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Category:Letter
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Boron Redistribution ML20195A5872020-07-13013 July 2020 LLC - Submittal of Draft Operator Licensing and Examination Standard for NuScale Small Modular Reactors ML20195C7662020-07-13013 July 2020 LLC Request for Standard Design Approval Based on the NuScale Standard Plant Design Certification Application ML20192A3262020-07-10010 July 2020 LLC, Submittal of Environmental Report: Revision Status ML20198M3932020-07-0202 July 2020 LLC Submittal of Revised Packing Slip for Submittal of the NuScale Standard Plant Design Certification Application, Revision 4.1, Dated June 19, 2020 ML20174A3472020-07-0101 July 2020 OEDO-20-00220 - Area of Focus - Probabilistic Risk Assessment and Emergency Core Cooling System Valve Performance ML20184A2872020-07-0101 July 2020 LLC Submittal of Presentation Materials Entitled ACRS Full Committee Presentation: NuScale Topic - Boron Redistribution and General Design Criterion 33, PM-0720-70785, Revision 0 ML20181A4322020-06-22022 June 2020 Final SER for NuScale TR-0516-49416 NON-Loss-of-Coolant Analysis Model, Rev 3 (Letter) ML20181A2702020-06-22022 June 2020 Final SER for NuScale TR-0516-49422 Loss-of-Coolant Analysis Model, Rev 2 (Letter) ML20198M3922020-06-19019 June 2020 LLC - 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Response to the ACRS Letter on Combustible Gas Monitoring ML20133K0882020-05-12012 May 2020 LLC, Submittal of Presentation Materials Entitled Public Meeting Presentation: Topic - Emergency Core Cooling System (ECCS) Boron Distribution (Closed Session), PM-0420-69512, Revision 0 ML20133J9142020-05-11011 May 2020 LLC Submittal of Presentation Materials Entitled Public Meeting Presentation: Topic - Emergency Core Cooling System (ECCS) Boron Distribution, PM-0420-69511, Revision 0 ML20112F4552020-05-0101 May 2020 LLC, Design Certification Application Phases 5 and 6 Review Status ML20107F8492020-05-0101 May 2020 OEDO-2000140 - NuScale Area of Focus - Helical Tube Steam Generator Design ML20104A0792020-04-27027 April 2020 OEDO-20-00115 - Safety Evaluation Report for Topical Report TR-0516-49416, Revision 2, Non-Loss-of-Coolant Accident Analysis Methodology ML20099H0802020-04-0808 April 2020 LLC - Submittal of Presentation Materials Entitled NRC Public Meeting: Revisions to Nuscale'S EPZ Sizing Methodology Topical Report, PM-0420-69598, Revision 0 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 LLC Submittal of Presentation Materials Entitled ACRS Full Committee Presentation NuScale Topic-Probabilistic Risk Assessment with a Focus on Emergency Core Cooling System Analysis PM-0420-69559, Revision 0 ML20092L8992020-04-0101 April 2020 LLC - Submittal of Updates to Standard Plant Design Certification Application, Revision 4 ML20072M6682020-03-30030 March 2020 Response to NuScale Letter Dated February 24, 2020, on Planned SDA Application Content 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] Category:Response to Request for Additional Information (RAI)
MONTHYEARRAIO-0420-69855, LLC, Submittal of Corrected Response to NRC Request for Additional Information No. 284 (Erai No. 9225) on the NuScale Design Certification2020-04-30030 April 2020 LLC, Submittal of Corrected Response to NRC Request for Additional Information No. 284 (Erai No. 9225) on the NuScale Design Certification ML19332A1202019-11-27027 November 2019 LLC Supplemental Response to NRC Request for Additional Information No. 484 (Erai No. 8930) on the NuScale Design Certification Application ML19304B4712019-10-31031 October 2019 LLC Supplemental Response to NRC Request for Additional Information No. 466 (Erai No. 9482) on the NuScale Design Certification Application ML19296D8052019-10-23023 October 2019 LLC Response to NRC Request for Additional Information No. 526 (Erai No. 9719) on the NuScale Design Certification Application ML19283E5302019-10-10010 October 2019 LLC Supplemental Response to NRC Request for Additional Information No. 522 (Erai No. 9681) on the NuScale Design Certification Application ML19260G7352019-10-0707 October 2019 Summary of Public Meeting with NuScale to Discuss Response to RAI 9681 ML19266A5872019-09-23023 September 2019 LLC Supplemental Response to NRC Request for Additional Information No. 518 (Erai No. 9659) on the NuScale Design Certification Application ML19262G9742019-09-19019 September 2019 LLC - Submittal of Changes to Final Safety Analysis Report, Tier 1, Section 3.11, Reactor Building and Section 3.13, Control Building, and Tier 2, Section 3.8.4, Design of Category I Structure and Section 14.3, Certified ... ML19262G5762019-09-19019 September 2019 LLC - Submittal of Changes to Final Safety Analysis Report, Section 14.2, Initial Plant Test Program, Table 14.2-2, Pool Cleanup Systems Test #2, and