and radioactive waste systems and associated handling areas (1) to detect conditions that may result in loss of residual heat removal capability and excessive radiation levels and (2) to initiate appropriate safety actions.
64 Monitoring radioactivity releases. In NUREG-1368, Table 3.3 (page 3-25), the NRC staff Means shall be provided for monitoring the reactor containment recommended deleting the GDC 64 phrase spaces containing atmosphere, spaces containing components for primary system components for recirculation of loss-of-coolant accident fluids.
sodium and cover gas cleanup and processing, effluent Otherwise, the NRC staff noted that criterion requirements are discharge paths, and the plant environs for radioactivity that independent of the design of SFRs (page 3-55).
Appendix B to RG 1.232, Rev. 0, Page B-34
APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA
VI. Fuel and Reactivity Control Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC
may be released from normal operations, including anticipated 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. SFR-DC 70, SFR-DC 75, and SFR-DC 76 describe the three functions of the ICS system: (1) to ensure that the ICS does not If an intermediate cooling system is provided, then the impact the safety of the primary coolant system, (2) to ensure that intermediate coolant system shall be designed with sufficient radioactivity in the primary coolant system does not transfer into the margin to assure that (1) the design conditions of the power conversion system, and (3) to ensure that the ICS is designed intermediate coolant boundary are not exceeded during normal to minimize the possibility of a large, uncontrolled release of operations, including anticipated occupational occurrences, and sodium. SFR-DC 77 provides verification that the ICS system can
(2) the integrity of the primary coolant boundary is maintained perform these functions through inspection. NUREG-1368 (Ref. 4)
during postulated accidents. (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.
Appendix B to RG 1.232, Rev. 0, Page B-35
APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA
VII. Additional SFR-DC
Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC
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.
71 Primary coolant and cover gas purity control. The NRC considered DOEs proposed SFR-DC 71 and made changes based on the Response to NRC Staff Questions on the Systems shall be provided as necessary to maintain the purity of U.S. Department of Energy Report, Guidance for Developing primary coolant sodium and cover gas within specified design Principal Design Criteria for Advanced Non-Light Water limits. These limits shall be based on consideration of Reactors (pages 12-13) (Ref. 19).
(1) chemical attack, (2) fouling and plugging of passages, and
(3) radionuclide concentrations, and (4) air or moisture ingress NUREG 1368 (Ref. 4) (page 3-57), Section 3.2.4.6, suggested the as a result of a leak of cover gas. need for a separate criterion for a sodium and cover gas purity control. Also a separate criterion was included in NUREG-0968 (Ref. 5) (Criterion 34, Reactor and Intermediate Coolant and Cover Gas Purity Control).
72 Sodium heating systems. The NRC considered DOEs proposed SFR-DC 72 and made Heating systems shall be provided for systems and components changes based on the Response to NRC Staff Questions on the that are important to safety, and that contain or could be U.S. Department of Energy Report, Guidance for Developing required to contain sodium. These heating systems and their Principal Design Criteria for Advanced Non-Light Water controls shall be appropriately designed to ensure that the Reactors (pages 13-15) (Ref. 19).
temperature distribution and rate of change of temperature in systems and components containing sodium are maintained NUREG-1368 (Ref. 4) (page 3-56), Section 3.2.4.2, suggested the within design limits assuming a single failure. If plugging of need for a separate criterion for sodium heating system. Also, a any cover gas line due to condensation or plate out of sodium separate criterion was included in NUREG-0968 (Ref. 5)
aerosol or vapor could prevent accomplishing a safety function, (Criterion 7, Sodium Heating Systems).
Appendix B to RG 1.232, Rev. 0, Page B-36
APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA
VII. Additional SFR-DC
Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC
the temperature control and the relevant corrective measures The phrase and the relevant corrective measures has been added, associated with that line shall be considered important to safety. in case the cover gas line design includes a feature for clearing an obstruction resulting from condensation or plate out of sodium aerosol or vapor.
73 Sodium leakage detection and reaction prevention and The NRC considered DOEs proposed SFR-DC 73 and made mitigation. changes based on the Response to NRC Staff Questions on the Means to detect and identify sodium leakage as practical and to U.S. Department of Energy Report, Guidance for Developing limit and control the extent of sodium-air and sodium-concrete Principal Design Criteria for Advanced Non-Light Water reactions and to mitigate the effects of fires resulting from these Reactors (pages 15-16) (Ref. 19).
sodium-air and sodium-concrete reactions shall be provided to ensure that the safety functions of structures, systems, and NUREG-1368 (Ref. 4) (page 3-56), Section 3.2.4.1, suggested the components important to safety are maintained. Systems from need for a separate criterion for protection against sodium reactions.
which sodium leakage constitutes a significant safety hazard Also, a separate criterion was included in NUREG-0968 (Ref. 5)
shall include measures for protection, such as inerted enclosures (Criterion 4, Protection against Sodium and NaK reactions).
or guard vessels.
74 Sodium/water reaction prevention/mitigation. The NRC considered DOEs proposed SFR-DC 74 and made Structures, systems, and components containing sodium shall be changes based on the Response to NRC Staff Questions on the designed and located to avoid contact between sodium and U.S. Department of Energy Report, Guidance for Developing water and to limit the adverse effects of chemical reactions Principal Design Criteria for Advanced Non-Light Water between sodium and water on the capability of any structure, Reactors (pages 16-18) (Ref.19)
system, or component to perform any of its intended safety functions. If steam-water is used for energy conversion, to NUREG-1368 (Ref 4) (page 3-56), Section 3.2.4.1, suggested the prevent loss of any plant safety function, the sodium-steam need for a separate criterion for protection against sodium reactions.
generator system shall be designed to detect and contain Also, a separate criterion was included in NUREG-0968 (Ref. 5)
sodium-water reactions and limit the effects of the energy and (Criterion 4, Protection against Sodium and NaK reactions).
reaction products released by such reactions, including mitigation of the effects of any resulting fire involving sodium. Fire considerations are added for consistency with SFR-DC 73.
75 Quality of the intermediate coolant boundary. This criterion is similar to GDC 30 in 10 CFR Part 50, Appendix A,
Components that are part of the intermediate coolant boundary and is intended to ensure that, similar to the reactor coolant pressure shall be designed, fabricated, erected, and tested to quality boundary, the intermediate coolant boundary is designed, Appendix B to RG 1.232, Rev. 0, Page B-37
APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA
VII. Additional SFR-DC
Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC
standards commensurate with the importance of the safety fabricated, and tested using quality standards and controls sufficient functions to be performed. to ensure that failure of the intermediate system would be unlikely.
The statement commensurate with the systems importance to safety clarifies that the staff expects a graded approach to be used in determining the quality requirements for the ICS. While not directly applicable to non-LWRs, RG 1.26 Quality Group Classifications and Standards for Water-, Steam-, and Radioactive- Waste-Containing Components of Nuclear Power Plants (Ref. 39)
provides a basis which can be used to develop a graded quality approach for non-LWR systems including the ICS.
