ML18271A152

LLC Response to NRC Request for Additional Information No. 460 (Erai No. 9481) on the NuScale Design Certification Application

ML18271A152
Person / Time
Site:	NuScale
Issue date:	09/28/2018
From:	Rad Z NuScale
To:	Document Control Desk, Office of New Reactors
References
RAIO-0918-61958
Download: ML18271A152 (4)

Text

RAIO-0918-61958 September 28, 2018 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.

460 (eRAI No. 9481) on the NuScale Design Certification Application

REFERENCE:

U.S. Nuclear Regulatory Commission, "Request for Additional Information No.

460 (eRAI No. 9481)," dated May 02, 2018 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. 9481:

15.06.05-8 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 Paul Infanger at 541-452-7351 or at pinfanger@nuscalepower.com.

Sincerely, Zackary W. Rad Director, Regulatory Affairs NuScale Power, LLC Distribution: Gregory Cranston, NRC, OWFN-8G9A Samuel Lee, NRC, OWFN-8G9A Rani Franovich, NRC, OWFN-8G9A : NuScale Response to NRC Request for Additional Information eRAI No. 9481 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvalis, Oregon 97330, Office: 541.360.0500, Fax: 541.207.3928 www.nuscalepower.com

RAIO-0918-61958 :

NuScale Response to NRC Request for Additional Information eRAI No. 9481 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.: 9481 Date of RAI Issue: 05/02/2018 NRC Question No.: 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 NuScale Nonproprietary

break case did not initialize the upper bound T-hot corresponding to the maximum T-avg required by Table 15.0-6.

NuScale Response:

Section 5.3 and Table 5-6 of the Loss-of-Coolant Accident (LOCA) Evaluation Model (EM) topical report (TR-0516-49422) provide methodology guidance for conservatively biasing the initial plant conditions for the LOCA analysis in Section 15.6.5 of the FSAR. At the time the FSAR analysis was docketed, the LOCA methodology prescribed the use of operating limit of 595 degrees F on maximum RCS hot temperature. This limited the maximum initial RCS average temperature (Tavg) below 555 degrees F. As described in response to RAI Question 15.06.05-17 of eRAI 9475 (Letter RAIO-0918-61890), the LOCA methodology (Reference 1) is being updated to remove the limitation on the maximum operating riser temperature as this temperature is protected by the analytical limit setpoint (i.e., a value at which safety system activation is assumed) of High RCS Hot Temperature. This analytical limit setpoint is defined in Table 15.0-7 of the FSAR (or Table 5-5 of the LOCA EM topical report) at 610 degrees F.

As described in response to eRAI 9475 Question 15.06.05-17, the updated LOCA methodology maximizes the initial RCS average temperature (Tavg) to a maximum value of this parameter in Table 15.0-6 of the FSAR (i.e., 555 degrees F). Furthermore, the RCS flow has been minimized to minimum value of this parameter in Table 15.0-6 of the FSAR (i.e., 535 kg/s). This approach results in maximization of the RCS hot temperature (i.e., riser temperature) and the initial RCS stored energy.

The results presented in the revised pages of the LOCA EM topical report are provided with the response to RAI Question 15.06.05-17 of eRAI 9475. This change in methodology has a negligible impact on the LOCA figures-of-merit. The FSAR LOCA analysis is currently being updated to address this methodology change and these updated results will be incorporated in Revision 3 of the FSAR.

Impact on DCA:

There are no impacts to the DCA as a result of this response.

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