NRC Rule: NuScale Small Modular Reactor Design Certification
Targeted News Service
WASHINGTON, Jan. 19 -- The U.S. Nuclear Regulatory Commission has issued a rule (10 CFR Part 52), published in the Federal Register on Jan. 19, 2022, entitled "NuScale Small Modular Reactor Design Certification." The rule was issued by Secretary Brooke P. Clark. DATES: This final rule is effective on February 21, 2023. The incorporation by reference of certain publications listed in the rule is approved by the Director of the Federal Register as of February 21, 2023. FOR FURTHER INFORMATION CONTACT: Yanely Malave, Office of Nuclear Material Safety and Safeguards, telephone: 301-415-1519, email: Yanely.Malave@nrc.gov, and Carolyn Lauron, Office of Nuclear Reactor Regulation, telephone: 301-415-2736, email: Carolyn.Lauron@nrc.gov. Both are staff of the U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001. * * * The U.S. Nuclear Regulatory Commission (NRC) is amending its regulations to certify the NuScale standard design for a small modular reactor. Applicants or licensees intending to construct and operate a NuScale standard design may do so by referencing this design certification rule. The applicant for certification of the NuScale standard design is NuScale Power, LLC. SUPPLEMENTARY INFORMATION: Table of Contents I. Background II. Opportunities for Public Participation III. Regulatory and Policy Issues IV. Technical Issues Associated With the NuScale Design V. Discussion A. Introduction (Section I) B. Definitions (Section II) C. Scope and Contents (Section III) D. Additional Requirements and Restrictions (Section IV) E. Applicable Regulations (Section V) F. Issue Resolution (Section VI) G. Duration of This Appendix (Section VII) H. Processes for Changes and Departures (Section VIII) I. [Reserved] (Section IX) J. Records and Reporting (Section X) VI. Public Comment Analysis VII. Section-by-Section Analysis VIII. Regulatory Flexibility Certification IX. Regulatory Analysis X. Backfitting and Issue Finality XI. Plain Writing XII. Environmental Assessment and Finding of No Significant Impact XIII. Paperwork Reduction Act XIV. Congressional Review Act XV. Agreement State Compatibility XVI. Voluntary Consensus Standards XVII. Availability of Documents XVIII. Incorporation by Reference--Reasonable Availability to Interested Parties I. Background Part 52 of title 10 of the Code of Federal Regulations (10 CFR), "Licenses, Certifications, and Approvals for Nuclear Power Plants," subpart B, "Standard Design Certifications," presents the process for obtaining standard design certifications. By letter dated December 31, 2016, NuScale Power, LLC, (NuScale Power) filed its application for certification of the NuScale standard design (hereafter referred to as NuScale). The NRC published a notification of receipt of the design certification application (DCA) in the Federal Register on February 22, 2017 (82 FR 11372). On March 30, 2017, the NRC published a notification of acceptance for docketing of the application in the Federal Register (82 FR 15717) and assigned docket number 52-048. The preapplication information submitted before the NRC formally accepted the application can be found in ADAMS under Docket No. PROJ0769. NuScale is the first small modular reactor design reviewed by the NRC. NuScale is based on a small light water reactor developed at Oregon State University in the early 2000s. It consists of one or more NuScale power modules (hereafter referred to as power module(s)). A power module is a natural circulation light water reactor composed of a reactor core, a pressurizer, and two helical coil steam generators located in a common reactor pressure vessel that is housed in a compact cylindrical steel containment. The NuScale reactor building is designed to hold up to 12 power modules. Each power module has a rated thermal output of 160 megawatt thermal (MWt) and electrical output of 50 megawatt electric (MWe), yielding a total capacity of 600 MWe for 12 power modules. All the NuScale power modules are partially submerged in a common safety-related pool, which is also the ultimate heat sink for up to 12 power modules. The pool portion of the reactor building is located below grade. The design utilizes several first-of-a-kind approaches for accomplishing key safety functions, resulting in no need for Class 1E safety-related power (no emergency diesel generators), no need for pumps to inject water into the core for post-accident coolant injection, and reduced need for control room staffing while providing safe operation of the plant during normal and post-accident operation. II. Opportunities for Public Participation The proposed rule and environmental assessment were published in the Federal Register on July 1, 2021, for a 60-day public comment period (86 FR 34999). The public comment period was scheduled to close on August 30, 2021. The NRC subsequently extended the comment period by 45 days (86 FR 47251; August 24, 2021), providing a total comment period of 105 days. The public comment period closed on October 14, 2021. The