76 Fracture prevention of the intermediate coolant boundary. This criterion addresses the need to maintain a sodium-steam The intermediate coolant boundary shall be designed with generator in a manner that minimizes the potential for system sufficient margin to ensure that, when stressed under operating, failure. This criterion is similar to GDC 31 in 10 CFR Part 50,
maintenance, testing, and postulated accident conditions, (1) the Appendix A, and is intended to ensure that, similar to the reactor boundary behaves in a nonbrittle manner and (2) the probability coolant pressure boundary, the ICS is designed to prevent a rapid of rapidly propagating fracture is minimized. and uncontrolled failure. A sudden rupture of the ICS could result in massive sodium-air, -water, or -concrete reactions and would constitute a risk to the safe operation of the plant and challenge the integrated safety of the plant. This criterion should not be interpreted to preclude the use of rupture discs for controlled, sudden evacuation of the ICS inventory into a vessel or system.
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.
77 Inspection of the intermediate coolant boundary. This criterion is similar to GDC 32 in 10 CFR Part 50, Appendix A,
Components that are part of the intermediate coolant boundary and is intended to ensure that, similar to the reactor coolant pressure shall be designed to permit (1) periodic inspection and boundary, the intermediate coolant boundary is inspected to ensure functional testing of important areas and features to assess their that the system is maintained to the quality standard defined in structural and leaktight integrity commensurate with the SFR-DC 75.
Appendix B to RG 1.232, Rev. 0, Page B-38
APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA
VII. Additional SFR-DC
Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC
systems importance to safety, and (2) an appropriate material A non-leaktight system may be acceptable for some designs surveillance program for the intermediate coolant boundary. 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 The consequence of leakage between the primary coolant system and a heat removal system (i.e. residual heat removal system, When the primary coolant system interfaces with a structure, intermediate coolant system) is more significant for primary coolant system, or component containing fluid that is chemically system (potentially impacting the fuel design limits or integrity of incompatible with the primary coolant, the interface location the primary coolant boundary) than it is for the heat removal system shall be designed to ensure that the primary coolant is separated (coolant drawdown or introduction of radioactive sodium).
from the chemically incompatible fluid by two redundant, passive barriers. When the primary coolant system interfaces Rather than creating two parallel requirements for the two systems, with a structure, system, or component containing fluid that is SFR-DC 78 was created to discuss leakage and required barriers as chemically compatible with the primary coolant, then the a generic criterion. The criterion allows for double walled steam generators, intermediate coolant systems connected to steam power Appendix B to RG 1.232, Rev. 0, Page B-39
APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA
VII. Additional SFR-DC
Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC
interface location may be a single passive barrier provided that system, and systems similar to the PRISM Direct Reactor Auxiliary the following conditions are met: Cooling System (DRACS).
(1) postulated leakage at the interface location does not A paragraph from NUREG 1368 (page 3-41) (Ref. 4) was added result in failure of the intended safety functions of describing the characteristics of the residual heat removal working structures, systems or components important to safety or fluid and its associated operating pressure. This SFR-DC has been result in exceeding the fuel design limits worded to explain that an intermediate coolant system may be used if the primary coolant is not chemically compatible with the energy
(2) the fluid contained in the structure, system, or conversion system coolant.
component is maintained at a higher pressure than the primary coolant during normal operation, anticipated A single passive barrier is adequate defense in depth when the heat operational occurrences, shutdown, and accident removal working fluid is chemically compatible with the primary conditions. 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 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 Appendix B to RG 1.232, Rev. 0, Page B-40
APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA
VII. Additional SFR-DC
Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC
accidents, as well as during normal operating and shutdown conditions.
79 Cover gas inventory maintenance. This criterion is similar to GDC 33 in 10 CFR Part 50, Appendix A
and SFR-DC 33 in this document. GDC 33 and SFR-DC 33 focus A system to maintain cover gas inventory shall be provided as on the effects of primary coolant (sodium) los
s. A leak in a SFR
necessary to ensure that the primary coolant sodium design primary coolant system may expel the cover gas rather than the limits are not exceeded as a result of cover gas loss due to primary coolant. The cover gas in the SFR performs an important to leakage from the primary coolant boundary and rupture of small safety function by protecting the sodium coolant from chemical piping or other small components that are part of the primary reactions. The staff created a new SFR-DC rather than adding the coolant boundary. cover gas in the term primary coolant. The term primary coolant sodium design limits is used to maintain consistent terminology with SFR-DC 71. The primary coolant sodium design limits consider the possibility of interactions between the primary coolant sodium and the primary coolant boundary or the fuel due to changes in the chemistry of the primary coolant sodium. The considerations include the possibility of (1) chemical attack, (2) fouling and plugging of passages, (3) radionuclide concentrations, and (4) air or moisture ingress as a result of a leak of cover gas.
The term as necessary is retained from SFR-DC 33 to permit designer flexibility if leakage of the system does not challenge the design limits of the primary coolant (for instance, an inerted containment filled with Argon).
Reactor coolant pressure boundary has been relabeled as primary coolant boundary to conform to standard terms used in the LMR
industry.
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 to RG 1.232, Rev. 0, Page B-41
APPENDIX B. SODIUM-COOLED FAST REACTOR DESIGN CRITERIA
VII. Additional SFR-DC
Criterion SFR-DC Title and Content NRC Rationale for Adaption to GDC
The cover gas boundary is included as part of the primary coolant boundary (referred to as RCPB by PRISM) per NUREG-1368 (page
3-38) (Ref. 4).
The GDC reference to electric power was removed. Refer to SFR-
DC 17 concerning those systems that require electric power.
Appendix B to RG 1.232, Rev. 0, Page B-42
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).14 These criteria are applicable to MHTGRs. MHTGR refers to the category of HTGRs that use the inherent high temperature characteristics of tristructural isotropic (TRISO) coated fuel particles, graphite moderator, and helium coolant, as well as passive heat removal from a low power density core with a relatively large height-to-diameter ratio within an uninsulated steel reactor vessel. The MHTGR is designed in such a way to ensure that during design basis events (including loss of forced cooling or loss of helium pressure conditions) radionuclides are retained at their source in the fuel and regulatory requirements for offsite dose are met at the exclusion area boundary. 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 IVHeat Transport Systems (Criteria 30-46)
Section VReactor Containment (Criteria 50-57)
14 The technology-specific design criteria were developed using available design information, previous NRC pre- application reviews of the design types, and more recent industry and DOE national laboratory initiatives in these technology areas (see Reference 17). It is the responsibility of the designer or applicant to provide and justify the PDC
for a specific design.
Appendix C to RG 1.232, Rev. 0, Page C-1
Section VIFuel and Radioactivity Control (Criteria 60-64)
Section VIIAdditional MHTGR-DC (Criteria 70-72)
Appendix B to RG 1.232, Rev. 0, Page B-2
APPENDIX C. MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGN CRITERIA
I. Overall Requirements Criterion MHTGR-DC Title and Content NRC Rationale for Adaptions to GDC 1 Quality standards and records.
Same as GDC
Structures, systems, and components important to safety shall be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed. Where generally recognized codes and standards are used, they shall be identified and evaluated to determine their applicability, adequacy, and sufficiency and shall be supplemented or modified as necessary to assure a quality product in keeping with the required safety function. A quality assurance program shall be established and implemented in order to provide adequate assurance that these structures, systems, and components will satisfactorily perform their safety functions. Appropriate records of the design, fabrication, erection, and testing of structures, systems, and components important to safety shall be maintained by or under the control of the nuclear power unit licensee throughout the life of the unit.
