ML18122A379

SMR DC RAI - Request for Additional Information No. 460 Erai No. 9481 (15.06.05)

ML18122A379
Person / Time
Site:	NuScale
Issue date:	05/02/2018
From:	NRC
To:	NRC/NRO/DNRL/LB1
References
Download: ML18122A379 (3)

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NuScaleDCRaisPEm Resource From: Chowdhury, Prosanta Sent: Wednesday, May 2, 2018 4:05 PM To: Request for Additional Information Cc: Lee, Samuel; Cranston, Gregory; Franovich, Rani; Karas, Rebecca; Lu, Shanlai; NuScaleDCRaisPEm Resource

Subject:

Request for Additional Information No. 460 eRAI No. 9481 (15.06.05)

Attachments: Request for Additional Information No. 460 (eRAI No. 9481).pdf Attached please find NRC staffs request for additional information (RAI) 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.

If you have any questions, please contact me.

Thank you.

Prosanta Chowdhury, Project Manager Licensing Branch 1 (NuScale)

Division of New Reactor Licensing Office of New Reactors U.S. Nuclear Regulatory Commission 301-415-1647 1

Hearing Identifier: NuScale_SMR_DC_RAI_Public Email Number: 491 Mail Envelope Properties (BN7PR09MB2609228A69203E462B2B89BE9E800)

Subject:

Request for Additional Information No. 460 eRAI No. 9481 (15.06.05)

Sent Date: 5/2/2018 4:04:50 PM Received Date: 5/2/2018 4:04:57 PM From: Chowdhury, Prosanta Created By: Prosanta.Chowdhury@nrc.gov Recipients:

"Lee, Samuel" <Samuel.Lee@nrc.gov>

Tracking Status: None "Cranston, Gregory" <Gregory.Cranston@nrc.gov>

Tracking Status: None "Franovich, Rani" <Rani.Franovich@nrc.gov>

Tracking Status: None "Karas, Rebecca" <Rebecca.Karas@nrc.gov>

Tracking Status: None "Lu, Shanlai" <Shanlai.Lu@nrc.gov>

Tracking Status: None "NuScaleDCRaisPEm Resource" <NuScaleDCRaisPEm.Resource@nrc.gov>

Tracking Status: None "Request for Additional Information" <RAI@nuscalepower.com>

Tracking Status: None Post Office: BN7PR09MB2609.namprd09.prod.outlook.com Files Size Date & Time MESSAGE 556 5/2/2018 4:04:57 PM Request for Additional Information No. 460 (eRAI No. 9481).pdf 13067 Options Priority: Standard Return Notification: No Reply Requested: No Sensitivity: Normal Expiration Date:

Recipients Received:

Request for Additional Information No. 460 (eRAI No. 9481)

Issue Date: 05/02/2018 Application

Title:

NuScale Standard Design Certification 048 Operating Company: NuScale Power, LLC Docket No.52-048 Review Section: 15.06.05 - Loss of Coolant Accidents Resulting From Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary Application Section: 15.6.5 QUESTIONS 15.06.05-8 Title 10, Part 50, Section 50.46 "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors" specifies the loss of coolant accident (LOCA) evaluation model includes one or more computer programs and all other information necessary for application of the calculational framework to a specific LOCA, such as mathematical models used, assumptions included in the programs, procedure for treating the program input and output information, specification of those portions of analysis not included in computer programs, values of parameters, and all other information necessary to specify the calculational procedure. Regulatory Guide 1.203 describes a process that the staff of the U.S. Nuclear Regulatory Commission (NRC) considers acceptable for use in developing and assessing evaluation models (EMs) that may be used to analyze transient and accident behavior that is within the design basis of a nuclear power plant. NuScale Design-Specific Review Standard Section 15.6.5, "Loss-Of-Coolant Accidents Resulting From Spectrum Of Postulated Piping Breaks Within The Reactor Coolant Pressure Boundary," directs the staff to evaluate whether the appropriate break locations, break sizes, and initial conditions were selected in a manner that conservatively predicts the consequences of the LOCA for evaluating emergency core cooling system performance.

Final Safety Analysis Report Tier 2, Section 15.6.5.3.2, "Input Parameters and Initial Conditions," states that, "[reactor coolant system (RCS)] average temperature is initialized to yield a maximum riser operation temperature of 595 oF." The 595 oF value of the riser operation temperature corresponds to a T-avg of 547.5 oF based on the chemical and volume control system (CVCS) line break LOCA NRELAP5 calculation. However, Table 15.0-6 lists all the module initial condition ranges for design basis event evaluation, with the T-avg value specified as 545 oF +/- 10 oF uncertainty. Therefore, the conservative and maximum initial condition T-avg should be 555 oF with a corresponding T-hot of 605 oF. Explain why the design basis CVCS line break case did not initialize the upper bound T-hot corresponding to the maximum T-avg required by Table 15.0-6.