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MONTHYEARML21287A4512021-10-15015 October 2021 Email for NuScale Topical Report Quality Assurance Program Description Topical Report -A Version Verification ML21154A1322021-05-26026 May 2021 Final Safety Evaluation Transmittal Email ML21053A2662021-02-22022 February 2021 SMR DC Docs - FW: NuScale EPZ Review Path Forward ML20203M1872020-07-14014 July 2020 Control Room Staffing Topical Report - NRC Staff'S Documentation of the Results of the Completeness Review ML20190A2352020-07-0808 July 2020 SMR DC Docs - Approved Version of NuScale Topical Report, Rod Ejection Accident Methodology, TR-0716-50350, Revision 1 ML20141L6102020-05-20020 May 2020 SMR DC Docs - NuScale Topical Report - Approved Version of NuScale Applicability of Areva Method for the Evaluation of Fuel Assembly Structural Response to Externally Applied Forces, TR-07116-50351, Revision 1 ML20090A8642020-03-30030 March 2020 SMR DC Docs - NuScale Topical Report - Approved Version of TR-0516-49417, Evaluation Methodology for the Stability of the NuScale Power Module, Revision 1 ML19331A7302019-11-27027 November 2019 SMR DC Docs - FW: Ibr Flow Chart ML19331A7382019-11-27027 November 2019 SMR DC Docs - FW: Chapter 18 (Di IP) Related (non-prop) ML19331A7262019-11-26026 November 2019 SMR DC Docs - (External_Sender) Ibr Flow Chart ML19329E7132019-11-25025 November 2019 SMR DC Docs - FW: Chapter 18 (Di IP) Related ML19323F6982019-11-19019 November 2019 DCA Technical Specification Confirmatory Items - Email Commitments ML19329E7022019-11-0606 November 2019 SMR DC Docs - (External_Sender) Chapter 18 (Di IP) Related ML19309D7862019-10-31031 October 2019 Reconciliation - NRR Response to Acrs' September 20 2019 Letter on NuScale Stability Analysis Topical Report ML19309E0172019-10-31031 October 2019 Reconciliation - NRR Response to Acrs'S September 24 2019 Letter on NuScale External Forces Topical Report ML19309F8612019-10-31031 October 2019 Reconciliation - NRR Response to Acrs'S September 25, 2019 Letter on the Focus Area Review Approach ML19276D2812019-10-0303 October 2019 SMR DC RAI - Request for Additional Information No. 526 Erai No. 9719 ML19235A1092019-08-23023 August 2019 SMR DC RAI - Request for Additional Information No. 525 Erai No. 9705 (19.02) ML19206B0772019-07-25025 July 2019 SMR DC Docs - FW: FW: NuScale Chapter 5 Section 5.2.3 - Changes to Information on Check Valves ML19171A0092019-06-20020 June 2019 SMR DC RAI - Request for Additional Information No. 524 Erai No. 9691 (3.9.4) ML19157A0352019-06-0606 June 2019 SMR DC RAI - Request for Additional Information No. 523 Erai No. 9682 (12.3-12.4, 9.3.2) ML19150A3172019-05-30030 May 2019 SMR DC RAI - Request for Additional Information No. 522 Erai No. 9681 (14) ML19151A0272019-05-30030 May 2019 SMR DC RAI - Request for Additional Information No. 522 Erai No. 9681 (14) ML19099A1272019-04-0909 April 2019 SMR DC Docs - FW: NuScale Comments Chapter 5 SER W/Ois ML19098A2362019-04-0808 April 2019 SMR DC Docs - FW: NuScale Comments Chapter 5 SER W/Ois ML19089A0112019-03-29029 March 2019 SMR DC Docs - NuScale Topical Report - Approved Version of Subchannel Analysis Methodology, TR-0915-17564, Revision 2 ML19081A2722019-03-22022 March 2019 SMR DC RAI - 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NuScaleDCRaisPEm Resource From: Cranston, Gregory Sent: Friday, December 29, 2017 12:00 PM To: RAI@nuscalepower.com Cc: NuScaleDCRaisPEm Resource; Lee, Samuel; Chowdhury, Prosanta; Karas, Rebecca; Schmidt, Jeffrey; Nolan, Ryan; Franovich, Rani
Subject:
Request for Additional Information No. 315 RAI No. 9237 (15.06.03)
Attachments: Request for Additional Information No. 315 (eRAI No. 9237).pdf Attached please find NRC staffs request for additional information concerning review of the NuScale Design Certification Application.