public comments informed the development of this final rule. III. Regulatory and Policy Issues A. Exemptions for Future Applicants Referencing NuScale 1. Control Room Staffing Requirements The requirements in Sections 50.54(k) and 50.54(m) identify the minimum number of licensed operators that must be on site, in the control room, and at the controls. The requirements are conditions in every nuclear power reactor operating license issued under 10 CFR part 50, "Domestic Licensing of Production and Utilization Facilities." The requirements also are conditions in every combined license (COL) issued under 10 CFR part 52; however, they are applicable only after the Commission makes the finding under Section 52.103(g) that the acceptance criteria in the COL are met. In a letter to the NRC, dated September 15, 2015, NuScale Power proposed that 6 licensed operators would operate up to 12 power modules from a single control room. The staffing proposal would meet the requirements of Section 50.54(k) but would not meet the requirements in Section 50.54(m)(2)(i) because the minimum requirements for the onsite staffing table in Section 50.54(m)(2)(i) do not address operation of more than two units from a single control room. The proposal also would not meet Section 50.54(m)(2)(iii), which requires a licensed operator at the controls for each fueled unit. Absent alternative staffing requirements, future applicants referencing the NuScale design would need to request an exemption. In DCA, Part 7, Section 6, NuScale requested that the NRC approve design-specific control room staffing requirements in lieu of the requirements in Section 50.54(m). In the DCA Part 7, Section 6.2, "Justification for Rulemaking," NuScale Power provided a technical basis for its proposed alternative control room staffing requirements. NuScale Power's proposed approach is consistent with SECY-11-0098, "Operator Staffing for Small or Multi-Module Nuclear Power Plant Facilities," dated July 22, 2011. For the reasons described in Chapter 18, Section 18.5.4.2, "Evaluation of the Applicant's Technical Basis," of the final safety evaluation report, the NRC found that NuScale Power's proposed staffing level, as described in the DCA Part 7, Section 6, is acceptable. Because Section V, "Applicable Regulations," of this final rule includes the alternative staffing requirement provisions, staffing table, and appropriate table notes, a future applicant or licensee that references appendix G to 10 CFR part 52 will not need to request an exemption from Section 50.54(m). 2. Preoperational and Periodic Testing of Primary Reactor Containment General Design Criterion (GDC) 52, "Capability for Containment Leakage Rate Testing," requires that the containment be designed so that periodic, integrated leakage rate testing can be conducted at containment design pressure; the underlying purpose of which is to provide design capability for testing that assures that containment leakage integrity is maintained and containment vessel leakage does not exceed allowable leakage rate values (see appendix J to 10 CFR part 50). Under 10 CFR 50.54(o), operating licenses and combined licenses for certain water-cooled power reactors must include a condition that the primary containment shall be subject to appendix J to 10 CFR part 50, "Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors." Appendix J to 10 CFR part 50 requires that primary reactor containments meet the containment leakage test requirements to provide for preoperational and periodic verification by tests of the leak-tight integrity of the primary reactor containment (Type A) and systems and components that penetrate containment (Type B and Type C). NuScale Power requested an exemption from GDC 52 in order to not design NuScale to include the capability for Type A testing and requested that the design certification rule exempt licensees referencing the NuScale design certification rule from the requirement for Type A testing in appendix J to 10 CFR part 50. NuScale Power's request was based on the NuScale small modular reactor design meeting the underlying purpose of the regulation through means not anticipated when the NRC issued GDC 52 and appendix J to 10 CFR part 50. NuScale Power stated that the NuScale containment has two primary features distinguishing it from containments at existing light water reactors that provide assurance that no unknown leakage pathways will be present. First, the NuScale containment is designed and would be constructed as a pressure vessel, and therefore leakage due to vessel design or fabrication flaws would be identified during a required preservice structural integrity test. In contrast to a Type A test, this test is a hydrostatic leakage test at design pressure, with no visible leakage as its acceptance criterion. Second, the containment is 100-percent inspectable, both inside and outside, whereby aging-related flaws leading to potential leakage could be observed. Containment leakage integrity assurance for NuScale is described in detail in technical report TR-1116-51962-NP, "NuScale