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Federal Register / Vol. 86, No. 124 / Thursday, July 1, 2021 / Proposed Rules
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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. A COL
applicant would be 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. A COL
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 NRCs review of the NuScale design.
Therefore, the NRC has determined that this information can be provided by a COL applicant that references this appendix without a demonstrable impact on safety or standardization.
Appendix G to 10 CFR part 52, Section VI, Issue Resolution, would clarify that this issue is not resolved within the meaning of 52.63a5, and Section IV, Additional Requirements and Restrictions, would state 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 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
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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. A COL applicant would be required 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 COL 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 would likely 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 COL application review to provide reasonable assurance of adequate protection. Therefore, the NRC has determined that this information can be provided by a COL
applicant that references this appendix without a demonstrable impact on safety or standardization. Appendix G to 10 CFR part 52, Section VI, Issue Resolution, would clarify that this issue is not resolved within the meaning of 52.63a5, and Section IV, Additional Requirements and Restrictions, would state 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, 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 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

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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 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 applicants 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 nuclear 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 part 20 and 10
CFR part 50, appendix A, General Design Criterion 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,
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