2 Design bases for protection against natural phenomena.
Same as GDC
Structures, systems, and components important to safety shall be designed to withstand the effects of natural phenomena such as earthquakes, tornadoes, hurricanes, floods, tsunami, and seiches without loss of capability to perform their safety functions. The design bases for these structures, systems, and components shall reflect: (1) Appropriate consideration of the most severe of the natural phenomena that have been historically reported for the site and surrounding area, with sufficient margin for the limited accuracy, quantity, and period of time in which the historical data have been accumulated, (2)
appropriate combinations of the effects of normal and accident Appendix C to RG 1.232, Rev. 0, Page C-3
APPENDIX C. MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGN CRITERIA
I. Overall Requirements Criterion MHTGR-DC Title and Content NRC Rationale for Adaptions to GDC
conditions with the effects of the natural phenomena and (3)
the importance of the safety functions to be performed.
3 Fire protection. The phrase containing examples where noncombustible and fire- Same as ARDC resistant materials must be used has been broadened to apply to all Structures, systems, and components important to safety shall advanced reactor designs.
be designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and Instead of and, the phrase locations with structures, systems, explosions. Noncombustible and fire-resistant materials shall and components (SSCs) important to safety uses or, which is be used wherever practical throughout the unit, particularly in logically correct in this case.
locations with structures, systems, or components important to safety. Fire detection and fighting systems of appropriate capacity and capability shall be provided and designed to minimize the adverse effects of fires on structures, systems, and components important to safety. Firefighting systems shall be designed to ensure that their rupture or inadvertent operation does not significantly impair the safety capability of these structures, systems, and components.
4 Environmental and dynamic effects design bases. This change removes the light-water reactor (LWR) emphasis on Structures, systems, and components important to safety shall loss-of-coolant accidents (LOCAs) that may not apply to every be designed to accommodate the effects of and to be design. For example, helium is not needed in a MHTGR to remove compatible with the environmental conditions associated with heat from the core during postulated accidents and does not have normal operation, maintenance, testing, and postulated the same importance as water does to LWR designs to ensure that accidents. These structures, systems, and components shall be fuel integrity is maintained. Therefore, a specific reference to appropriately protected against dynamic effects, including the LOCAs is not applicable to all designs. LOCAs may still require effects of missiles originating both inside and outside the analysis in conjunction with postulated accidents if they are reactor helium pressure boundary, pipe whipping, and relevant to the design.
discharging fluids, that may result from equipment failures and from events and conditions outside the nuclear power unit. If an MHTGR design proposes using a direct power cycle in which However, dynamic effects associated with postulated pipe one or more very high-speed, very high-energy gas turbines are ruptures in nuclear power units may be excluded from the located inside the reactor helium pressure boundary. The presence design basis when analyses reviewed and approved by the of one or more very high-energy turbines inside the primary Commission demonstrate that the probability of fluid system helium pressure boundary creates the potential that a catastrophic Appendix C to RG 1.232, Rev. 0, Page C-4
APPENDIX C. MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGN CRITERIA
I. Overall Requirements Criterion MHTGR-DC Title and Content NRC Rationale for Adaptions to GDC
piping rupture is extremely low under conditions consistent dynamic failure of the gas turbine (e.g., at power) could result in with the design basis for the piping. the consequential catastrophic failure of the primary system pressure boundary caused by the failure of rotating turbine components. To account for the possibility of an MHTGR design that locates high-energy gas turbines inside the reactor helium pressure boundary, the MHTGR-DC language in the area of prevention, protection, and mitigation of turbine dynamic failure is strengthened to support such a power conversion system design approach.
5 Sharing of structures, systems, and components.
Same as GDC
Structures, systems, and components important to safety shall not be shared among nuclear power units unless it can be shown that such sharing will not significantly impair their ability to perform their safety functions, including, in the event of an accident in one unit, an orderly shutdown and cooldown of the remaining units.
II. Multiple Barriers Criterion MHTGR-DC Title and Content NRC Rationale for Adaptions to GDC 10 Reactor design. The concept of specified acceptable fuel design limits, which The reactor system and associated heat removal, control, and prevent additional fuel failures during anticipated operational protection systems shall be designed with appropriate margin to occurrences (AOOs), has been replaced with that of the specified ensure that specified acceptable system radionuclide release acceptable system radionuclide release design limits (SARRDL),
design limits are not exceeded during any condition of normal which limits the amount of radionuclide inventory that is released operation, including the effects of anticipated operational by the system under normal and AOO conditions. The term occurrences. system refers to the fuel, the helium coolant circuit and all connected systems that are not isolated and may contribute to dose.
Appendix C to RG 1.232, Rev. 0, Page C-5
APPENDIX C. MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGN CRITERIA
II. Multiple Barriers Criterion MHTGR-DC Title and Content NRC Rationale for Adaptions to GDC
Design features within the reactor system must ensure that the SARRDLS are not exceeded during normal operations and AOOs.
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 Appendix C to RG 1.232, Rev. 0, Page C-6
APPENDIX C. MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGN CRITERIA
II. Multiple Barriers Criterion MHTGR-DC Title and Content NRC Rationale for Adaptions to GDC
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.
The word coolant has been replaced with heat removal, as helium coolant inventory control for normal operation and AOOs is not necessary to meet the SARRDLs, due to the reactor system design.
11 Reactor inherent protection. The wording has been changed to broaden the applicability from Same as ARDC coolant systems to additional factors (including structures or The reactor core and associated systems that contribute to other fluids) that may contribute to reactivity feedback. These reactivity feedback shall be designed so that, in the power systems are to be designed to compensate for rapid reactivity operating range, the net effect of the prompt inherent nuclear increase.
feedback characteristics tends to compensate for a rapid increase in reactivity.
12 Suppression of reactor power oscillations. Helium in the MHTGR does not affect reactor core susceptibility to The reactor core and associated control and protection systems coolant-induced power oscillations; therefore, a separate MHTGR-
shall be designed to ensure that power oscillations that can specific DC is appropriate. The word coolant was deleted and the result in conditions exceeding specified acceptable system specified acceptable fuel design limits were replaced by radionuclide release design limits are not possible or can be SARRDLs. The discussion on the SARRDL is given in reliably and readily detected and suppressed. MHTGR-DC 10.
13 Instrumentation and control. Reactor coolant pressure boundary has been relabeled as reactor Instrumentation shall be provided to monitor variables and helium pressure boundary to conform to standard terms used for systems over their anticipated ranges for normal operation, for MHTGRs.
anticipated operational occurrences, and for accident conditions, as appropriate, to ensure adequate safety, including The criterion has been modified to reflect the use of the MHTGR
those variables and systems that can affect the fission process functional containment. See the MHTGR-DC 16 rationale.
and the integrity of the reactor core, reactor helium pressure boundary, and functional containment. Appropriate controls shall be provided to maintain these variables and systems within prescribed operating ranges.
Appendix C to RG 1.232, Rev. 0, Page C-7
APPENDIX C. MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGN CRITERIA
II. Multiple Barriers Criterion MHTGR-DC Title and Content NRC Rationale for Adaptions to GDC 14 Reactor helium pressure boundary. Reactor coolant pressure boundary has been relabeled as reactor The reactor helium pressure boundary shall be designed, helium pressure boundary to conform to standard terms used for fabricated, erected, and tested so as to have an extremely low MHTGRs.
probability of abnormal leakage, of rapidly propagating failure, of gross rupture, and of unacceptable ingress of moisture, air, The MHTGR-DC 14 addresses the need to consider leakage of secondary coolant, or other fluids. contaminants into the helium used to transport heat from the reactor to the heat exchangers for power production, residual heat removal, and process heat. The phrase reactor helium pressure boundary encompasses the entire volume containing helium used to cool the reactor, not just the volume within the reactor vessel.
