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Federal Register / Vol. 86, No. 124 / Thursday, July 1, 2021 / Proposed Rules
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, an appropriate method of analysis has not been provided to the NRC.
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 NuScales 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 TR101651669
describe validation for determining density wave oscillation thermalhydraulic 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 acknowledges 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
VerDate Sep<11>2014

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TR101651669, NuScale Power Module Short-Term Transient Analysis, incorporated by reference in DCA Part 2, Tier 2, Table 1.62, 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 would 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 would be accomplished 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 NRCs review of the design of the NuScale nuclear power plant. Therefore, the NRC has determined that this information can be provided by a COL 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, 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.

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IV. Technical Issues Associated With the NuScale Design The NRC identified significant technical issues associated with the following design areas that were resolved by NuScale Power during the review:
Comprehensive vibration assessment program;
Containment safety analysis;
Emergency core cooling system inadvertent actuation block valve;
Conformance with GDC 27, Combined Reactivity Control Systems Capability, of appendix A, General Design Criteria for Nuclear Power Plants, to 10 CFR part 50;
Absence of safety-related Class 1E
alternating current AC or direct current DC electrical power;
Accident source term methodology;
Boron redistribution during passive cooling modes.
In addition, the NRC granted 17
exemptions from 10 CFR part 50 to address various aspects of NuScales design.
A. Comprehensive Vibration Assessment Program The NuScale comprehensive vibration assessment program limits potentially adverse effects from flow, acoustic, and mechanically induced vibrations and resonances on NuScale power module components, including the helical coil steam generators. The NuScale steam generators are different from those of operating pressurized-water reactors in that the primary reactor coolant is on the outside of the steam generator tubes and the steam is on the inside. Because of this design, there is the possibility of density wave oscillation instabilities in the secondary coolant which could challenge the integrity of the tubes. The NRCs review and findings, including independent analyses and observation of vibration testing, are documented in detail in Chapter 3, Design of Structures, Components, Equipment, and Systems, Section 3.9.2, Dynamic Testing and Analysis of Systems, Structures, and Components, of the final safety evaluation report. The review focused on assuring that the design of the helical coil steam generator tubes would not result in issues with flow-induced vibration.
As part of the comprehensive vibration assessment, the NRC also reviewed and found acceptable the steam generator tube margin against fluid-elastic instability, steam generator tube margin against vortex shedding, control rod drive shaft margin against vortex shedding, in-core instrument guide tube against vortex shedding,
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