ML19087A329

LLC Supplemental Response to NRC Request for Additional Information No. 506 (Erai No. 9614) on the NuScale Design Certification Application

ML19087A329
Person / Time
Site:	NuScale
Issue date:	03/28/2019
From:	Rad Z NuScale
To:	Document Control Desk, Office of New Reactors
References
RAIO-0319-64996
Download: ML19087A329 (5)

Text

RAIO-0319-64996 March 28, 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 Supplemental Response to NRC Request for Additional Information No. 506 (eRAI No. 9614) on the NuScale Design Certification Application

REFERENCES:

1. U.S. Nuclear Regulatory Commission, "Request for Additional Information No. 506 (eRAI No. 9614)," dated October 16, 2018

2. NuScale Power, LLC Response to NRC "Request for Additional Information No. 506 (eRAI No.9614)," dated December 12, 2018 The purpose of this letter is to provide the NuScale Power, LLC (NuScale) supplemental response to the referenced NRC Request for Additional Information (RAI).

The Enclosure to this letter contains NuScale's supplemental response to the following RAI Question from NRC eRAI No. 9614:

16-55 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 Carrie Fosaaen at 541-452-7126 or at cfosaaen@nuscalepower.com.

Sincerely, Zackary W. Rad Director, Regulatory Affairs NuScale Power, LLC Distribution: Gregory Cranston, NRC, OWFN-8H12 Samuel Lee, NRC, OWFN-8H12 Getachew Tesfaye, NRC, OWFN-8H12 : NuScale Supplemental Response to NRC Request for Additional Information eRAI No. 9614 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvalis, Oregon 97330, Office: 541.360.0500, Fax: 541.207.3928 www.nuscalepower.com

RAIO-0319-64996 :

NuScale Supplemental Response to NRC Request for Additional Information eRAI No. 9614 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.: 9614 Date of RAI Issue: 10/16/2018 NRC Question No.: 16-55 Paragraph (a)(11) of 10 CFR 52.47 and paragraph (a)(30) of 10 CFR 52.79 state that a design certification (DC) applicant and a combined license (COL) applicant, respectively, are to propose technical specifications (TS) prepared in accordance with 10 CFR 50.36 and 50.36a.

10 CFR 50.36 sets forth requirements for TS to be included as part of the operating license for a nuclear power facility.

Regarding Surveillances unique to the NuScale design, SR 3.6.2.5 is provided in lieu of conducting an integrated containment leak rate test, which is described and explained in FSAR Section 6.2.6.1; containment penetration leakage rate testing is described in FSAR Section 6.2.6.2; CIV leakage rate testing is described in FSAR Section 6.2.6.3. This Surveillance states:

SR 3.6.2.5 Verify the combined leakage rate for all containment bypass leakage paths is 0.6 La when pressurized to 951 psia.

It is unclear to the staff (1) why this surveillance statement does not identify the pressure criterion of 951 psia as the calculated peak containment internal pressure (Pa); and (2) how this SR relates to SR 3.6.1.1 ("Perform required visual examinations and leakage rate testing in accordance with the Containment Leakage Rate Testing Program."). The applicant is requested to revise SR 3.6.2.5 and its Bases by identifying 951 psia as Pa; and if necessary, by updating this pressure to the most up to date value. The applicant is also requested to explain how SR 3.6.2.5 is related to SR 3.6.1.1, and incorporate this explanation in the SRs section of Subsections B 3.6.1 and B 3.6.2.

NuScale Nonproprietary

NuScale Response:

The peak containment internal pressure (Pa) in the technical specification bases for LCO 3.6.1 was revised to align with the applicable accident analysis pressure of 986 psia.

Impact on DCA:

The Technical Specifications have been been revised as described in the response above and as shown in the markup provided in this response.

NuScale Nonproprietary

Containment B 3.6.1 BASES BACKGROUND (continued)

b. De-activated automatic valves secured in their closed positions, except as provided in LCO 3.6.2, Containment Isolation Valves; and

c. The sealing mechanism associated with each containment penetration (e.g. welds, flanges, or o-rings) is OPERABLE (i.e.,

OPERABLE such that the containment leakage limits are met).

APPLICABLE The safety design basis for the containment is that the containment SAFETY must withstand the pressures and temperatures of the limiting Design ANALYSES Basis Accident (DBA) without exceeding the design leakage rates.

The DBAs that result in a challenge to containment OPERABILITY from high pressures and temperatures are a loss of coolant accident (LOCA), a steam line break, and a rod ejection accident (REA) (Ref. 2). In addition, release of significant fission product radioactivity within containment can occur from a LOCA or REA. The DBA analyses assume that the containment is OPERABLE such that, for the DBAs involving release of fission product radioactivity, release to the environment is controlled by the rate of containment leakage. The containment is designed with an allowable leakage rate of 0.20% of containment air weight after a DBA per day (Ref. 3). This leakage rate, used in the evaluation of offsite doses resulting from accidents, is defined in 10 CFR 50, Appendix J (Ref. 1), as La: the maximum allowable containment leakage rate at the calculated peak containment internal pressure 951986 psia (Pa) resulting from the limiting DBA. The allowable leakage rate represented by La forms the basis for the acceptance criteria imposed on containment leakage rate testing. La is assumed to be 0.20% per day in the safety analysis.

Satisfactory leakage rate test results are a requirement for the establishment of containment OPERABILITY.

The containment satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO The containment is designed to maintain leakage integrity < 1.0 La.

Leakage integrity is assured by performing local leak rate testing (LLRT) and containment inservice inspection. Total LLRT leakage is maintained

< 0.60 La in accordance with 10 CFR 50, Appendix J (Ref. 1). Satisfactory LLRT and ISI examination are required for containment OPERABILITY.

NuScale B 3.6.1-2 Draft Revision 3.0