For consistency, a specific requirement is appended to MHTGR-DC 30 for a means of detecting ingress of moisture, air, secondary coolant, or other fluids. Although other fluids could be interpreted as including water and steam, for emphasis, the word moisture is included in the list of contaminants in both MHTGR-DC 14 and MHTGR-DC 30.
15 Reactor helium pressure boundary design. Reactor coolant system has been relabeled as reactor helium All systems that are part of the reactor helium pressure pressure boundary to conform to standard terms used for boundary, such as the reactor system, vessel system, and heat MHTGRs.
removal systems, and the associated auxiliary, control, and protection systems, shall be designed with sufficient margin to ensure that the design conditions of the reactor helium pressure boundary are not exceeded during any condition of normal operation, including anticipated operational occurrences.
16 Containment design. The term functional containment is applicable to advanced A reactor functional containment, consisting of multiple non-LWRs without a pressure retaining containment structure.
barriers internal and/or external to the reactor and its cooling A functional containment can be defined as a barrier, or set of system, shall be provided to control the release of radioactivity barriers taken together, that effectively limit the physical transport to the environment and to ensure that the functional and release of radionuclides to the environment across a full range containment design conditions important to safety are not of normal operating conditions, AOOs, and accident conditions.
exceeded for as long as postulated accident conditions require.
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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 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 Next Generation Nuclear Plant Assessment of Key Licensing Issues (Ref. 11 Enclosure 1), 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 Appendix C to RG 1.232, Rev. 0, Page C-9
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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 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. The electric power systems are required to provide reliable power Electric power systems shall be provided when required to for SSCs during anticipated operational occurrences or postulated permit functioning of structures, systems, and components. The accident conditions when those SSCs safety functions require safety function for each power system shall be to provide electric power. The safety functions are established by the safety sufficient capacity and capability to ensure that (1) that the analyses (i.e. design basis accidents). Where electric power is specified acceptable system radionuclide release design limits needed for anticipated operational occurrences or postulated and the reactor helium pressure boundary design limits are not accidents, the electric power systems shall be sufficient in capacity exceeded as a result of anticipated operational occurrences and and capability to ensure that safety functions as well as important
(2) safety functions that rely on electric power are maintained to safety functions are maintained. The electric power systems in the event of postulated accidents. provide redundancy and defense-in-depth since there would be a minimum of two power systems.
The electric power systems shall include an onsite power system and an additional power system. The onsite electric Compared to GDC 17, more emphasis is placed herein on requiring power system shall have sufficient independence, redundancy, reliability of the overall power supply scheme rather than fully and testability to perform its safety functions, assuming a single prescribing how such reliability can be attained. For example, failure. An additional power system shall have sufficient reference to offsite electric power systems was deleted to provide independence and testability to perform its safety function. for those reactor designs that do not depend on offsite power for the functioning of SSCs important to safety or do not connect to a If electric power is not needed for anticipated operational power grid.
occurrences or postulated accidents, the design shall demonstrate that power for important to safety functions is The onsite power system is envisioned as a fully Class 1E power provided. system and the additional power system is left to the discretion of the designer as long as it meets the performance criteria in Appendix C to RG 1.232, Rev. 0, Page C-10
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paragraph one and the design criteria of paragraph two. For example, the additional independent power source could be from the electrical grid, a diesel generator, a combustion gas turbine or some other alternative, again, at the discretion of the designer.
In this context, important to safety functions refer to the broader, potentially non-safety related functions such as e post-accident monitoring, control room habitability, emergency lighting, radiation monitoring, communications and/or any others that may be deemed appropriate for the given design. The electric power system for important to safety functions could be non-Class 1E and would not be required to have redundant power sources.
Reactor coolant pressure boundary has been relabeled as reactor helium pressure boundary to conform to standard terms used for MHTGRs.
18 Inspection and testing of electric power systems. ARDC 18 is a design-independent companion criterion to ARDC
Same as ARDC 17.
Electric power systems important to safety shall be designed to permit appropriate periodic inspection and testing of important Wording pertaining to additional system examples has been deleted areas and features, such as wiring, insulation, connections, and to allow increased flexibility associated with various designs.
switchboards, to assess the continuity of the systems and the Specifically, the text related to the nuclear power unit, offsite condition of their components. The systems shall be designed power system, and onsite power system was deleted to be with a capability to test periodically (1) the operability and consistent with ARDC 17.
functional performance of the components of the systems, such as onsite power sources, relays, switches, and buses, and (2) the operability of the systems as a whole and, under conditions as close to design as practical, the full operation sequence that brings the systems into operation, including operation of applicable portions of the protection system, and the transfer of power among systems.
Appendix C to RG 1.232, Rev. 0, Page C-11
APPENDIX C. MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGN CRITERIA
II. Multiple Barriers Criterion MHTGR-DC Title and Content NRC Rationale for Adaptions to GDC 19 Control room. ARDC 19 preserves the language of GDC 19 which states (with Same as ARDC emphasis added) A control room shall be provided from which A control room shall be provided from which actions can be actions can be taken to operate the nuclear power unit safely taken to operate the nuclear power unit safely under normal However some clarification of this language is warranted.
conditions and to maintain it in a safe condition under accident conditions. Adequate radiation protection shall be provided to It is clear from this language that there is a need to for operators to permit access and occupancy of the control room under be able to take actions to control the plant. Therefore, designers accident conditions without personnel receiving radiation must consider how the design of controls support safe operator exposures in excess of 5 rem total effective dose equivalent as actions. In addition, NRC staff recognize that in order for defined in § 50.2 for the duration of the accident. operators to act safely as stated in ARDC, that operators must have certain knowledge about the status of the plant and be able to Adequate habitability measures shall be provided to permit make decisions about the appropriate course of action given a access and occupancy of the control room during normal particular operating circumstance. Therefore, these cognitive needs operations and under accident conditions. Equipment at of operators should also be considered by designers when appropriate locations outside the control room shall be provided interpreting ARDC 19.
(1) with a design capability for prompt hot shutdown of the reactor, including necessary instrumentation and controls to This consideration should be reflected in the design of indications, maintain the unit in a safe condition during hot shutdown, and displays, alarms, controls or other future technologies which are
(2) with a potential capability for subsequent cold shutdown of used to inform operators of plant status and may be used to support the reactor through the use of suitable procedures. the decision making process (such as computer based procedures).
This position is consistent with 10 CFR 50.34(f)(2)(iii) which describes the contents required in applications for construction permits. Amongst many other requirements, this rule indicates that the control room design must reflect state-of-the-art human factors principles. These state-of-the-art principles inherently consider both the cognitive and physical aspects of operator action as described above.
The criterion was updated to remove specific emphasis on LOCAs, which may be not appropriate for advanced designs such as the MHTGR.
Appendix C to RG 1.232, Rev. 0, Page C-12
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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.
An additional control room habitability requirement has been proposed. It addresses a new concern: accidents that are not radiological in nature 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.
III. Reactivity Control Criterion MHTGR-DC Title and Content NRC Rationale for Adaptions to GDC 20 Protection system functions. Specified acceptable fuel design limits has been replaced with The protection system shall be designed (1) to initiate SARRDLs. The concept of using SARRDLs is discussed in automatically the operation of appropriate systems, including MHTGR-DC 10. The quantitative value of the SARRDL will be the reactivity control systems, to ensure that the specified design specific. The protection aspect of automatic operation, to acceptable system radionuclide release design limits is not protect normal operation and AOO limits, to sense accident exceeded as a result of anticipated operational occurrences and conditions, and to initiate mitigating equipment has been preserved.
