10 C.F.R. § 52 app G to Part 52
I. Introduction
Appendix G constitutes the standard design certification for the NuScale design (hereinafter referred to as NuScale), in accordance with 10 CFR part 52, subpart B. The applicant for this standard design certification NuScale is NuScale Power, LLC.
II. Definitions
A. Generic design control document (generic DCD) means the documents containing the Tier 1 and Tier 2 information (including the technical and topical reports referenced in Chapter 1) and generic technical specifications that are incorporated by reference into this appendix.
B. Generic technical specifications (generic TS) means the information required by 10 CFR 50.36 and 50.36a for the portion of the plant that is within the scope of this appendix.
C. Plant-specific DCD means that portion of the combined license (COL) final safety analysis report (FSAR) that sets forth both the generic DCD information and any plant-specific changes to generic DCD information.
D. Tier 1 means the portion of the design-related information contained in the generic DCD that is approved and certified by this appendix (Tier 1 information). The design descriptions, interface requirements, and site parameters are derived from Tier 2 information. Tier 1 information includes:
1. Definitions and general provisions;
2. Design descriptions;
3. Inspections, tests, analyses, and acceptance criteria (ITAAC);
4. Significant site parameters; and
5. Significant interface requirements.
E. Tier 2 means the portion of the design-related information contained in the generic DCD that is approved but not certified by this appendix (Tier 2 information). Compliance with Tier 2 is required, but generic changes to and plant-specific departures from Tier 2 are governed by Section VIII of this appendix. Compliance with Tier 2 provides a sufficient, but not the only acceptable, method for complying with Tier 1. Compliance methods differing from Tier 2 must satisfy the change process in Section VIII of this appendix. Regardless of these differences, an applicant or licensee must meet the requirement in paragraph III.B of this appendix to reference Tier 2 when referencing Tier 1. Tier 2 information includes:
1. Information required by § 52.47(a) and (c) , with the exception of generic TS and conceptual design information;
2. Supporting information on the inspections, tests, and analyses that will be performed to demonstrate that the acceptance criteria in the ITAAC have been met; and
3. COL action items (COL license information) identify certain matters that must be addressed in the site-specific portion of the FSAR by an applicant who references this appendix. These items constitute information requirements but are not the only acceptable set of information in the FSAR. An applicant may depart from or omit these items, provided that the departure or omission is identified and justified in the FSAR. After issuance of a construction permit or COL, these items are not requirements for the licensee unless such items are restated in the FSAR.
F. Departure from a method of evaluation described in the plant-specific DCD used in establishing the design bases or in the safety analyses means:
1. Changing any of the elements of the method described in the plant-specific DCD unless the results of the analysis are conservative or essentially the same; or
2. Changing from a method described in the plant-specific DCD to another method unless that method has been approved by the NRC for the intended application.
G. Nuclear power unit, as applied to this certified design, means a nuclear power module and associated equipment necessary for electric power generation and includes those structures, systems, and components required to provide reasonable assurance the facility can be operated without undue risk to the health and safety of the public.
H. All other terms in this appendix have the meaning set out in 10 CFR 50.2 , 10 CFR 52.1 , or Section 11 of the Atomic Energy Act of 1954, as amended, as applicable.
III. Scope and Contents
A. Incorporation by reference.
1. Certain material listed in paragraph III.A.2 of this appendix is incorporated by reference into this appendix G with the approval of the Director of the Federal Register in accordance with 5 U.S.C. 552(a) and 1 CFR part 51 . All approved incorporation by reference (IBR) material in paragraph III.A.2 of this appendix may be obtained from NuScale Power, LLC, 6650 SW Redwood Lane, Suite 210, Portland, Oregon 97224, telephone: 1-971-371-1592, email: RegulatoryAffairs@nuscalepower.com, and can be inspected as follows:
a. Contact the U.S. Nuclear Regulatory Commission at: U.S. Nuclear Regulatory Commission, Two White Flint North, 11545 Rockville Pike, Rockville, Maryland 20852; telephone: 301-415-7000; email: Library.Resource@nrc.gov; https://www.nrc.gov/reading-rm/pdr.html.
b. Access ADAMS and view the material online in the NRC Library at https://www.nrc.gov/reading-rm/adams.html. In ADAMS, search under ADAMS Accession No. ML20225A071. The material is available in the ADAMS Public Documents collection.
c. If you do not have access to ADAMS or if you have problems accessing documents located in ADAMS, contact the NRC's Public Document Room (PDR) reference staff at 1-800-397-4209, 301-415-3747, or by email at PDR.Resource@nrc.gov.
d. For information on the availability of this material at the National Archives and Records Administration, visit www.archives.gov/federal-register/cfr/ibr-locations.html or email: fr.inspection@nara.gov.