Containment Leakage Integrity Assurance," Rev. 1 (May 2019), which this final rule incorporates by reference. NuScale Power stated that the required preservice tests and inservice inspections described in TR-1116-51962-NP, including Type B and Type C testing without Type A testing, ensure that containment leakage rates remain acceptable. In Chapter 6, Section 6.2.6.4, " Technical Evaluation for Exemption Request No. 7," of the final safety evaluation report, the NRC staff concluded that granting this exemption from Type A testing, and associated design features required by GDC 52 to provide for Type A testing, is acceptable because the NuScale design relies on the preservice pressure test, successful Type B and C testing at each refueling as required in appendix J to 10 CFR part 50, periodic inservice inspections, and direct observation of the entire vessel to identify potential degradation or unknown leakage pathways for the remainder of the service life for the containment. The NRC received a comment that the exemption from the requirement for Type A testing in appendix J to 10 CFR part 50 should have been listed in the proposed rule. The NRC agrees that the exemption should have been included in the proposed rule. The NRC's conclusion that Type A testing is not necessary for NuScale was noticed for comment as the basis for the exemption from GDC 52. The exemption from Type A testing itself was discussed in detail in the same section of final safety evaluation report that evaluated the exemption from GDC 52. Although the exemption from Type A testing was not included in the proposed rule, the change to this final rule only specifies that future licensees that reference this final rule will not be required to perform Type A testing for which NuScale is not designed or required to be capable of. Therefore, the NRC concludes that the exemption from the Type A test in appendix J to 10 CFR part 50 is a logical outgrowth of the proposed rule. In addition, because the issue of whether Type A testing is necessary for NuScale was noticed in the proposed rule and the NRC received no comments on the matter, the NRC finds that notice and comment on this exemption from Type A testing is unnecessary within the meaning of 5 U.S.C. 553(b). Thus, Section V, "Applicable Regulations," in this final rule includes an exemption for licensees referencing appendix G to 10 CFR part 52 from the requirement of appendix J to 10 CFR part 50 to conduct Type A testing. B. Incorporation by Reference Section III.A, "Incorporation by reference approval," of appendix G to 10 CFR part 52 lists documents that were approved by the Director of the Office of the Federal Register for incorporation by reference into this appendix. Section III.B.2 identifies information that is not within the scope of the design certification and, therefore, is not incorporated by reference into this appendix. This information includes conceptual design information, as defined in Section 52.47(a)(24), and the discussion of "first principles" described in the Design Control Document (DCD) Part 2, Tier 2, Section 14.3.2, "Tier 1 Design Description and Inspections, Tests, Analyses, and Acceptance Criteria First Principles." The final rule has been updated to align with the Office of the Federal Register's latest guidance for incorporation by reference, issued on March 1, 2022, as supplemented by Release 1-2022 to the Incorporation by Reference Handbook. C. Issues Not Resolved by the Design Certification The NRC identified three issues as not resolved within the meaning of Section 52.63(a)(5). There was insufficient information available for the NRC to resolve issues regarding (1) the shielding wall design in certain areas of the plant, (2) the potential for containment leakage from the combustible gas monitoring system, and (3) the ability of the steam generator tubes to maintain structural and leakage integrity during density wave oscillations in the secondary fluid system, including the method of analysis to predict the thermal-hydraulic conditions of the steam generator secondary fluid system and resulting loads, stresses, and deformations from density wave oscillations from reverse flow. 1. Shielding Wall Design As discussed in Section 12.3.4.1.2 of the final safety evaluation report, the NRC found that there were insufficient design details available regarding shielding wall design with the presence of large penetrations, such as the main steam lines; main feedwater lines; and power module bay heating, ventilation, and air conditioning lines in the radiation shield wall between the power module bay and the reactor building steam gallery area. Without this shielding design information, the NRC is unable to confirm that the radiological doses to workers will be maintained within the radiation zone limits specified in the application. This issue is narrowly focused on the shielding walls between the reactor module bays and the reactor building steam gallery areas. The radiation zones and dose calculations, including dose calculations for the dose to