(2) to sense accident conditions and to initiate the operation of systems and components important to safety.
21 Protection system reliability and testability.
Same as GDC
The protection system shall be designed for high functional reliability and inservice testability commensurate with the safety functions to be performed. Redundancy and independence designed into the protection system shall be Appendix C to RG 1.232, Rev. 0, Page C-13
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sufficient to assure that (1) no single failure results in loss of the protection function and (2) removal from service of any component or channel does not result in loss of the required minimum redundancy unless the acceptable reliability of operation of the protection system can be otherwise demonstrated. The protection system shall be designed to permit periodic testing of its functioning when the reactor is in operation, including a capability to test channels independently to determine failures and losses of redundancy that may have occurred.
22 Protection system independence.
Same as GDC
The protection system shall be designed to assure that the effects of natural phenomena, and of normal operating, maintenance, testing, and postulated accident conditions on redundant channels do not result in loss of the protection function, or shall be demonstrated to be acceptable on some other defined basis. Design techniques, such as functional diversity or diversity in component design and principles of operation, shall be used to the extent practical to prevent loss of the protection function.
23 Protection system failure modes.
Same as GDC
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, pressure, steam, water, and radiation) are experienced.
24 Separation of protection and control systems.
Same as GDC
Appendix C to RG 1.232, Rev. 0, Page C-14
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The protection system shall be separated from control systems to the extent that failure of any single control system component or channel, or failure or removal from service of any single protection system component or channel which is common to the control and protection systems leaves intact a system satisfying all reliability, redundancy, and independence requirements of the protection system. Interconnection of the protection and control systems shall be limited so as to assure that safety is not significantly impaired.
25 Protection system requirements for reactivity control Specified acceptable fuel design limits is replaced with malfunctions. SARRDLs. The concept of using SARRDLs is discussed in The protection system shall be designed to ensure that specified MHTGR-DC 10.
acceptable system radionuclide release design limits are not exceeded during any anticipated operational occurrence, accounting for a single malfunction of the reactivity control systems.
26 Reactivity control systems. Recent licensing activity, associated with the application of GDC
A minimum of two reactivity control systems or means shall 26 and GDC 27 to new reactor designs (ADAMS Accession Nos.
provide: ML16116A083 (Ref. 29) and ML16292A589) (Ref. 30), revealed that additional clarity could be provided in the area of reactivity
(1) A means of inserting negative reactivity at a sufficient rate control requirements. MHTGR-DC 26 combines the scope of and amount to assure, with appropriate margin for GDC 26 and GDC 27. The development of MHTGR-DC 26 is malfunctions, that the specified acceptable system radionuclide informed by the proposed general design criteria of 1965 (AEC-R
release design limits and the reactor helium pressure boundary 2/49, November 5), 1967 (32 FR 10216) (Ref. 31), current GDC 26 design limits are not exceeded and safe shutdown is achieved and 27, the definition of safety-related SSC in 10 CFR 50.2, and maintained during normal operation, including anticipated SECY-94-084, Policy and Technical Issues Associated with the operational occurrences. Regulatory Treatment of Non-Safety Systems in Passive Plant Designs (Ref. 32), and the prior application of reactivity control
(2) A means which is independent and diverse from the requirements.
other(s), shall be capable of controlling the rate of reactivity changes resulting from planned, normal power changes to (1) Currently the second sentence of GDC 26 states, that one of the assure that the specified acceptable system radionuclide release reactivity control systems shall use control rods and shall be Appendix C to RG 1.232, Rev. 0, Page C-15
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design limits and the reactor helium pressure boundary design capable of reliably controlling reactivity changes to ensure that, limits are not exceeded. under conditions of normal operation, including AOOs, and with appropriate margin for malfunctions such as stuck rods, specified
(3) A means of inserting negative reactivity at a sufficient rate acceptable fuel design limits are not exceeded. The staff recognizes and amount to assure, with appropriate margin for that specifying control rods may not be suitable for advanced malfunctions, that the capability to cool the core is maintained reactors. Additionally, reliably controlling reactivity, as applied to and a means of shutting down the reactor and maintaining, at a GDC 26, has been interpreted as ensuring the control rods are minimum, a safe shutdown condition following a postulated capable of rapidly (i.e., within a few seconds) shutting down the accident. reactor (ADAMS Accession No. ML16292A589) (Ref. 30).
(4) A means for holding the reactor shutdown under The staff changed control rods to means in recognition that conditions which allow for interventions such as fuel loading, advanced reactor designs may not rely on control rods to rapidly inspection and repair shall be provided. shut down the reactor (e.g., alternative system designs or inherent feedback mechanisms may be relied upon to perform this function).
The wording of reliably controlling reactivity in GDC 26 has been replaced with inserting negative reactivity at a sufficient rate and amount to more clearly define the requirement. For a non- LWR design the rate of negative reactivity insertion may not necessitate rapid (within seconds) insertion but should occur in a time frame such that the fission product barrier design limits are not exceeded.
The term design limits for fission product barriers is replaced with specified acceptable system radionuclide release design limits (SARRDLs) to be consistent with the AOO acceptance criteria associated with MHTGR-DC 10 (SARRDL) and MHTGR-
DC 15 (helium pressure boundary).
The wording safe shutdown is achieved and maintained has been added again to more clearly define the requirements associated with reliably controlling reactivity in GDC 26. SECY-
94-084, Policy and Technical Issues Associated with the Regulatory Treatment of Non-Safety Systems in Passive Plant Appendix C to RG 1.232, Rev. 0, Page C-16
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Designs (Ref. 32), describes the characteristics of a safe shutdown condition as reactor subcriticality, decay heat removal, and radioactive materials containment. MHTGR-DC 26 (1) clearly defines that reactor shutdown at any time during the transient is the performance requirement. The distinction between during and following the transient is discussed in (2) below.
In regards to safety class, the capability to insert negative reactivity at a rate and amount to preserve the fission product barrier(s) and to shut down the reactor during an AOO is identified as a function performed by safety-related SSCs in the 10 CFR 50.2 definition of safety-related SSCs.
(2) The first sentence of GDC 26, states that two independent reactivity control systems of different design principles shall be provided. The third sentence of GDC 26, 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 specified acceptable fuel design limits are not exceeded. MHTGR-DC 26 (2) is consistent with the current requirements of the second reactivity control system specified in GDC 26. The words including xenon burnout have been deleted as this may not be as important for some non- LWR reactor designs. Also, of different design principles from the first sentence of GDC 26 has been replaced with independent and diverse to clarify the requirement. The reactivity means given by MHTGR-DC 26 (2) is a system important to safety but not necessarily safety-related as it does not mitigate an AOO or accident but is used to control planned, normal reactivity changes such that SARRDLs and the helium pressure boundary design limits are preserved thereby minimizing challenges to the safety- related reactivity control means or protection system.
Appendix C to RG 1.232, Rev. 0, Page C-17
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The term independent and diverse indicates no shared systems or components and a design which is different enough such that no common failure modes exist between the system or means in MHTGR-DC 26 (2) and safety-related systems in MHTGR-DC 26
(1) and (3).