2. Material incorporated by reference.
a. NuScale Standard Plant Design Certification Application, Certified Design Descriptions and Inspections, Tests, Analyses, & Acceptance Criteria (ITAAC), Part 2-Tier 1, Revision 5, July 2020.
b. NuScale Standard Plant Design Certification Application, Part 2-Tier 2, Revision 5, July 2020, including:
i. Chapter One, Introduction and General Description of the Plant.
ii. Chapter Two, Site Characteristics and Site Parameters.
iii. Chapter Three, Design of Structures, Systems, Components and Equipment.
iv. Chapter Four, Reactor.
v. Chapter Five, Reactor Coolant System and Connecting Systems.
vi. Chapter Six, Engineered Safety Features.
vii. Chapter Seven, Instrumentation and Controls.
viii. Chapter Eight, Electric Power.
ix. Chapter Nine, Auxiliary Systems.
x. Chapter Ten, Steam and Power Conversion System.
xi. Chapter Eleven, Radioactive Waste Management.
xii. Chapter Twelve, Radiation Protection.
xiii. Chapter Thirteen, Conduct of Operations.
xiv. Chapter Fourteen, Initial Test Program and Inspections, Tests, Analyses, and Acceptance Criteria.
xv. Chapter Fifteen, Transient and Accident Analyses.
xvi. Chapter Sixteen, Technical Specifications.
xvii. Chapter Seventeen, Quality Assurance and Reliability Assurance.
xviii. Chapter Eighteen, Human Factors Engineering.
xix. Chapter Nineteen, Probabilistic Risk Assessment and Severe Accident Evaluation.
xx. Chapter Twenty, Mitigation of Beyond-Design-Basis Events.
xxi. Chapter Twenty-One, Multi-Module Design Considerations.
c. DCA Part 4, Volume 1, Revision 5.0, Generic Technical Specifications, NuScale Nuclear Power Plants, Volume 1: Specifications.
d. DCA Part 4, Volume 2, Revision 5.0, Generic Technical Specifications, NuScale Nuclear Power Plants, Volume 2: Bases.
e. ES-0304-1381-NP, Human-System Interface Style Guide, December 2019, Revision 4.
f. RP-0215-10815-NP, Concept of Operations, May 2019, Revision 3.
g. RP-0316-17614-NP, Human Factors Engineering Operating Experience Review Results Summary Report, December 7, 2016, Revision 0.
h. RP-0316-17615-NP, Human Factors Engineering Functional Requirements Analysis and Function Allocation Results Summary Report, December 2, 2016, Revision 0.
i. RP-0316-17616-NP, Human Factors Engineering Task Analysis Results Summary Report, April 2019, Revision 2.
j. RP-0316-17617-NP, Human Factors Engineering Staffing and Qualifications Results Summary Report, December 2, 2016, Revision 0.
k. RP-0316-17618-NP, Human Factors Engineering Treatment of Important Human Actions Results Summary Report, December 2, 2016, Revision 0.
l. RP-0316-17619-NP, Human Factors Engineering Human-System Interface Design Results Summary Report, April 2019, Revision 2.
m. RP-0516-49116-NP, Control Room Staffing Plan Validation Results, December 2, 2016, Revision 1.
n. RP-0914-8534-NP, Human Factors Engineering Program Management Plan, April 2019, Revision 5.
o. RP-0914-8543-NP, Human Factors Verification and Validation Implementation Plan, April 2019, Revision 5.
p. RP-0914-8544-NP, Human Factors Engineering Design Implementation Plan, November 2019, Revision 4.
q. RP-1018-61289-NP, Human Factors Engineering Verification and Validation Results Summary Report, July 2019, Revision 1.
r. RP-1215-20253-NP, Control Room Staffing Plan Validation Methodology, December 2, 2016, Revision 3.
s. TR-0116-20781-NP, Fluence Calculation Methodology and Results, July 2019, Revision 1.
t. TR-0116-20825-NP-A, Applicability of AREVA Fuel Methodology for the NuScale Design, June 2016, Revision 1.
u. TR-0116-21012-NP-A, NuScale Power Critical Heat Flux Correlations, December 2018, Revision 1.
v. TR-0316-22048-NP, Nuclear Steam Supply System Advanced Sensor Technical Report, May 2020, Revision 3.
w. TR-0515-13952-NP-A, Risk Significance Determination, October 2016, Revision 0.
x. TR-0516-49084-NP, Containment Response Analysis Methodology Technical Report, May 2020, Revision 3.
y. TR-0516-49416-NP-A, Non-Loss-of-Coolant Accident Analysis Methodology, July 2020, Revision 3.