workers, members of the public, and environmental qualification, in areas outside of the reactor module bay are calculated assuming a solid wall and currently do not account for penetrations in the shield wall. An applicant is required to demonstrate penetration shielding adequate to address the following issues in the NuScale DCD: the plant radiation zones, environmental qualification dose calculations, and dose estimates for workers and the public. An applicant can provide this information for the NRC to review because this issue involves a localized area of the plant without affecting other aspects of the NRC's review of the NuScale design. Therefore, the NRC has determined that this information can be provided by an applicant that references this appendix without a demonstrable impact on safety or standardization. Appendix G to 10 CFR part 52, Section VI, "Issue Resolution," clarifies that this issue is not resolved within the meaning of Section 52.63(a)(5), and Section IV, "Additional Requirements and Restrictions," states that the COL applicant is responsible for providing the design information to address this issue. 2. Containment Leakage From the Combustible Gas Monitoring System As documented in Section 12.3.4.1.3 of the final safety evaluation report, there was insufficient information available regarding the NuScale combustible gas monitoring system and the potential for leakage from this system outside containment. Without additional information regarding the potential for leakage from this system, the NRC was unable to determine whether this leakage could impact analyses performed to assess main control room dose consequences, offsite dose consequences to members of the public, and whether this system can be safely re-isolated after monitoring is initiated due to potentially high dose levels at or near the isolation valve location. The isolation valve can only be operated locally, and dose levels at the valve location have not been determined. This issue is narrowly focused on the radiation dose implications as a result of using the post-accident combustible gas monitoring loop. An applicant is required under Sections 50.34(f)(2) and 52.47(a)(2) to demonstrate either that offsite and main control room dose calculations are not exceeded or that the system can be safely re-isolated, if needed. This issue does not affect normal plant operation or non-core damage accidents. The issue may be resolved by performing radiation dose calculations and demonstrating that doses would remain within applicable dose limits in 10 CFR part 20, "Standards for Protection Against Radiation." More information may be available at the application stage that would allow for more detailed calculations. Any design changes to address this issue would only affect the combustible gas monitoring loop to ensure it can be re-isolated or to ensure that dose limits are not exceeded. Such design changes likely would not have an impact on other systems or equipment, and the NRC would review such changes and any resulting effects on other structures, systems, and components during the application review to determine whether there is reasonable assurance of adequate protection of public health and safety. Therefore, the NRC has determined that this information can be provided by an applicant that references this appendix without a demonstrable impact on safety or standardization. Appendix G to 10 CFR part 52, Section VI, "Issue Resolution," clarifies that this issue is not resolved within the meaning of Section 52.63(a)(5), and Section IV, "Additional Requirements and Restrictions," states that the COL applicant is responsible for providing the design information to address this issue. 3. Steam Generator Stability During Density Wave Oscillations and Associated Method of Analysis Section 5.4.1.2, "System Design," in Revision 2 of the DCA Part 2, Tier 2 (ADAMS Accession No. ML18310A345), stated that a flow restriction device at the inlet to each steam generator tube "ensures secondary-side flow stability and precludes density wave oscillations." However, the applicant modified this section in Revision 3 of the DCA Part 2, Tier 2 (ADAMS Accession No. ML19241A431), to state that the steam generator inlet flow restrictors provide the necessary secondary-side pressure drop "to reduce flow oscillations to acceptable limits." Revision 4.1 of the DCA (ADAMS Accession No. ML20205L562) revised Section 5.4.1.2 to state that the steam generator inlet flow restrictors are designed "to reduce the potential for density wave oscillations." Revision 5 of this section of the DCA (ADAMS Accession No. ML20225A071) provides only editorial changes to Revision 4.1 and does not change the technical content or conclusions. Sections 3.9.2, 3.9.5, and 5.4.1 of the final safety evaluation report relied on the applicant's statements in Revision 2 and Revision 3 of the DCA that flow oscillations in the secondary fluid system of the steam generators would either be precluded or minimal. After issuance of the advanced safety evaluation report, the NRC noted