(3) 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 applied to GDC 27 requires that the reactor achieve and maintain a safe, stable condition, including subcriticality, using only safety related equipment with margin for stuck rods (ADAMS
Accession No. ML16116A083) (Ref. 29).
MHTGR-DC 26 (3) is written to clarify that shut down following a postulated accident using safety-related equipment or means is required. The term following a postulated accident refers to the time when plant parameters are relatively stable, no additional means of mitigation are needed and margins to acceptance criteria are constant or increasing. MHTGR-DC 26 allows for a return to power during a postulated accident consistent with the current licensing basis of some existing PWRs if sufficient heat removal capability exists (e.g., PWR main steam line break accident), but MHTGR-DC 26 (1) precludes a return to power during an AOO.
(4) The fourth sentence of GDC 26 regarding the capability to reach cold shutdown has been generalized in MHTGR-DC 26 (4)
to refer to activities which are performed at conditions below (less limiting than) those normally associated with safe shutdown.
SECY-94-084 (Ref. 32) describes staff positions on obtaining a cold shutdown and explains that the requirement to bring the plant Appendix C to RG 1.232, Rev. 0, Page C-18
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to cold shutdown is driven by the need to inspect and repair a plant following an accident. In regards to safety class, the capability to bring the plant to a cold shutdown is not covered by the definition of safety-related SSCs in 10 CFR 50.2, and most operating pressurized-water reactors have not credited safety-related SSCs to satisfy this requirement of GDC 26. Based on the information provided above, the system credited for holding the reactor subcritical under conditions necessary for activities such as refueling, inspection and repair is identified as an important to safety system.
27 Combined reactivity control systems capability.
Same as ARDC
DELETEDInformation incorporated into MHTGR-DC 26
28 Reactivity limits. Reactor coolant pressure boundary has been relabeled as reactor The reactor core, including the reactivity control systems, shall helium pressure boundary to conform to standard terms used for be designed with appropriate limits on the potential amount and MHTGRs.
rate of reactivity increase to ensure that the effects of postulated reactivity accidents can neither (1) result in damage to the The list of postulated reactivity accidents has been deleted. Each reactor helium pressure boundary greater than limited local design will have to determine its postulated reactivity accidents yielding, nor (2) sufficiently disturb the core, its support based on the specific design and associated risk evaluation.
structures, or other reactor vessel internals to impair significantly the capability to cool the core.
29 Protection against anticipated operational occurrences.
Same as GDC
The protection and reactivity control systems shall be designed to assure an extremely high probability of accomplishing their Appendix C to RG 1.232, Rev. 0, Page C-19
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safety functions in the event of anticipated operational occurrences.
IV. Heat Transport Systems Criterion MHTGR-DC Title and Content NRC Rationale for Adaptions to GDC 30 Quality of reactor helium pressure boundary. Reactor coolant pressure boundary has been relabeled as reactor Components that are part of the reactor helium pressure helium pressure boundary to conform to standard terms used for boundary shall be designed, fabricated, erected, and tested to MHTGRs.
the highest quality standards practical. Means shall be provided for detecting and, to the extent practical, identifying the The MHTGR-DC 14 addresses the need to consider leakage of location of the source of reactor helium leakage. Means shall be contaminants into the helium used to transport heat from the reactor provided for detecting ingress of moisture, air, secondary to the heat exchangers for power production, residual heat removal, coolant, or other fluids to within the reactor helium pressure and process heat. The phrase reactor helium pressure boundary boundary. encompasses the entire volume containing helium used to cool the reactor, not just the volume within the reactor vessel. For consistency, a specific requirement is appended to MHTGR-DC 30
for a means of detecting ingress of moisture, air, secondary coolant, or other fluids. Although other fluids could be interpreted as including water and steam, for emphasis, the word moisture is included in the list of contaminants in both MHTGR-DC 14 and MHTGR-DC 30.
31 Fracture prevention of reactor helium pressure boundary. Reactor coolant pressure boundary has been relabeled as reactor The reactor helium pressure boundary shall be designed with helium pressure boundary to conform to standard terms used for sufficient margin to ensure that, when stressed under operating, MHTGRs.
maintenance, testing, and postulated accident conditions,
(1) the boundary behaves in a nonbrittle manner and (2) the Specific examples are added to the MHTGR DC to account for the probability of rapidly propagating fracture is minimized. The high design and operating temperatures, helium coolant, design shall reflect consideration of service temperatures, contaminants, and reaction products.
service degradation of material properties, creep, fatigue, stress Appendix C to RG 1.232, Rev. 0, Page C-20
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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 helium composition, including contaminants and reaction products, on material properties, (3) residual, steady-state, and transient stresses, and (4) size of flaws.
32 Inspection of reactor helium pressure boundary. Reactor coolant pressure boundary has been relabeled as reactor Components that are part of the reactor helium pressure helium pressure boundary to conform to standard terms used for boundary shall be designed to permit (1) periodic inspection MHTGRs.
and functional testing of important areas and features to assess their structural and leaktight integrity, and (2) an appropriate The staff modified the LWR GDC by replacing the term reactor material surveillance program for the reactor vessel. pressure vessel with reactor vessel, which the staff believes is a more generically applicable term.
A non-leaktight system may be acceptable for some designs provided that (1) the system leakage does not impact safety functions under all conditions, and (2) leakage is consistent with SARRDL.
Functional testing is testing that assesses component and system operational readiness such as required in the ASME
OM Code as incorporated by reference in 10 CFR 50.55a and in Plant Technical Specifications.
33 Reactor coolant makeup. The MHTGR does not require reactor coolant inventory Not applicable to MHTGR. maintenance for small leaks to meet the SARRDLs, which replaces the concept of the specified acceptable fuel design limits, as discussed in GDC 10. Therefore, ARDC 33 is not applicable to the MHTGR design.
Appendix C to RG 1.232, Rev. 0, Page C-21
APPENDIX C. MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGN CRITERIA
IV. Heat Transport Systems Criterion MHTGR-DC Title and Content NRC Rationale for Adaptions to GDC 34 Passive residual heat removal.
A passive system to remove residual heat shall be provided. For The word passive was added, based on the definition of a normal operations and anticipated operational occurrences, the MHTGR. In definitions Section 3.1 of the DOE report titled system safety function shall be to transfer fission product decay Guidance for Developing Principal Design Criteria for Advanced heat and other residual heat from the reactor core to an ultimate (Non-Light-Water) Reactors (Ref. 17), the MHTGR design has a heat sink at a rate such that specified acceptable system low power density and hence residual heat is removed by a passive radionuclide release design limits and the design conditions of system.
the reactor helium pressure boundary are not exceeded. Ultimate heat sink has been added to explain that, if MHTGR-DC 44 is deemed not applicable to the design, the During postulated accidents, the system safety function shall residual heat removal system is then required to provide the heat provide effective cooling. removal path to the ultimate heat sink.
Suitable redundancy in components and features and suitable Reactor coolant pressure boundary has been relabeled as reactor interconnections, leak detection, and isolation capabilities shall helium pressure boundary to conform to standard terms used for be provided to ensure the system safety function can be MHTGRs.
accomplished, assuming a single failure.
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 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 the integrity of the core, the core structural components, and the reactor vessel is maintained under postulated accident conditions, thereby ensuring a geometry required for passive heat removal.
The GDC reference to electric power was removed. Refer to the rationale for ARDC 17 on electric power systems.