Please submit your technically correct and complete response within 60 days of the date of this RAI to the NRC Document Control Desk.
The NRC Staff recognizes that NuScale has preliminarily identified that the response to the question in this RAI is likely to require greater than 60 days.
If you have any questions, please contact me.
Thank you.
Gregory Cranston, Senior Project Manager Licensing Branch 1 (NuScale)
Division of New Reactor Licensing Office of New Reactors U.S. Nuclear Regulatory Commission 301-415-0546 1
Hearing Identifier: NuScale_SMR_DC_RAI_Public Email Number: 342 Mail Envelope Properties (MWHPR09MB1200B4DF0B1B3E140B911EF490050)
Subject:
Request for Additional Information No. 315 RAI No. 9237 (15.06.03)
Sent Date: 12/29/2017 11:59:54 AM Received Date: 12/29/2017 12:00:05 PM From: Cranston, Gregory Created By: Gregory.Cranston@nrc.gov Recipients:
"NuScaleDCRaisPEm Resource" <NuScaleDCRaisPEm.Resource@nrc.gov>
Tracking Status: None "Lee, Samuel" <Samuel.Lee@nrc.gov>
Tracking Status: None "Chowdhury, Prosanta" <Prosanta.Chowdhury@nrc.gov>
Tracking Status: None "Karas, Rebecca" <Rebecca.Karas@nrc.gov>
Tracking Status: None "Schmidt, Jeffrey" <Jeffrey.Schmidt2@nrc.gov>
Tracking Status: None "Nolan, Ryan" <Ryan.Nolan@nrc.gov>
Tracking Status: None "Franovich, Rani" <Rani.Franovich@nrc.gov>
Tracking Status: None "RAI@nuscalepower.com" <RAI@nuscalepower.com>
Tracking Status: None Post Office: MWHPR09MB1200.namprd09.prod.outlook.com Files Size Date & Time MESSAGE 726 12/29/2017 12:00:05 PM Request for Additional Information No. 315 (eRAI No. 9237).pdf 87386 Options Priority: Standard Return Notification: No Reply Requested: No Sensitivity: Normal Expiration Date:
Recipients Received:
Request for Additional Information No. 315 (eRAI No. 9237)
Issue Date: 12/29/2017 Application Title: NuScale Standard Design Certification 048 Operating Company: NuScale Power, LLC Docket No.52-048 Review Section: 15.06.03 - Radiological Consequences of Steam Generator Tube Failure (PWR) 07/1981 Application Section: 15.6.3 QUESTIONS 15.06.03-3 GDC 1 requires structures, systems, and components important to safety to be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed. In addition, 10 CFR 50.2 defines safety-related as structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) the integrity of the reactor coolant pressure boundary; (2) the capability to shut down the reactor and maintain it in a safe shutdown condition; or (3) the capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to the applicable guideline exposures set forth in § 50.34(a)(1) or § 100.11 of this chapter, as applicable.
FSAR Tier 2, Section 15.6.3, Steam Generator Tube Failure (Thermal Hydraulic), credits the nonsafety-related secondary main steam isolation valve (MSIV) to provide isolation of the faulted steam generator when single failure criteria is applied (i.e. MSIV fails to close). FSAR Tier 2, Section 10.3.3 states:
The nonsafety-related secondary MSIVs downstream of the MSIVs are credited as backup isolation components in the event that an MSIV fails to close. Although not safety-related, the secondary MSIVs are designed to close under postulated worst-case conditions and are included in technical specification surveillance requirements to ensure their reliability and operability. Thus, consistent with the position established in NUREG-0138, Issue Number 1, the secondary MSIVs ensure that the blowdown is limited if a steamline were to break upstream of the MSIV.
NUREG-0138, Issue Number 1, states that GDC 1 permits flexibility in the acceptance level for safety-related equipment. NUREG-0138, Issue Number 1 also fully documents that its position is only applicable to spontaneous failures of secondary system piping not part of the primary system boundary, and where the potential for a release of fission products is significantly lower compared to a breach of the primary system boundary. As a steam generator tube rupture is a breach of the primary system boundary which bypasses multiple fission product barriers, provide additional justification as to why NUREG-0138, Issue Number 1 applies and/or provide a dose analysis which demonstrates the applicable accident dose limits are met with the single failure of the MSIV and the failure of the secondary, non-safety-related MSIV to close.