z. TR-0516-49417-NP-A, Evaluation Methodology for Stability Analysis of the NuScale Power Module, March 2020, Revision 1.
aa. TR-0516-49422-NP-A, Loss-of-Coolant Accident Evaluation Model, July 2020, Revision 2.
ab. TR-0616-48793-NP-A, Nuclear Analysis Codes and Methods Qualification, November 2018, Revision 1.
ac. TR-0616-49121-NP, NuScale Instrument Setpoint Methodology Technical Report, May 2020, Revision 3.
ad. TR-0716-50350-NP-A, Rod Ejection Accident Methodology, June 2020, Revision 1.
ae. TR-0716-50351-NP-A, NuScale Applicability of AREVA Method for the Evaluation of Fuel Assembly Structural Response to Externally Applied Forces, April 2020, Revision 1.
af. TR-0716-50424-NP, Combustible Gas Control, March 2019, Revision 1.
ag. TR-0716-50439-NP, NuScale Comprehensive Vibration Assessment Program Analysis Technical Report, July 2019, Revision 2.
ah. TR-0815-16497-NP-A, Safety Classification of Passive Nuclear Power Plant Electrical Systems, January 2018, Revision 1.
ai. TR-0816-49833-NP, Fuel Storage Rack Analysis, November 2018, Revision 1.
aj. TR-0816-50796-NP, Loss of Large Areas Due to Explosions and Fires Assessment, June 2019, Revision 1.
ak. TR-0816-50797, Mitigation Strategies for Loss of All AC Power Event [NuScale Nonproprietary], October 2019, Revision 3.
al. TR-0816-51127-NP, NuFuel-HTP2TM Fuel and Control Rod Assembly Designs, December 2019, Revision 3.
am. TR-0818-61384-NP, Pipe Rupture Hazards Analysis, July 2019, Revision 2.
an. TR-0915-17564-NP-A, Subchannel Analysis Methodology, February 2019, Revision 2.
ao. TR-0915-17565-NP-A, Accident Source Term Methodology, February 2020, Revision 4.
ap. TR-0916-51299-NP, Long-Term Cooling Methodology, May 2020, Revision 3.
aq. TR-0916-51502-NP, NuScale Power Module Seismic Analysis, April 2019, Revision 2.
ar. TR-0917-56119-NP, CNV Ultimate Pressure Integrity, June 2019, Revision 1.
as. TR-0918-60894-NP, NuScale Comprehensive Vibration Assessment Program Measurement and Inspection Plan Technical Report, August 2019, Revision 1.
at. NP-TR-1010-859-NP-A, NuScale Topical Report: Quality Assurance Program Description for the NuScale Power Plant, May 2020, Revision 5.
au. TR-1015-18177-NP, Pressure and Temperature Limits Methodology, October 2018, Revision 2.
av. TR-1015-18653-NP-A, Design of the Highly Integrated Protection System Platform, May 2017, Revision 2.
aw. TR-1016-51669-NP, NuScale Power Module Short-Term Transient Analysis, July 2019, Revision 1.
ax. TR-1116-51962-NP, NuScale Containment Leakage Integrity Assurance, May 2019, Revision 1.
ay. TR-1116-52065-NP, Effluent Release (GALE Replacement) Methodology and Results, November 2018, Revision 1.
B.
1. An applicant or licensee referencing this appendix, in accordance with Section IV of this appendix, shall incorporate by reference and comply with the requirements of this appendix except as otherwise provided in this appendix.
2. Conceptual design information, as set forth in the design certification application Part 2, Tier 2, Section 1.2 , and the discussion of "first principles" contained in design certification application Part 2, Tier 2, Section 14.3.2 , are not incorporated by reference into this appendix.
C. If there is a conflict between Tier 1 and Tier 2 of the DCD, then Tier 1 controls.
D. If there is a conflict between the generic DCD and either the application for the design certification of NuScale or the final safety evaluation report related to certification of the NuScale standard design, then the generic DCD controls.
E. Design activities for structures, systems, and components that are wholly outside the scope of this appendix may be performed using site characteristics, provided the design activities do not affect the DCD or conflict with the interface requirements.
