|
|---|
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
OEDO-21-00155 - Lessons-Learned from the Design Certification Review of the NuScale Power, LLC, Small Modular Reactor
ML21050A4312021-02-19019 February 2021
LLC - Lessons-Learned from the Design Certification Review of the NuScale Power, LLC Small Modular Reactor
ML20247J5642020-09-11011 September 2020
Standard Design Approval for the NuScale Power Plant Based on the NuScale 600 Standard Plant Design Certification Application
ML20231A8042020-08-28028 August 2020
Final Safety Evaluation Report for the NuScale Standard Plant Design
ML20224A4602020-08-25025 August 2020
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
ML20231A5982020-08-25025 August 2020
OEDO-20-00285_NuScale Area of Focus - Boron Redistribution
ML20210M8902020-07-29029 July 2020
Area of Focus - 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 - Submittal of the NuScale Standard Plant Design Certification Application, Revision 4.1
ML20171A7312020-06-19019 June 2020
LLC, Submittal of Riser Flow Hole Methodology and Associated Changes to Final Safety Analysis Report Incorporating Its Use
ML20157A2232020-06-0303 June 2020
Letter to NuScale Requesting -A for TR-0716-50350
ML20150C5172020-05-29029 May 2020
LLC Submittal of Presentation Materials Entitled NRC Public Meeting Presentation: Boron Redistribution and Associated Design and DCA Changes, PM-0620-70336, Revision 0
ML20150E1772020-05-29029 May 2020
LLC - Submittal of Presentation Materials Entitled ACRS Full Committee Presentation: NuScale Topic - Extended Dhrs Operation and RCS Boron Redistribution (Closed Session), PM-0620-70243, Revision 0
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
ML20149M1192020-05-28028 May 2020
LLC Summary of Impacts to Erai 8930 Response and Discussion on the Exemption from General Design Criterion 33
ML20141L8082020-05-20020 May 2020
LLC Submittal of Containment Response Analysis Methodology Technical Report, TR-0516-49084, Revision 3
ML20141N0122020-05-20020 May 2020
LLC Submittal of Changes to Final Safety Analysis Report, Section 6.2, Containment Systems, Section 6.3, Emergency Core Cooling System, and Technical Report TR-0516-49084, Containment Response Analysis Methodology Technical Report
ML20141M7642020-05-20020 May 2020
LLC Submittal of Nuclear Steam Supply System Advanced Sensor Technical Report, TR-0316-22048, Revision 3
ML20141L7872020-05-20020 May 2020
LLC, Submittal of Second Updates to Standard Plant Design Certification Application, Revision 4
ML20141L8162020-05-20020 May 2020
LLC, Submittal of Long-Term Cooling Methodology, TR-0916-51299, Revision 3
ML20141M1142020-05-20020 May 2020
LLC Submittal of NuScale Instrument Setpoint Methodology Technical Report, TR-0616-49121, Revision 3
ML20141L8042020-05-20020 May 2020
LLC Submittal of Technical Specifications Regulatory Conformance and Development, TR-1116-52011, Revision 4
ML20128J3162020-05-18018 May 2020
OEDO-20-00167 - 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:Topical Report
MONTHYEARML20099H0802020-04-0808 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
ML20069A1632020-03-0505 March 2020
ACRS Full Committee Presentation: NuScale Topical Report-Rod Ejection Accident Methodology, PM-0320-69146, Revision 0
ML20069A9692020-03-0505 March 2020
ACRS Full Committee Presentation: NuScale Topical Report, Loss-of-Coolant Accident Evaluation Model, PM-0320-69138, Revision O
ML20069A1542020-03-0404 March 2020
ACRS Full Committee Presentation: NuScale Topical Report- Non-Loss-of-Coolant Accident, PM-0320-69141, Revision 0
ML19352G6972019-12-18018 December 2019
LLC Submittal of Long-Term Cooling Methodology, TR-0916-51299, Revision 2
ML19319C7872019-11-15015 November 2019
LLC Submittal of Technical Specifications Regulatory Conformance and Development, TR-1116-52011, Revision 3
ML19324C8392019-11-13013 November 2019
LLC Submittal of Presentation Materials Entitled ACRS Subcommittee Presentation NuScale Topical Report -Accident Source Term Methodology, PM-1119-67927, Revision 0
ML19211C9012019-07-30030 July 2019
LLC Submittal of NuFuel-HTP2 Fuel and Control Rod Assembly Designs, TR-0816-51127, Revision 2
ML19192A1662019-07-10010 July 2019
Presentation Entitled, Evaluation Methodology for Stability Analysis of the NuScale Power Module.