Table 14.2-50, Module Assembly Equipment Test #50 ML19259B8102019-09-16016 September 2019 LLC Supplemental Response to NRC Request for Additional Information No. 205 (Erai No. 9044) on the NuScale Design Certification Application ML19259A0922019-09-16016 September 2019 LLC Response to NRC Request for Additional Information No. 525 (Erai No. 9705) on the NuScale Design Certification Application ML19238A3722019-08-26026 August 2019 LLC Supplemental Response to NRC Request for Additional Information No. 197 (Erai No. 9051) on the NuScale Design Certification Application ML19238A3662019-08-23023 August 2019 LLC - Response to NRC Request for Additional Information No. 523 (Erai No. 9682) on the NuScale Design Certification Application ML19215A0032019-08-0202 August 2019 LLC - Supplemental Response to NRC Request for Additional Information No. 202 (Erai No. 8911) on the NuScale Design Certification Application ML19215A0062019-08-0202 August 2019 LLC - 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Supplemental Response to NRC Request for Additional Information No. 325 (Erai No. 9268) on the NuScale Design Certification Application ML19164A1452019-06-13013 June 2019 LLC - Submittal of Containment Response Analysis Methodology Technical Report, TR-0516 -49 08 4, Revision 1 ML19157A3262019-06-0606 June 2019 LLC - Supplemental Response to NRC Request for Additional Information No. 232 (Erai No. 9113) on the NuScale Design Certification Application ML19154A6222019-06-0303 June 2019 LLC Supplemental Response to NRC Request for Additional Information No. 202 (Erai No. 8911) on the NuScale Design Certification Application ML19154A6052019-06-0303 June 2019 LLC Response to NRC Request for Additional Information No. 514 (Erai No. 9645) on the NuScale Design Certification Application ML19151A8372019-05-31031 May 2019 LLC - Supplemental Response to NRC Request for Additional Information No. 377 (Erai No. 9380) on the NuScale Design Certification Application ML19140A4592019-05-20020 May 2019 LLC Supplemental Response to NRC Request for Additional Information No. 401 (Erai No. 9447) on the NuScale Design Certification Application ML19137A2902019-05-17017 May 2019 LLC Supplemental Response to NRC Request for Additional Information No. 156 (Erai No. 9031) on the NuScale Design Certification Application ML19137A2872019-05-15015 May 2019 LLC Response to NRC Request for Additional Information No. 519 (Erai No. 9656) on the NuScale Design Certification Application ML19126A2942019-05-0606 May 2019 LLC Supplemental Response to NRC Request for Additional Information No. 26 (Erai No. 8840) on the NuScale Design Certification Application ML19122A5092019-05-0202 May 2019 LLC - Supplemental Response to NRC Request for Additional Information No. 494 (Erai No. 9548)on the Design Certification Application ML19121A6002019-05-0101 May 2019 LLC - Supplemental Response to NRC Request for Additional Information No. 202 (Erai No. 8911) on Design Certification Application 2020-04-30
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Text
RAIO-0419-65197 April 11, 2019 Docket No.52-048 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852-2738
SUBJECT:
NuScale Power, LLC Response to NRC Request for Additional Information No.
518 (eRAI No. 9659) on the NuScale Design Certification Application
REFERENCE:
U.S. Nuclear Regulatory Commission, "Request for Additional Information No.
518 (eRAI No. 9659)," dated March 04, 2019 The purpose of this letter is to provide the NuScale Power, LLC (NuScale) response to the referenced NRC Request for Additional Information (RAI).
The Enclosure to this letter contains NuScale's response to the following RAI Question from NRC eRAI No. 9659:
19-39 This letter and the enclosed response make no new regulatory commitments and no revisions to any existing regulatory commitments.
If you have any questions on this response, please contact Rebecca Norris at 541-452-7539 or at rnorris@nuscalepower.com.
Sincerely, Zackary W. Rad Director, Regulatory Affairs NuScale Power, LLC Distribution: Gregory Cranston, NRC, OWFN-8H12 Samuel Lee, NRC, OWFN-8H12 Rani Franovich, NRC, OWFN-8H12 : NuScale Response to NRC Request for Additional Information eRAI No. 9659 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvalis, Oregon 97330, Office: 541.360.0500, Fax: 541.207.3928 www.nuscalepower.com
RAIO-0419-65197 :
NuScale Response to NRC Request for Additional Information eRAI No. 9659 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvalis, Oregon 97330, Office: 541.360.0500, Fax: 541.207.3928 www.nuscalepower.com
Response to Request for Additional Information Docket No.52-048 eRAI No.: 9659 Date of RAI Issue: 03/04/2019 NRC Question No.: 19-39 Regulatory basis 10 CFR 52.47(a)(27) requires a description of the design-specific probabilistic risk assessment (PRA) and its results.