IV. Additional Requirements and Restrictions
A. An applicant for a COL that wishes to reference this appendix shall, in addition to complying with the requirements of §§ 52.77 , 52.79 , and 52.80 , comply with the following requirements:
1. Incorporate by reference, as part of its application, this appendix.
2. Include, as part of its application:
a. A plant-specific DCD containing the same type of information and using the same organization and numbering as the generic DCD for NuScale, either by including or incorporating by reference the generic DCD information, and as modified and supplemented by the applicant's exemptions and departures;
b. The reports on departures from and updates to the plant-specific DCD required by paragraph X.B of this appendix;
c. Plant-specific TS, consisting of the generic and site-specific TS that are required by 10 CFR 50.36 and 50.36a ;
d. Information demonstrating that the site characteristics fall within the site parameters and that the interface requirements have been met;
e. Information that addresses the COL action items;
f. Information required by § 52.47(a) that is not within the scope of this appendix;
g. Information demonstrating that necessary shielding to limit radiological dose consistent with the radiation zones specified in design certification application Part 2, Tier 2, Chapter 12, Figure 12.3-1, "Reactor Building Radiation Zone Map," is provided to account for penetrations in the radiation shield wall between the power module bay and the reactor building steam gallery area;
h. Information demonstrating that the requirements of 10 CFR 50.34(f)(2) (xxviii) are met with respect to potential radiological releases under accident conditions from the systems used for post-accident hydrogen and oxygen monitoring described in design certification application Part 2, Tier 2, Section 6.2.5 ; information demonstrating that post-accident leakage from these systems does not result in the total main control room dose exceeding the dose criteria for the surrogate event with significant core damage, which may include use of design features compliant with 10 CFR 50.34(f)(2)(vii) , as appropriate; and information demonstrating that post-accident leakage from these systems does not result in the total dose for the surrogate event with significant core damage exceeding the offsite dose criteria, as required by 10 CFR 52.47(a)(2)(iv) ; and
i. Information demonstrating that the requirements of 10 CFR 52.47(a)(2)(iv) and General Design Criterion (GDC) 4 and GDC 31 of appendix A to 10 CFR part 50 are met with respect to the structural and leakage integrity of the steam generator tubes that might be compromised by effects from 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 and reverse flow. This information must be consistent with the other design information regarding steam generator integrity contained in design certification application Part 2, Tier 2, Sections 3.9.2 and 5.4.1 .
3. Include, in the plant-specific DCD, the sensitive, unclassified, non-safeguards information (including proprietary information and security-related information) and safeguards information referenced in the NuScale generic DCD.
4. Include, as part of its application, a demonstration that an entity other than NuScale Power, LLC, is qualified to supply the NuScale generic DCD, unless NuScale Power, LLC, supplies the design for the applicant's use.
B. The Commission reserves the right to determine in what manner this appendix may be referenced by an applicant for a construction permit or operating license under 10 CFR part 50.
C. A licensee referencing the NuScale design certification is exempt from portions of the following regulation:
1. Paragraph (m) of 10 CFR 50.54 -Minimum Staffing. In lieu of these requirements, a licensee that references this appendix must comply with the following:
a. A senior operator licensed pursuant to part 55 of this chapter shall be present at the facility or readily available on call at all times during its operation, and shall be present at the facility during initial startup and approach to power, recovery from an unplanned or unscheduled shutdown or significant reduction in power, and refueling, or as otherwise prescribed in the facility license.
b. Licensees shall meet the following requirements:
i. Each licensee shall meet the minimum licensed operator staffing requirements identified in Table 1:
Table 1-Minimum Requirements per Shift for On-Site Staffing of NuScale Power Plants by Operators and Senior Operators Licensed Under 10 CFR Part 55
| Number of units operating (a nuclear power unit is considered to be operating when it is in MODE 1, 2, or 3 as defined by the unit's technical specifications) | Position | One to twelve units |
| One control room | ||
| None | Senior operator Operator | 1 2 |
| One to twelve | Senior operator Operator | 3 3 |
Source: Design Certification Application, Part 7, Section 6.1.3 , "Requested Action."
ii. Each facility licensee shall have at its site a person holding a senior operator license for all fueled units at the site who is assigned responsibility for overall plant operation at all times there is fuel in any unit. At all times any module is fueled, regardless of mode, there must be a licensed operator or senior operator in the control room.
iii. When a nuclear power unit is in MODE 1, 2, or 3, as defined by the unit's technical specifications, each licensee shall have a person holding a senior operator license for the nuclear power unit in the control room at all times. In addition to this senior operator, a second person who is either a licensed operator or licensed senior operator shall be present at the controls at all times. A third person who is either a licensed operator or licensed senior operator shall be in the control room envelope at all times.
iv. Each licensee shall have present, during alteration or movement of the core of a nuclear power unit (including fuel loading, fuel transfer, or movement of a module that contains fuel), a person holding a senior operator license or a senior operator license limited to fuel handling to directly supervise the activity and, during this time, the licensee shall not assign other duties to this person.