ML19175A2992019-06-24024 June 2019
LLC Submittal of NuScale Topical Report: Quality Assurance Program Description for the NuScale Power Plant, TR-1010-859-NP, Revision 4
ML19136A3322019-05-16016 May 2019
LLC Submittal of Changes to the Pipe Rupture Hazards Analysis Technical Report, TR-0818-61384, Revision 1
ML19087A2882019-03-28028 March 2019
LLC Updates to Final Safety Analysis Report, Section 1.6,Materials Referenced, Section 3.6, Protection Against Dynamic Effects Associated with Postulated Rupture of Piping, and the Pipe Rupture Hazards Analysis, TR-0818-61384
ML18354B4002018-12-20020 December 2018
LLC - Submittal of Pipe Rupture Hazards Analysis, TR-0818-61384, Revision 1
ML18360A6322018-12-19019 December 2018
LLC - Submittal of Approved Version of Topical Report TR-0116-21012, NuScale Power Critical Heat Flux Corrections, Revision 1
ML18341A3372018-12-0707 December 2018
LLC Submittal of NuScale Comprehensive Vibration Assessment Program Measurement and Inspection Plan Technical Report, TR-0918-60894, Revision 0
ML18338A0312018-12-0404 December 2018
Final Safety Evaluation for NuScale Power, TR-0915-17564-P, Subchannel Analyses Methodology, Revision 2
ML18310A1542018-11-0505 November 2018
LLC Submittal of Fuel Storage Rack Analysis, TR-0816-49833, Revision 1
ML18096B8852018-04-0606 April 2018
LLC Submittal of Technical Report TR-1015-18177, Pressure and Temperature Limits Methodology, Revision 1
2020-04-08
[Table view] |
Text
LO-0519-65598 May 16, 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 Submittal of Changes to the Pipe Rupture Hazards Analysis Technical Report, TR-0818-61384, Revision 1
REFERENCE:
Letter from NuScale Power, LLC to Nuclear Regulatory Commission, NuScale Power, LLC Updates to Final Safety Analysis Report, Section 1.6, Materials Referenced, Section 3.6, Protection Against Dynamic Effects Associated with Postulated Rupture of Piping, and the Pipe Rupture Hazards Analysis, TR-0818-61384, dated March 28, 2019 (ML19087A288)
The referenced letter addressed inconsistencies, identified during the NRC Design Certification Application review, between the Final Safety Analysis Report, Section 1.6, Materials Referenced, Section 3.6, Protection Against Dynamic Effects Associated with Postulated Rupture of Piping, and the Pipe Rupture Hazards Analysis Technical Report, TR-0818-61384, Revision 1.
In a subsequent follow-up communication on May 2, 2019, NRC requested a rewording of the text changes made to the Pipe Rupture Hazards Analysis (PRHA) Technical Report, TR-0818-61384, Sections 4.2.1 and 4.2.3.
The Enclosure to this letter provides a mark-up of the PRHA technical report incorporating the text clarifications as requested, in redline/strikeout format. NuScale will include this change as part of a future revision to the NuScale PRHA technical report.
This letter makes no regulatory commitments or revisions to any existing regulatory commitments.
If you have any questions, please feel free to contact Marty Bryan at (541) 452-7172 or at mbryan@nuscalepower.com.
Sincerely, Zackary W. Rad Director, Regulatory Affairs NuScale Power, LLC Distribution: Samuel Lee, NRC, OWFN-8H12 Gregory Cranston, NRC, OWFN-8H12 Marieliz Vera, NRC, OWFN-8H12
Enclosure:
Changes to the Pipe Rupture Hazards Analysis Technical Report, TR-0818-61384, Revision 1 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com
LO-0519-65598
Enclosure:
Changes to the Pipe Rupture Hazards Analysis Technical Report, TR-0818-61384, Revision 1 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com
Pipe Rupture Hazards Analysis TR-0818-61384-NP Draft Rev. 12 4.2 Blast Effects Blast effects results are based on three-dimensional CFD analysis discussed in Appendix F.
4.2.1 In the Containment Vessel Because only NPS 2 lines are postulated to break, the mass and energy release feeding the blast formation is small. Only the degasification line has the potential for forming a blast, because the other CVCS lines contain subcooled liquid. The magnitude of the blast wave pressures is low, and the maximum force imposed on any component is limited to 6,000 lbf. In addition, the load is of very short duration, (a few milliseconds). Because the blast load is small, its effect on load combinations is inconsequential, and its exclusion does not affect compliance with the ASME allowable limits.and does not need to be included in load combinations.
4.2.2 In the NPM Bay under the Bioshield Not applicable.
4.2.3 In the Reactor Building Breaks are postulated in MSS lines at three locations in a pipe gallery. Only MSS lines have a potential for forming a blast, because the other CVCS lines contain subcooled liquid. The maximum force on any component is less than 10,000 lbf. Although a force of 103,000 lbf on the pool wall was calculated, it is distributed over a surface area with a radius of about 100 inches, yielding a momentary overpressure of less than 15 psig and no damage to the structure. No damage occurs as a result, and the shortness of the loading eliminates the need to consider it in load combinations. In addition, the load is of very short duration (a few milliseconds). Because the blast load is small, its effect on load combinations is inconsequential, and its exclusion does not affect compliance with the ACI 349 allowable limits.
4.3 Pipe Whip Results of pipe whip evaluations are detailed in Appendix C.
4.3.1 In the Containment Vessel Pipe whip for breaks at the RPV and CNV terminals ends has been evaluated. The nozzle/safe end end does not whip. For the piping end, the motion of the pipe is such that no safety-related or essential SSC are impacted. Even if an impact did occur, the SSC are of heavy wall construction so that they neither leak nor crack. There is one exception: the ECCS trip/reset line. If a whipping pipe strikes a trip/reset line, the line is severed, causing it to vent. This has the same effect as opening the trip valve and allows the ECCS main valve to open once the IAB clears. As the response to the HELB is ECCS initiation, the severance of a trip/reset line has no effect on response to the event.
© Copyright 2018 2019 by NuScale Power, LLC 86