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Category:Letter
MONTHYEARML23180A1512023-06-29029 June 2023
LLC, Request for Exemption to the Reporting Requirements of 10 CFR 50.46(a)(3)
ML21102A3072021-04-15015 April 2021
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OEDO-20-00292-Response to the Advisory Committee on Reactor Safeguards Letter on NuScale Power, LLC, Report on the Safety Aspects of the NuScale Small Modular Reactor
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Area of Focus - Boron Redistribution
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LLC Request for Standard Design Approval Based on the NuScale Standard Plant Design Certification Application
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ML20150C5172020-05-29029 May 2020
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ML20150C8812020-05-29029 May 2020
LLC Submittal of Presentation Materials Entitled ACRS Full Committee Presentation: NuScale Topic - Boron Redistribution and Associated Design and DCA Changes, PM-0620-70244, Revision 0
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ML20141L8082020-05-20020 May 2020
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ML20141N0122020-05-20020 May 2020
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ML20098G2372020-04-0707 April 2020
Nuscale Power, LLC Submittal of Remaining Closure Items for the Emergency Core Cooling System Valve Failure Mode Effects Analysis Audit Items
ML20097F1922020-04-0606 April 2020
Nuscale Power, LLC Submittal of Presentation Materials Entitled ACRS Full Committee Presentation: Nuscale Topic - Hydrogen/Oxygen Monitoring, PM-0420-69518, Revision 0
ML20094H6742020-04-0303 April 2020
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LLC - Submittal of Updates to Standard Plant Design Certification Application, Revision 4
ML20072M6682020-03-30030 March 2020
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ML20072H3332020-03-0909 March 2020
LLC - Submittal of Presentation Materials Entitled Public Meeting Presentation: Topic - Emergency Core Cooling System (ECCS) Boron Distribution, PM-0320-69218, Revision 0
ML20057D9002020-03-0606 March 2020
Submittal of Errata to Final SE for NuScale Power, LLC TR-1010-859-NP-A, Quality Assurance Program Description for the NuScale Power Plant
ML20062F7262020-03-0505 March 2020
Request for Withholding Information from Public Disclosure for Nuscale Power, LLC Letter Public
ML20069A1572020-03-0404 March 2020
LLC - Submittal of Presentation Materials Entitled ACRS Full Committee Presentation: NuScale Topical Report-Rod Ejection Accident Methodology, PM-0320-69146, Revision 0
ML20069A1772020-03-0404 March 2020
LLC - Submittal of Presentation Materials Entitled ACRS Full Committee Presentation: NuScale Topical Report-Non-Loss-of-Coolant Accident, PM-0320-69141, Revision 0
ML20069A9632020-03-0404 March 2020
LLC Submittal of Presentation Materials Entitled ACRS Full Committee Presentation: NuScale Topical Report, Loss-of-Coolant Accident Evaluation Model, PM-0320-69138, Revision 0
ML20066G2802020-03-0303 March 2020
LLC, Submittal of Presentation Materials Entitled ACRS Subcommittee Presentation: NuScale Topic - Hydrogen Monitoring, PM-0220-69071, Revision 0
ML20066G2882020-02-28028 February 2020
LLC Submittal of Presentation Materials Titled ACRS Full Committee Presentation: NuScale - Steam Generator Design (Closed Session), PM-0220-69053, Revision 0
2023-06-29
[Table view] |
Text
May 1, 2020 Matthew W. Sunseri, Chairman Advisory Committee on Reactor Safeguards U.S. Nuclear Regulatory Commission Washington, DC 20555-0001
SUBJECT:
NUSCALE AREA OF FOCUS - HELICAL TUBE STEAM GENERATOR DESIGN
Dear Mr. Sunseri:
In your letter dated March 24, 2020 (Agencywide Documents Access and Management System Accession No. ML20091G387), the Advisory Committee on Reactor Safeguards (ACRS or the Committee) reported on the Committees review of the U.S. Nuclear Regulatory Commission (NRC) staffs safety evaluation (SE) for the NuScale Power, LLC. (NuScale), area of focus -
Helical Tube Steam Generator Design. I appreciate the time and effort that the ACRS has devoted to this review, as reflected in meetings held with the ACRS Subcommittee on February 4, 2020, and the ACRS Full Committee held between March 5 - 6, 2020.
