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Federal Register / Vol. 86, No. 124 / Thursday, July 1, 2021 / Proposed Rules 26. TR051649417NPA, Evaluation Methodology for Stability Analysis of the NuScale Power Module, March 2020, Revision 1, Docket: PROJ0769.
27. TR051649422NPA, Loss-ofCoolant Accident Evaluation Model, July 2020, Revision 2, Docket: PROJ0769.
28. TR061648793NPA, Nuclear Analysis Codes and Methods Qualification, November 2018, Revision 1, Docket:
PROJ0769.
29. TR061649121NP, NuScale Instrument Setpoint Methodology Technical Report, May 2020, Revision 3, Docket: 52
048.
30. TR071650350NPA, Rod Ejection Accident Methodology, June 2020, Revision 1, Docket: PROJ0769.
31. TR071650351NPA, NuScale Applicability of AREVA Method for the Evaluation of Fuel Assembly Structural Response to Externally Applied Forces, April 2020, Revision 1, Docket: PROJ0769.
32. TR071650424NP, Combustible Gas Control, March 2019, Revision 1, Docket:
PROJ0769.
33. TR071650439NP, NuScale Comprehensive Vibration Assessment Program Analysis Technical Report, July 2019, Revision 2, Docket: 52048.
34. TR081516497NPA, Safety Classification of Passive Nuclear Power Plant Electrical Systems, January 2018, Revision 1, Docket: PROJ0769.
35. TR081649833NP, Fuel Storage Rack Analysis, November 2018, Revision 1, Docket: 52048.
36. TR081650796NP, Loss of Large Areas Due to Explosions and Fires Assessment, June 2019, Revision 1, Docket:
52048.
37. TR081650797, Mitigation Strategies for Loss of All AC Power Event, October 2019, Revision 3, Docket: 52048, NuScale Nonproprietary.
38. TR081651127NP, NuFuel-HTP2TM
Fuel and Control Rod Assembly Designs, December 2019, Revision 3, Docket: 52048.
39. TR081861384NP, Pipe Rupture Hazards Analysis, July 2019, Revision 2, Docket No.: 52048.
40. TR091517564NPA, Subchannel Analysis Methodology, February 2019, Revision 2, Docket: PROJ0769.
41. TR091517565NPA, Accident Source Term Methodology, February 2020, Revision 4, Docket: PROJ0769.
42. TR091651299NP, Long-Term Cooling Methodology, May 2020, Revision 3, Docket: 52048.
43. TR091651502NP, NuScale Power Module Seismic Analysis, April 2019, Revision 2, Docket: 52048.
44. TR091756119NP, CNV Ultimate Pressure Integrity, June 2019, Revision 1, Docket No. 52048.
45. TR091860894NP, NuScale Comprehensive Vibration Assessment Program Measurement and Inspection Plan Technical Report, August 2019, Revision 1, Docket No.: 52048.
46. NPTR1010859NPA, NuScale Topical Report: Quality Assurance Program Description for the NuScale Power Plant, May 2020, Revision 5, Docket: PROJ0769, NuScale Nonproprietary.

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47. TR101518177NP, Pressure and Temperature Limits Methodology, October 2018, Revision 2, Docket: 52048.
48. TR101518653NPA, Design of the Highly Integrated Protection System Platform, May 2017, Revision 2, Docket:
PROJ0769.
49. TR101651669NP, NuScale Power Module Short-Term Transient Analysis, July 2019, Revision 1, Docket: 52048.
50. TR111651962NP, NuScale Containment Leakage Integrity Assurance, May 2019, Revision 1, Docket: 52048.
51. TR111652065NP, Effluent Release GALE Replacement Methodology and Results, November 2018, Revision 1, Docket:
52048.
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 entirely 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 applicants 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 52.47a that is not within the scope of this appendix;

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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 12.31, 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 50.34f2xxviii 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 50.34f2vii, 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 52.47a2iv; and i. Information demonstrating that the criteria of 10 CFR part 20 and the requirements of 10 CFR part 50, appendix A, General Design Criterion GDC 4 and GDC 31
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 applicants 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.
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 DATE 120 DAYS
AFTER DATE OF PUBLICATION OF FINAL
RULE IN THE Federal Register, that are applicable and technically relevant, as described in the final safety evaluation report.

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