2. Appendix J to 10 CFR part 50, Type A testing-Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors.
V. Applicable Regulations
A. Except as indicated in paragraph B of this section, the regulations that apply to NuScale are in 10 CFR parts 20 , 50 , 52 , 73 , and 100 , codified as of February 21, 2023, that are applicable and technically relevant, as described in the final safety evaluation report.
B. The NuScale design is exempt from portions of the following regulations:
1. Paragraph (f)(2)(vi) of 10 CFR 50.34 and 10 CFR 50.46a -High point venting for the reactor coolant system and reactor pressure vessel head.
2. Paragraph (f)(2)(viii) of 10 CFR 50.34 -Post-accident sampling of the reactor coolant system and containment.
3. Paragraph (f)(2)(xiii) of 10 CFR 50.34 -Power supplies for pressurizer heaters.
4. Paragraph (f)(2)(xiv)(E) of 10 CFR 50.34 -Automatic closing of containment isolation systems on a high radiation signal.
5. Paragraph (f)(2)(xx) of 10 CFR 50.34 -Power from vital buses and emergency power sources for pressurizer level indication.
6. Paragraph (c)(2) of 10 CFR 50.44 -Combustible gas control.
7. Paragraph (a)(1)(i) of 10 CFR 50.46 -Applicability limited to reactor designs that use zircaloy or ZIRLO fuel rod cladding material.
8. Paragraph (c)(1) of 10 CFR 50.62 -Diverse equipment to initiate a turbine trip under conditions indicative of an anticipated transient without scram.
9 Appendix A of 10 CFR part 50-Electric Power Systems GDCs:
a. GDC 17-Electric power systems for safety-related functions;
b. GDC 18-Design to permit periodic inspection and testing of electric power systems;
c. GDC 34-Electric power systems for residual heat removal;
d. GDC 35-Electric power systems for emergency core cooling;
e. GDC 38-Electric power systems for containment heat removal;
f. GDC 41-Electric power systems for containment atmosphere cleanup; and
g. GDC 44-Electric power systems for cooling.
10. Appendix A to 10 CFR part 50, GDC 19-Equipment outside the control room with capability for cold shutdown of the reactor.
11. Appendix A to 10 CFR part 50, GDC 27-Demonstration of long-term shutdown under post-accident conditions with an assumed worst rod stuck out.
12. Appendix A to 10 CFR part 50, GDC 33-Reactor coolant makeup for protection against small breaks in the reactor coolant pressure boundary.
13. Appendix A to 10 CFR part 50, GDC 40-Periodic pressure and functional testing of containment heat removal system.
14. Appendix A to 10 CFR part 50, GDC 52-Design to allow periodic containment leakage rate testing.
15. Appendix A of 10 CFR part 50, GDCs 55, 56, and 57-Containment Isolation:
a. GDC 55-Isolation valves for certain reactor coolant pressure boundary lines penetrating containment;
b. GDC 56-Isolation valves for certain primary containment lines; and
c. GDC 57-Isolation valves for certain closed systems lines.
16. Appendix K to 10 CFR part 50-Emergency Core Cooling System Evaluation Models:
a. Section I.A.4-Heat generation rates from radioactive decay of fission products;
b. Section I.A.5-Rate of energy release, hydrogen generation, and cladding oxidation from the metal/water reaction;
c. Section I.B-Predicting cladding swelling and rupture;
d. Section I.C.1.b-Calculation of the discharge rate for all times after the discharging fluid has been calculated to be two-phase;
e. Section I.C.5.a-Post-critical heat flux correlations of heat transfer from the fuel cladding to the surrounding fluid; and
f. Section I.C.7.a-Calculation of cross-flow between the hot and average channel regions of the core during blowdown.
VI. Issue Resolution
A. The Commission has determined that the structures, systems, and components and design features of NuScale comply with the provisions of the Atomic Energy Act of 1954, as amended, and the applicable regulations identified in Section V of this appendix; and therefore, provide adequate protection to the health and safety of the public. A conclusion that a matter is resolved includes the finding that additional or alternative structures, systems, and components, design features, design criteria, testing, analyses, acceptance criteria, or justifications are not necessary for NuScale.