inconsistencies and gaps in the information provided in Sections 3.9.1, 3.9.2, and 5.4.1 of Revision 4.1 of the DCA Part 2, Tier 2, regarding the potential for significant density wave oscillations in the steam generator tubes, including both forward and reverse secondary flow. The testing performed by the applicant on various conceptual designs of the steam generator inlet flow restrictors only involved flow in the forward direction without oscillation or reverse flow. As a result, NuScale Power has not demonstrated that the flow oscillations that are predicted to occur on the secondary side of the steam generators will not cause failure of the inlet flow restrictors. Structural and leakage integrity of the inlet flow restrictors in the steam generators is necessary to avoid damage to multiple steam generator tubes, caused directly by broken parts or indirectly by unexpected density wave oscillation loads. Damage to multiple steam generator tubes could disrupt natural circulation in the reactor coolant pathway and interfere with the decay heat removal system and the emergency core cooling system, which is relied upon to cool the reactor core in a NuScale power module. The failure of multiple steam generator tubes resulting from failure of an inlet flow restrictor has not been included within the scope of the NuScale accident analyses in DCA Part 2, Tier 2, Chapter 15. Therefore, the NRC concludes that NuScale Power has not demonstrated compliance with 10 CFR 52.47(a)(2)(iv) and appendix A to 10 CFR part 50, GDC 4 and GDC 31, relative to potential impacts on steam generator tube integrity from inlet flow restrictor failure. As described previously, NuScale Power made a change to the description of inlet flow restrictor performance beginning with DCA Part 2, Tier 2, Revision 3, that indicates that the design no longer precludes density wave oscillations in the secondary side of the steam generators. As a result, the design needs a method of analysis to predict the thermal-hydraulic conditions of the steam generator secondary fluid system and resulting loads, stresses, and deformations from density wave oscillations including reverse flow. However, as described in the next paragraph, NuScale power did not provide verification and validation for its proposed method of analysis to demonstrate it is appropriate for this purpose. The DCA Part 2, Tier 2, Section 3.9.1.2, "Computer Programs Used in Analyses," lists the computer programs used by NuScale Power in the dynamic and static analyses of mechanical loads, stresses, and deformations, and in the hydraulic transient load analyses of seismic Category I components and supports for the NuScale nuclear power plant. Section 3.9.1.2 states that NRELAP5 is NuScale's proprietary system thermal-hydraulics code for use in safety-related design and analysis calculations and is pre-verified and configuration-managed. The advanced safety evaluation report, Section 3.9.1.4.9, "Computer Programs Used in Analyses," states that the NRELAP5 computer program had received verification and validation. Following preparation of the advanced safety evaluation report, the NRC noted a discrepancy between two statements in the DCA about validation for NRELAP5: DCA Part 2, Tier 2, Section 5.4.1.3, in Revision 4 stated that NRELAP5 was validated for determining density wave oscillation thermal-hydraulic conditions, referring to Section 15.0.2 for more information, but neither Section 15.0.2 nor technical report TR-1016-51669-NP describe validation for determining density wave oscillation thermal-hydraulic conditions. On June 19, 2020, NuScale submitted Revision 4.1 of the DCA Part 2, Tier 2 (ADAMS Accession No. ML20205L562; subsequently included in Revision 5 of the DCA submitted on July 29, 2020 (ADAMS Accession No. ML20225A071)), to correct the discrepancies and acknowledge the need for a COL applicant to address secondary-side instabilities in the steam generator design. Specifically, the update to Section 3.9.1.2 in Revision 4.1 of DCA Part 2, Tier 2, references DCA Part 2, Tier 2, Section 15.0.2, "Review of Transient and Accident Analysis Methods," for the discussion of the development, use, verification, validation, and code limitations of the NRELAP5 computer program for application to transient and accident analyses. The correction to Section 3.9.1.2 also references technical report TR-1016-51669-NP, "NuScale Power Module Short-Term Transient Analysis," incorporated by reference in DCA Part 2, Tier 2, Table 1.6-2, for application of the NRELAP5 computer program to short-term transient dynamic mechanical loads, such as pipe breaks and valve actuations. In addition, the correction to Section 3.9.1.2 includes a new COL item specifying that a COL applicant that references the NuScale DCD will develop an evaluation methodology for the analysis of secondary-side instabilities in the steam generator design. The COL item states that this methodology would address the identification of potential density wave oscillations in the steam generator tubes and qualification of the applicable