Appendix C to RG 1.232, Rev. 0, Page C-22
APPENDIX C. MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGN CRITERIA
IV. Heat Transport Systems Criterion MHTGR-DC Title and Content NRC Rationale for Adaptions to GDC 35 Emergency core cooling. In the MHTGR design maintaining the helium inventory is not Not applicable to MHTGR. necessary to maintain effective cooling. Postulated accident heat removal is accomplished by the residual heat removal system described in MHTGR DC 34.
36 Inspection of passive residual heat removal system. The word passive was added, based on the definition of a MHTGR. In definitions Section 3.1 of DOE report titled Guidance The passive residual heat removal system shall be designed to for Developing Principal Design Criteria for Advanced (Non- permit appropriate periodic inspection of important components Light-Water) Reactors (Ref. 17), the MHTGR design has a low to ensure the integrity and capability of the system. power density and hence residual heat is removed by a passive system.
The GDC 36 system is renamed and revised to provide for inspection of the residual heat removal systems as required for MHTGR-DC 34.
The list of examples was deleted, as they apply to LWR designs and each specific design will have different important components associated with residual heat removal.
37 Testing of passive residual heat removal system. Criterion 37 has been renamed and revised for testing the passive residual heat removal system required by MHTGR-DC 34.
The passive residual heat removal system shall be designed to permit appropriate periodic functional testing to ensure (1) the Section 2.3.4 of INL/EXT-10-17997, Mechanistic Source Terms structural and leaktight integrity of its components, (2) the White Paper, (Ref. 37) notes that the passive reactor cavity operability and performance of the system components, and cooling system (RCCS) (using either air or water as heat transfer
(3) the operability of the system as a whole and, under fluid) contributes to the MHTGR safety basis and is subject to conditions as close to design as practical, the performance of component integrity testing. However, Section 6.1 of the full operational sequence that brings the system into INL/EXT-11-22708, Modular HTGR Safety Basis and operation, including associated systems, for AOO or postulated Approach, (Ref. 38), indicates that RCCS performance does not accident decay heat removal to the ultimate heat sink and, if require leaktight conditions. For an RCCS which is an open applicable, any system(s) necessary to transition from active system, the normal and expected loss of RCCS coolant through normal operation to passive mode. the exhaust structure would not be considered leakage. Abnormal leakage of RCCS coolant to locations other than the exhaust Appendix C to RG 1.232, Rev. 0, Page C-23
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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.
Functional testing is testing that assesses component and system operational readiness such as required in the ASME
OM Code as incorporated by reference in 10 CFR 50.55a and in Plant Technical Specifications.
The criterion was modified to reflect the passive nature of the MHTGR RCCS to mitigate AOOs or postulated accidents and the need to verify the ability to transition the RCCS from active mode (if present) to passive mode. 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.
Associated systems means testing any auxiliary or secondary systems needed to perform the passive residual heat removal function.
38 Containment heat removal. This criterion is not applicable to the MHTGR. The MHTGR
Not applicable to MHTGR. designs do not have a pressure retaining reactor containment structure but instead rely on a multibarrier functional containment configuration to control the release of radionuclides. See the MHTGR DC 16 rationale.
39 Inspection of containment heat removal system. This criterion is not applicable to the MHTGR. The MHTGR
Not applicable to MHTGR. designs do not have a pressure retaining reactor containment structure but instead rely on a multibarrier functional containment configuration to control the release of radionuclides. See the MHTGR-DC 16 rationale.
Appendix C to RG 1.232, Rev. 0, Page C-24
APPENDIX C. MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGN CRITERIA
IV. Heat Transport Systems Criterion MHTGR-DC Title and Content NRC Rationale for Adaptions to GDC 40 Testing of containment heat removal system. This criterion is not applicable to the MHTGR. The MHTGR
Not applicable to MHTGR. designs do not have a pressure retaining reactor containment structure but instead rely on a multibarrier functional containment configuration to control the release of radionuclides. See the MHTGR-DC 16 rationale.
41 Containment atmosphere cleanup. This criterion is not applicable to the MHTGR. The MHTGR
Not applicable to MHTGR. designs do not have a pressure retaining reactor containment structure but instead rely on a multibarrier functional containment configuration to control the release of radionuclides. See the MHTGR-DC 16 rationale.
42 Inspection of containment atmosphere cleanup systems. This criterion is not applicable to the MHTGR. The MHTGR
Not applicable to MHTGR. designs do not have a pressure retaining reactor containment structure but instead rely on a multibarrier functional containment configuration to control the release of radionuclides. See the MHTGR-DC 16 rationale.
43 Testing of containment atmosphere cleanup systems. This criterion is not applicable to the MHTGR. The MHTGR
Not applicable to MHTGR. designs do not have a pressure retaining reactor containment structure but instead rely on a multibarrier functional containment configuration to control the release of radionuclides. See the MHTGR-DC 16 rationale.
44 Structural and equipment cooling. This renamed MHTGR-DC accounts for advanced reactor system In addition to the heat rejection capability of the passive design differences to include cooling requirements for SSCs residual heat removal system, systems to transfer heat from important to safety, if applicable; this MHTGR-DC does not structures, systems, and components important to safety to an address the residual heat removal system required under MHTGR-
ultimate heat sink shall be provided, as necessary, to transfer DC 34.
the combined heat load of these structures, systems, and components under normal operating and accident conditions. The staff inserted passive based on the system design for residual heat removal. If a specific MHTGR design can demonstrate that the Suitable redundancy in components and features and suitable reactor cavity cooling system (RCCS) provides indefinite core interconnections leak detection, and isolation capabilities shall cooling capability, then structural and equipment cooling systems would not be needed.
Appendix C to RG 1.232, Rev. 0, Page C-25
APPENDIX C. MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGN CRITERIA
IV. Heat Transport Systems Criterion MHTGR-DC Title and Content NRC Rationale for Adaptions to GDC
be provided to ensure that the system safety function can be accomplished, assuming a single failure. The GDC reference to electric power was removed. Refer to the rationale for ARDC 17 on electric power systems.
45 Inspection of structural and equipment cooling systems. This renamed MHTGR-DC accounts for advanced reactor system Same as ARDC design differences to include possible cooling requirements for The structural and equipment cooling systems shall be designed SSCs important to safety.
to permit appropriate periodic inspection of important components, such as heat exchangers and piping, to assure the integrity and capability of the systems.
46 Testing of structural and equipment cooling systems. This renamed MHTGR-DC accounts for advanced reactor system Same as ARDC design differences to include possible cooling requirements for The structural and equipment cooling systems shall be designed SSCs important to safety. Specific mention of pressure testing to permit appropriate periodic functional testing to assure (1) has been removed yet remains a potential requirement should it be the structural and leaktight integrity of their components, (2) necessary as a component of appropriate periodic functional the operability and the performance of the system components, testing... of cooling systems. A non-leaktight system may be and (3) the operability of the systems as a whole and, under acceptable for some designs provided that (1) the system leakage conditions as close to design as practical, the performance of does not impact safety functions under all conditions, and (2)
the full operational sequences that bring the systems into defense in depth is not impacted by system leakage.
operation for reactor shutdown and postulated accidents, including operation of associated systems. Functional testing is testing that assesses component and system operational readiness such as required in the ASME
OM Code as incorporated by reference in 10 CFR 50.55a and in Plant Technical Specifications.
Active has been deleted in item (2) because appropriate operability and performance tests of system components are required regardless of their active or passive natur
e. 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.