B. The Commission considers the following matters resolved within the meaning of § 52.63(a)(5) in subsequent proceedings for issuance of a COL, amendment of a COL, or renewal of a COL, proceedings held under § 52.103 , and enforcement proceedings involving plants referencing this appendix:
1. All nuclear safety issues associated with the information in the final safety evaluation report, Tier 1, Tier 2, and the rulemaking record for certification of the NuScale design, with the exception of the following:
a. generic TS and other operational requirements;
b. the adequacy of the design of the shield wall between the NuScale power module and the reactor building steam gallery to limit potential radiological doses consistent with the radiation zones specified in design certification application Part 2, Tier 2, Chapter 12, Figure 12.3-1, "Reactor Building Radiation Zone Map";
c. the adequacy of the design of the systems used for post-accident hydrogen and oxygen monitoring described in design certification application Part 2, Tier 2, Section 6.2.5 to meet the requirements of 10 CFR 50.34(f)(2)(vii) , 10 CFR 50.34(f)(2) (xxviii), and 10 CFR 52.47(a)(2)(iv) , with respect to radiological releases caused by leakage from these systems under accident conditions; and
d. 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 and reverse flow, consistent with the other design information regarding steam generator integrity described in DCA Part 2, Tier 2, Sections 3.9.1 , 3.9.2 , 5.4.1 , and 15.6.3 , and in accordance with 10 CFR part 50, GDC 4 and 31;
2. All nuclear safety and safeguards issues associated with the referenced information in the non-public documents in Tables 1.6-1 and 1.6-2 of Tier 2 of the DCD, which contain sensitive unclassified non-safeguards information (including proprietary information and security-related information) and safeguards information and which, in context, are intended as requirements in the generic DCD for the NuScale design;
3. All generic changes to the DCD under and in compliance with the change processes in paragraphs VIII.A.1 and VIII.B.1 of this appendix;
4. All exemptions from the DCD under and in compliance with the change processes in paragraphs VIII.A.4 and VIII.B.4 of this appendix, but only for that plant;
5. All departures from the DCD that are approved by license amendment, but only for that plant;
6. Except as provided in paragraph VIII.B.5.g of this appendix, all departures from Tier 2 under and in compliance with the change processes in paragraph VIII.B.5 of this appendix that do not require prior NRC approval, but only for that plant; and
7. All environmental issues concerning severe accident mitigation design alternatives associated with the information in the NRC's environmental assessment for NuScale (ADAMS Accession No. ML22004A006) and DCD Part 3, "Applicant's Environmental Report-Standard Design Certification," Revision 5, dated July 2020 (ADAMS Accession No. ML20224A512), for plants referencing this appendix whose site characteristics fall within the site parameters of the representative site specified in the NuScale environmental report.
C. The Commission does not consider operational requirements for an applicant or licensee who references this appendix to be matters resolved within the meaning of § 52.63(a)(5) . The Commission reserves the right to require operational requirements for an applicant or licensee who references this appendix by rule, regulation, order, or license condition.
D. Except under the change processes in Section VIII of this appendix, the Commission may not require an applicant or licensee who references this appendix to:
1. Modify structures, systems, and components or design features as described in the generic DCD;
2. Provide additional or alternative structures, systems, and components or design features not discussed in the generic DCD; or
3. Provide additional or alternative design criteria, testing, analyses, acceptance criteria, or justification for structures, systems, and components or design features discussed in the generic DCD.
E. The NRC will specify, at an appropriate time, the procedures to be used by an interested person who wishes to review portions of the design certification or references containing safeguards information or sensitive unclassified non-safeguards information (including proprietary information, such as trade secrets and commercial or financial information obtained from a person that are privileged or confidential (10 CFR 2.390 and 10 CFR part 9 ), and security-related information), for the purpose of participating in the hearing required by § 52.85 , the hearing provided under § 52.103 , or in any other proceeding relating to this appendix, in which interested persons have a right to request an adjudicatory hearing.
VII. Duration of This Appendix
This appendix may be referenced for a period of 15 years from February 21, 2023, except as provided for in §§ 52.55(b) and 52.57(b) . This appendix remains valid for an applicant or licensee who references this appendix until the application is withdrawn or the license expires, including any period of extended operation under a renewed license.
VIII. Processes for Changes and Departures
A. Tier 1 Information
1. Generic changes to Tier 1 information are governed by the requirements in § 52.63(a)(1) .
2. Generic changes to Tier 1 information are applicable to all applicants or licensees who reference this appendix, except those for which the change has been rendered technically irrelevant by action taken under paragraphs A.3 or A.4 of this section.
3. Departures from Tier 1 information that are required by the Commission through plant-specific orders are governed by the requirements in § 52.63(a)(4) .
4. Exemptions from Tier 1 information are governed by the requirements in §§ 52.63(b)(1) and 52.98(f) . The Commission will deny a request for an exemption from Tier 1, if it finds that the design change will result in a significant decrease in the level of safety otherwise provided by the design.
B. Tier 2 Information
1. Generic changes to Tier 2 information are governed by the requirements in § 52.63(a)(1) .
2. Generic changes to Tier 2 information are applicable to all applicants or licensees who reference this appendix, except those for which the change has been rendered technically irrelevant by action taken under paragraphs B.3, B.4, or B.5, of this section.