Discussion Standard Review Plan, Chapter 19.0, Revision 3, page 19.0-22 identifies the need for in-depth NRC review of refueling operations for small, modular reactors, which are different from traditional LWRs, to ensure that the PRA model is of acceptable scope and level of detail. This same page also requires staff to verify that applicants for plants with multiple modules use a systematic process to identify accident sequences, including significant human errors that could lead to core damage or large release from multiple modules.
Section 19.1.7.4 of the NuScale Design Certification Application (DCA) discusses how a module dropped during refueling transport might impact other modules. Rev. 1 of the DCA states that if the module is dropped on an operating module near the top, it could damage the DHRS piping or heat exchangers. In Rev. 2 of the DCA, NuScale added that "additional pipe breaks may occur, leading to a [chemical and volume control system] CVCS line break outside containment." Additionally, Rev. 1 of the DCA states that if the operating module was struck near the bottom, the safety systems would remain nominally available, whereas Rev. 2 replaced this conclusion with "the collision is expected to cause a torque about the module support lugs, resulting in similar stresses to the piping on top of the operating module." The risk insights from this evaluation, which is the same in both revisions, are that a dropped module may incur core damage while the struck modules incur initiating events at full power.
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Because Rev. 2 of the DCA postulates additional damage to the operating module beyond what was described in Rev. 1, the staff needs additional information to conclude the qualitative multi-module risk assessment is technically adequate and complete.
Request for Additional Information Provide justification that multi-module risk insights for the struck module that is assumed to be operating at full power are unaffected by the additional damage described in Rev. 2 of the DCA.
Specifically, describe which pipes in the CVCS, decay heat removal system and the containment flooding and drain system are assumed to fail and why. Also explain if the capability of the containment isolation valves to close is compromised, given that the strike to the operating module has sufficient force to cause pipe breaks.
NuScale Response:
The risk insights are unaffected by the changes in the description of potential multiple-module effects included in FSAR Revision 2. As described in FSAR Section 19.1.7.4, the effects of a module being struck by a dropped module are determined by engineering judgment.
Accordingly, the modified FSAR wording in Revision 2 is intended to depict a range of possible effects on an operating module rather than a prediction of a specific effect.
Potential damage to the decay heat removal system (DHRS) heat exchangers is identified in the FSAR because the heat exchangers are located external to the containment; damage to these heat exchangers could affect secondary side heat removal, which is considered in the secondary side line break initiating event TGS---FMSLB-UD, described in FSAR Table 19.1-8.
Potential damage to the chemical and volume control system (CVCS) piping is identified because it could result in a pathway from the reactor coolant system to outside of containment, as considered in the CVCS pipe break outside containment initiating events CVCS--ALOCA-COC and CVCS--ALOCA-LOC, described in FSAR Table 19.1-8. Several conditions are necessary to result in such an impact-induced pipe break on an operating module. For example, in a postulated "controlled" module drop (i.e., the module is still attached to the module lifting adapter and crane) the dropped module is likely to remain vertical, precluding impact on an operating module. In a postulated "uncontrolled" module drop (e.g., catastrophic failure of both of the redundant rope reeving systems) which could result in module tipping, the geometry of the refueling pool, operating bay, and bay walls is such that a dropped module is unlikely NuScale Nonproprietary
to impact an operating module; additionally, the steel module platform columns and cross braces minimize the potential for a significant impact.
Considering only the probability of a load drop, the contribution of a potential module drop to the initiating event frequency of an operating module is judged to be negligible both in absolute terms and in comparison to the frequency of a randomly occurring pipe break outside of containment. The frequency of a dropped module is about 1E-7 per year as indicated in Table 19.1-68 and the initiating event frequency for a CVCS pipe break outside containment is about 4E-4 per module critical year as indicated in FSAR Table 19.1-8.
As discussed in FSAR Section 6.2.4.1, there are two containment isolation valves (CIVs) in each line that penetrates the containment boundary and connects to the reactor coolant pressure boundary. The CIVs on the CVCS line are open during power operation. The CIVs fail closed on loss of power or loss of hydraulic pressure and are located on top of the containment vessel head and under the module platform. Thus, if failure of a CVCS line on an operating module were postulated in response to a dropped module, impact damage to piping would likely occur downstream of the CIVs, and therefore the probability of failure-to-close of the CVCS CIVs is judged to be small. Because the containment flooding and drain system (CFDS) is normally isolated during module operation (i.e., the CFDS CIVs are closed) and not connected to the reactor coolant system, a postulated impact-induced CFDS pipe break would not lead to a loss of coolant or compromise the ability of the CIVs to maintain containment integrity.
Impact on DCA:
There are no impacts to the DCA as a result of this response.
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