Your letter offered the following conclusions and recommendations:
- 1. The design and performance of the steam generators have not yet been sufficiently validated because of uncertainties associated with unstable density wave oscillations (DWO) on the steam generator secondary side.
- 2. Accelerated wear of the alloy 690TT steam generator tubing material is a potential concern.
- 3. Having determined that steam generator integrity is not resolved, NuScale and the staff have proposed the following solutions.
- a. The staff has proposed that the steam generator design not receive finality in the NuScale design certification.
- b. NuScale has proposed a combined license (COL) item and Inspections, Tests, Analyses and Acceptance Criteria (ITAAC) to address steam generator DWO.
- 4. Successful completion of these activities will address our concerns on steam generator performance at the design stage. Some uncertainty will remain until a NuScale Power Module is built and operated.
The NRC staff agrees with ACRS Conclusions and Recommendations items 1, 3, and 4. The staff has also found that NuScale has not yet sufficiently validated the design and performance of the steam generators to support its design certification application because of uncertainties associated with potential DWO on the steam generator secondary side. Therefore, the staff has proposed that the steam generator design not receive finality in the NuScale design certification.
M. Sunseri As such, the COL applicant will be required to provide information, such as analysis or testing results, to demonstrate the design and performance of the steam generators associated with DWO. Once a COL is issued, the COL holder will be required to confirm the structural integrity of the steam generators by the successful completion of inspections, tests, and acceptance criteria related to the American Society of Mechanical Engineers Boiler and Pressure Vessel Code component analysis, and by the implementation of a Comprehensive Vibration Assessment Program, which would include analysis and testing. During the initial startup testing, the COL holder will confirm that the performance of the steam generators provides acceptable vibration and stress in the reactor internals.
Regarding Conclusion and Recommendation item 2, the NRC staff agrees that potential wear of the steam generator tubing is a concern, regardless of the material used for the tubing. The staff agrees that rapid forms of degradation could cause wear of the steam generator tubes and should be addressed in the reactor design. The staff has proposed that the design and performance of the steam generator, with respect to DWO, not receive finality in the NuScale design certification. Rapid wear could also result from flow-induced vibration in the steam generators, which a COL applicant will need to address in its development of a Comprehensive Vibration Assessment Program. Similarly, a COL applicant will address progressive forms of tubing wear in its steam generator inspection program included in the plant technical specifications. The NRC staff agrees that NuScale should continue to develop the wear model for the steam generator tubes as it finalizes the reactor design. As such, regardless of the type of wear analysis that is performed, actual wear patterns in the steam generators will be revealed during operation, and they will be managed by the COL holder.
The NRC staff appreciates the ACRS review of this highly complex issue. The NRC staff plans to issue its SE for the NuScale design certification in September 2020.
Sincerely, Robert Digitally signed by Robert M. Taylor M. Taylor Date: 2020.05.01 13:56:47 -04'00' Ho K. Nieh, Director Office of Nuclear Reactor Regulation Docket No.: 52-048 cc: Chairman Svinicki Commissioner Baran Commissioner Caputo Commissioner Wright SECY
Pkg: ML20105A222 Ltr: ML20107F849 *via e-mail NRR-106 OFFICE NRR/DNRL/NRLB: PM NRR/DNRL/NRLB: LA QTE NRR/DSS/SNRB: BC NAME MJohnson* CSmith* QTE* RPatton*
DATE 04/13/2020 04/15/2020 04/14/2020 04/15/2020 OFFICE NRR/DEX/EMIB: BC (Acting) NRR/DNRL/NCSG: BC NRR/DNRL/NRLB: BC NRR/DSS :D NAME TScarbrough* SBloom* MDudek* JDonoghue*
DATE 04/15/2020 04/15/2020 04/17/2020 04/24/2020 OFFICE NRR/DEX: D NRR/DNRL: D NRR: D NAME EBenner* ABradford* HNieh (RTaylor for)*
DATE 04/21/2020 04/21/2020 05/01/2020