3. The Commission may not require new requirements on Tier 2 information by plant-specific order, while this appendix is in effect under § 52.55 or § 52.61 , unless:
a. A modification is necessary to secure compliance with the Commission's regulations applicable and in effect at the time this appendix was approved, as set forth in Section V of this appendix, or to ensure adequate protection of the public health and safety or the common defense and security; and
b. Special circumstances as defined in 10 CFR 50.12(a) are present.
4. An applicant or licensee who references this appendix may request an exemption from Tier 2 information. The Commission may grant such a request only if it determines that the exemption will comply with the requirements of 10 CFR 50.12(a) . The Commission will deny a request for an exemption from Tier 2, if it finds that the design change will result in a significant decrease in the level of safety otherwise provided by the design. The granting of an exemption to an applicant must be subject to litigation in the same manner as other issues material to the license hearing. The granting of an exemption to a licensee must be subject to an opportunity for a hearing in the same manner as license amendments.
5.
a. An applicant or licensee who references this appendix may depart from Tier 2 information, without prior NRC approval, unless the proposed departure involves a change to or departure from Tier 1 information, or the TS, or requires a license amendment under paragraph B.5.b or B.5.c of this section. When evaluating the proposed departure, an applicant or licensee shall consider all matters described in the plant-specific DCD.
b. A proposed departure from Tier 2, other than one affecting resolution of a severe accident issue identified in the plant-specific DCD or one affecting information required by § 52.47(a)(28) to address aircraft impacts, requires a license amendment if it would:
(1) Result in more than a minimal increase in the frequency of occurrence of an accident previously evaluated in the plant-specific DCD;
(2) Result in more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component important to safety and previously evaluated in the plant-specific DCD;
(3) Result in more than a minimal increase in the consequences of an accident previously evaluated in the plant-specific DCD;
(4) Result in more than a minimal increase in the consequences of a malfunction of a structure, system, or component important to safety previously evaluated in the plant-specific DCD;
(5) Create a possibility for an accident of a different type than any evaluated previously in the plant-specific DCD;
(6) Create a possibility for a malfunction of a structure, system, or component important to safety with a different result than any evaluated previously in the plant-specific DCD;
(7) Result in a design-basis limit for a fission product barrier as described in the plant-specific DCD being exceeded or altered; or
(8) Result in a departure from a method of evaluation described in the plant-specific DCD used in establishing the design bases or in the safety analyses.
c. A proposed departure from Tier 2, affecting resolution of an ex-vessel severe accident design feature identified in the plant-specific DCD, requires a license amendment if:
(1) There is a substantial increase in the probability of an ex-vessel severe accident such that a particular ex-vessel severe accident previously reviewed and determined to be not credible could become credible; or
(2) There is a substantial increase in the consequences to the public of a particular ex-vessel severe accident previously reviewed.
d. A proposed departure from Tier 2 information required by § 52.47(a)(28) to address aircraft impacts shall consider the effect of the changed design feature or functional capability on the original aircraft impact assessment required by 10 CFR 50.150(a) . The applicant or licensee shall describe, in the plant-specific DCD, how the modified design features and functional capabilities continue to meet the aircraft impact assessment requirements in 10 CFR 50.150(a)(1) .
e. If a departure requires a license amendment under paragraph B.5.b or B.5.c of this section, it is governed by 10 CFR 50.90 .
f. A departure from Tier 2 information that is made under paragraph B.5 of this section does not require an exemption from this appendix.
g. A party to an adjudicatory proceeding for either the issuance, amendment, or renewal of a license or for operation under § 52.103(a) , who believes that an applicant or licensee who references this appendix has not complied with paragraph VIII.B.5 of this appendix when departing from Tier 2 information, may petition to admit into the proceeding such a contention. In addition to complying with the general requirements of 10 CFR 2.309 , the petition must demonstrate that the departure does not comply with paragraph VIII.B.5 of this appendix. Further, the petition must demonstrate that the change bears on an asserted noncompliance with an ITAAC acceptance criterion in the case of a § 52.103 preoperational hearing, or that the departure bears directly on the amendment request in the case of a hearing on a license amendment. Any other party may file a response. If, on the basis of the petition and any response, the presiding officer determines that a sufficient showing has been made, the presiding officer shall certify the matter directly to the Commission for determination of the admissibility of the contention. The Commission may admit such a contention if it determines the petition raises a genuine issue of material fact regarding compliance with paragraph VIII.B.5 of this appendix.
C. Operational Requirements
1. Changes to NuScale design certification generic TS and other operational requirements that were completely reviewed and approved in the design certification rule and do not require a change to a design feature in the generic DCD are governed by the requirements in 10 CFR 50.109 . Changes that require a change to a design feature in the generic DCD are governed by the requirements in paragraphs A or B of this section.