portions of the reactor coolant system integral reactor pressure vessel and steam generator given the occurrence of density wave oscillations, including the effects of reverse fluid flows within the tubes. These corrections to the DCA clarify that the evaluation methodology for the analysis of secondary-side instabilities in the steam generator design was not verified and validated as part of the NuScale DCA but will need to be established by the COL applicant. This steam generator design issue is narrowly focused on the effects of density wave oscillations in the secondary fluid system on steam generator tubes to maintain structural and leakage integrity, including the method of analysis to predict the thermal-hydraulic conditions of the steam generator secondary fluid system and resulting loads, stresses, and deformations from density wave oscillations including reverse flow. No other reactor safety aspect of the steam generators is impacted by this design issue. As a result, the NRC finds that this is an isolated issue that does not affect other aspects of the NRC's review of the design of the NuScale nuclear power plant. Therefore, the NRC has determined that this information can be provided by an applicant that references this appendix, consistent with the other design information regarding steam generator integrity described in DCA Part 2, Tier 2, Sections 3.9.1, 3.9.2, and 5.4.1, without a demonstrable impact on safety or standardization. Therefore, appendix G to 10 CFR part 52, Section VI, "Issue Resolution," clarifies that this issue is not resolved within the meaning of Section 52.63(a)(5), and Section IV, "Additional Requirements and Restrictions," states that the COL applicant is responsible for providing the design information to address this issue. D. The Term "Multi-Unit" as Applied to NuScale In a letter response to NuScale Power dated October 25, 2016, the NRC staff explained how the staff's review of NuScale would apply the definitions for "nuclear power unit" from Appendix A to 10 CFR part 50, "General Design Criteria for Nuclear Power Plants," and "modular design" from Section 52.1, "Definitions." As defined in Appendix A to 10 CFR part 50, a nuclear power unit is the combination of a nuclear reactor and the equipment for power generation. As defined in Section 52.1, modular design means that the nuclear power station consists of two or more essentially identical nuclear reactors (modules) and that each module is capable of operation independent of the other modules, even if they have some shared systems. The NuScale modular design combines one or more nuclear reactors (up to 12) with the necessary equipment for power generation, such that each separate nuclear reactor can be operated independent of the stage of completion or operating condition of any other nuclear reactor on the same site. Therefore, each reactor ( i.e., power module) is a separate nuclear power unit. However, NuScale's modular design means that some multi-unit considerations are integral to the design. The NuScale DCD addresses multi-unit considerations other than construction for up to 12 power modules in a single reactor building, but the NuScale DCD does not address multi-unit issues that may arise if a NuScale facility is constructed and operated on the same site as another nuclear facility. For previously certified or licensed power reactor designs (one nuclear power unit per reactor building), multi-unit site considerations arose when multiple nuclear power units (in separate reactor buildings) on the same site could affect the construction or operation of another unit in a manner not previously reviewed by the NRC. However, because the NuScale design has been reviewed and is certified for multiple units in a single reactor building, issues related to multiple NuScale units in the same reactor building constructed at the same time have been resolved. Future applicants referencing the NuScale design certification will need to address multi-unit construction issues and, if applicable, multi-unit issues for a proposed NuScale facility to be constructed and operated on the same site as another nuclear facility, including adding additional NuScale modules to a previously licensed NuScale reactor building. The NRC has added a definition of the term "nuclear power unit" to this final rule. Dated: January 11, 2023. For the Nuclear Regulatory Commission. Brooke P. Clark, Secretary of the Commission. [FR Doc. 2023-00729 Filed 1-18-23; 8:45 am] BILLING CODE 7590-01-P The document was published in the Federal Register: https://www.federalregister.gov/documents/2023/01/19/2023-00729/nuscale-small-modular-reactor-design-certification TARGETED NEWS SERVICE (founded 2004) features non-partisan 'edited journalism' news briefs and information for news organizations, public policy groups and individuals; as well as 'gathered' public policy information, including news releases, reports, speeches. For more information contact MYRON STRUCK, editor, editor@targetednews.com, Springfield, Virginia; 703/304-1897; https://targetednews.com