Appendix C to RG 1.232, Rev. 0, Page C-26
APPENDIX C. MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGN CRITERIA
V. Reactor Containment Criterion MHTGR-DC Title and Content NRC Rationale for Adaptions to GDC 50 Containment design basis. This criterion is not applicable to the MHTGR. The MHTGR
Not applicable to MHTGR. designs do not have a pressure retaining reactor containment structure but instead rely on a multibarrier functional containment configuration to control the release of radionuclides. See the MHTGR-DC 16 rationale.
51 Fracture prevention of containment pressure boundary. This criterion is not applicable to the MHTGR. The MHTGR
Not applicable to MHTGR. designs do not have a pressure retaining reactor containment structure but instead rely on a multibarrier functional containment configuration to control the release of radionuclides. See the MHTGR-DC 16 rationale.
52 Capability for containment leakage rate testing. This criterion is not applicable to the MHTGR. The MHTGR
Not applicable to MHTGR. designs do not have a pressure retaining reactor containment structure but instead rely on a multibarrier functional containment configuration to control the release of radionuclides. See the MHTGR-DC 16 rationale.
53 Provisions for containment testing and inspection. This criterion is not applicable to the MHTGR. The MHTGR
Not applicable to MHTGR. designs do not have a pressure retaining reactor containment structure but instead rely on a multibarrier functional containment configuration to control the release of radionuclides. See the MHTGR-DC 16 rationale.
54 Piping systems penetrating containment. This criterion is not applicable to the MHTGR. The MHTGR
Not applicable to MHTGR. designs do not have a pressure retaining reactor containment structure but instead rely on a multibarrier functional containment configuration to control the release of radionuclides. See the MHTGR-DC 16 rationale.
55 Reactor coolant boundary penetrating containment. This criterion is not applicable to the MHTGR. The MHTGR
Not applicable to MHTGR. designs do not have a pressure retaining reactor containment structure but instead rely on a multibarrier functional containment Appendix C to RG 1.232, Rev. 0, Page C-27
APPENDIX C. MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGN CRITERIA
V. Reactor Containment Criterion MHTGR-DC Title and Content NRC Rationale for Adaptions to GDC
configuration to control the release of radionuclides. See the MHTGR-DC 16 rationale.
56 Primary Containment isolation. This criterion is not applicable to the MHTGR. The MHTGR
Not applicable to MHTGR. designs do not have a pressure retaining reactor containment structure but instead rely on a multibarrier functional containment configuration to control the release of radionuclides. See the MHTGR-DC 16 rationale.
57 Closed system isolation valves. This criterion is not applicable to the MHTGR. The MHTGR
Not applicable to MHTGR. designs do not have a pressure retaining reactor containment structure but instead rely on a multibarrier functional containment configuration to control the release of radionuclides. See the MHTGR-DC 16 rationale.
VI. Fuel and Reactivity Control Criterion MHTGR-DC Title and Content NRC Rationale for Adaptions to GDC 60 Control of releases of radioactive materials to the environment.
Same as GDC
The nuclear power unit design shall include means to control suitably the release of radioactive materials in gaseous and liquid effluents and to handle radioactive solid wastes produced during normal reactor operation, including anticipated operational occurrences. Sufficient holdup capacity shall be provided for retention of gaseous and liquid effluents containing radioactive materials, particularly where unfavorable site environmental conditions can be expected to impose unusual operational limitations upon the release of such effluents to the environment.
61 Fuel storage and handling and radioactivity control. The underlying concept of establishing functional requirements for Same as ARDC radioactivity control in fuel storage and fuel handling systems is independent of the design of non-LWR advanced reactors.
Appendix C to RG 1.232, Rev. 0, Page C-28
APPENDIX C. MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGN CRITERIA
VI. Fuel and Reactivity Control Criterion MHTGR-DC Title and Content NRC Rationale for Adaptions to GDC
The fuel storage and handling, radioactive waste, and other However, some advanced designs may use dry fuel storage that systems which may contain radioactivity shall be designed to incorporates cooling jackets that can be liquid-cooled or air-cooled assure adequate safety under normal and postulated accident to remove heat. This modification to this GDC allows for both conditions. These systems shall be designed (1) with a capability liquid and air-cooling of the dry fuel storage containers.
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.
62 Prevention of criticality in fuel storage and handling.
Same as GDC
Criticality in the fuel storage and handling system shall be prevented by physical systems or processes, preferably by use of geometrically safe configurations.
63 Monitoring fuel and waste storage.
Same as GDC
Appropriate systems shall be provided in fuel storage and radioactive waste systems and associated handling areas (1) to detect conditions that may result in loss of residual heat removal capability and excessive radiation levels and (2) to initiate appropriate safety actions.
64 Monitoring radioactivity releases. The underlying concept of monitoring radioactivity releases from Means shall be provided for monitoring the reactor building the MHTGR particle fuel to the reactor building, effluent discharge atmosphere, effluent discharge paths, and plant environs for paths, and plant environs applies. High radioactivity in the reactor radioactivity that may be released from normal operations, building provides input to the plant protection system. In addition, including anticipated operational occurrences, and from the reactor building atmosphere is monitored for personnel postulated accidents.
Appendix C to RG 1.232, Rev. 0, Page C-29
APPENDIX C. MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGN CRITERIA
VI. Fuel and Reactivity Control Criterion MHTGR-DC Title and Content NRC Rationale for Adaptions to GDC
protection. Recirculation of loss-of-coolant fluids (i.e., water) does not apply to the MHTGR.
The descriptions of the associated atmospheres and spaces that are required to be monitored are revised to reflect the MHTGRs different design configuration and functional containment arrangement.
VII. Additional MHTGR-DC
Criterion MHTGR-DC Title and Content NRC Rationale for Adaptions to GDC 70 Reactor vessel and reactor system structural design basis. New MHTGR design-specific GDC are necessary to ensure that the The design of the reactor vessel and reactor system shall be such reactor vessel and reactor system (including the fuel, reflector, that their integrity is maintained during postulated accidents control rods, core barrel, and structural supports) integrity is
(1) to ensure the geometry for passive removal of residual heat preserved for passive heat removal and for the insertion of neutron from the reactor core to the ultimate heat sink and (2) to permit absorbers.
sufficient insertion of the neutron absorbers to provide for reactor shutdown.
71 Reactor building design basis. The reactor building functions are to protect and maintain passive The design of the reactor building shall be such that, during cooling geometry and to provide a pathway for the release of postulated accidents, it structurally protects the geometry for helium from the building in the case of a line break in the reactor passive removal of residual heat from the reactor core to the helium pressure boundary. This newly established criterion ensures ultimate heat sink and provides a pathway for the release of that these safety functions are provided.
reactor helium from the building in the event of depressurization It is noted that the reactor building is not relied upon to meet the accidents. offsite dose requirements of 10 CFR 50.34 (10 CFR 52.79).
72 Provisions for periodic reactor building inspection. This newly established criterion on periodic inspection and The reactor building shall be designed to permit (1) appropriate surveillance provides assurance that the reactor building will periodic inspection of all important structural areas and the perform its safety functions of protecting and maintaining the depressurization pathway, and (2) an appropriate surveillance configuration needed for passive cooling and providing a discharge program. pathway for helium depressurization events.
Appendix C to RG 1.232, Rev. 0, Page C-30
APPENDIX C. MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGN CRITERIA
Appendix C to RG 1.232, Rev. 0, Page C-31