2. Changes to NuScale design certification generic TS and other operational requirements are applicable to all applicants who reference this appendix, except those for which the change has been rendered technically irrelevant by action taken under paragraphs C.3 or C.4 of this section.
3. The Commission may require plant-specific departures on generic TS and other operational requirements that were completely reviewed and approved, provided a change to a design feature in the generic DCD is not required and special circumstances, as defined in 10 CFR 2.335 are present. The Commission may modify or supplement generic TS and other operational requirements that were not completely reviewed and approved or require additional TS and other operational requirements on a plant-specific basis, provided a change to a design feature in the generic DCD is not required.
4. An applicant who references this appendix may request an exemption from the generic TS or other operational requirements. The Commission may grant such a request only if it determines that the exemption will comply with the requirements of § 52.7 . The granting of an exemption must be subject to litigation in the same manner as other issues material to the license hearing.
5. A party to an adjudicatory proceeding for the issuance, amendment, or renewal of a license, or for operation under § 52.103(a) , who believes that an operational requirement approved in the DCD or a TS derived from the generic TS must be changed, may petition to admit such a contention into the proceeding. The petition must comply with the general requirements of § 2.309 of this chapter and must either demonstrate why special circumstances as defined in § 2.335 of this chapter are present or demonstrate that the proposed change is necessary for compliance with the Commission's regulations in effect at the time this appendix was approved, as set forth in Section V of this appendix. Any other party may file a response to the petition. If, on the basis of the petition and any response, the presiding officer determines that a sufficient showing has been made, the presiding officer shall certify the matter directly to the Commission for determination of the admissibility of the contention. All other issues with respect to the plant-specific TS or other operational requirements are subject to a hearing as part of the licensing proceeding.
6. After issuance of a license, the generic TS have no further effect on the plant-specific TS. Changes to the plant-specific TS will be treated as license amendments under 10 CFR 50.90 .
IX. [Reserved]
X. Records and Reporting
A. Records
1. The applicant for this appendix shall maintain a copy of the generic DCD that includes all generic changes that are made to Tier 1 and Tier 2, and the generic TS and other operational requirements. The applicant shall maintain the sensitive unclassified non-safeguards information (including proprietary information and security-related information) and safeguards information referenced in the generic DCD for the period that this appendix may be referenced, as specified in Section VII of this appendix.
2. An applicant or licensee who references this appendix shall maintain the plant-specific DCD to accurately reflect both generic changes to the generic DCD and plant-specific departures made under Section VIII of this appendix throughout the period of application and for the term of the license (including any periods of renewal).
3. An applicant or licensee who references this appendix shall prepare and maintain written evaluations that provide the bases for the determinations required by Section VIII of this appendix. These evaluations must be retained throughout the period of application and for the term of the license (including any periods of renewal).
4.
a. The applicant for NuScale shall maintain a copy of the aircraft impact assessment performed to comply with the requirements of 10 CFR 50.150(a) for the term of the certification (including any period of renewal).
b. An applicant or licensee who references this appendix shall maintain a copy of the aircraft impact assessment performed to comply with the requirements of 10 CFR 50.150(a) throughout the pendency of the application and for the term of the license (including any periods of renewal).
B. Reporting
1. An applicant or licensee who references this appendix shall submit a report to the NRC containing a brief description of any plant-specific departures from the DCD, including a summary of the evaluation of each departure. This report must be filed in accordance with the filing requirements applicable to reports in § 52.3 .
2. An applicant or licensee who references this appendix shall submit updates to its plant-specific DCD, which reflect the generic changes to and plant-specific departures from the generic DCD made under Section VIII of this appendix. These updates shall be filed under the filing requirements applicable to final safety analysis report updates in 10 CFR 50.71(e) and 52.3 .
3. The reports and updates required by paragraphs X.B.1 and X.B.2 of this appendix must be submitted as follows:
a. On the date that an application for a license referencing this appendix is submitted, the application must include the report and any updates to the generic DCD.
b. During the interval from the date of application for a license to the date the Commission makes its finding required by § 52.103(g) , the report must be submitted semiannually. Updates to the plant-specific DCD must be submitted annually and may be submitted along with amendments to the application.
c. After the Commission makes the finding required by § 52.103(g) , the reports and updates to the plant-specific DCD must be submitted, along with updates to the site-specific portion of the final safety analysis report for the facility, at the intervals required by 10 CFR 50.59(d)(2) and 50.71(e)(4) , respectively, or at shorter intervals as specified in the license.
10 C.F.R. 52 app G to Part 52