ML19203A315

LLC Supplemental Response to NRC Request for Additional Information No. 325 (Erai No. 9268) on the NuScale Design Certification Application

ML19203A315
Person / Time
Site:	NuScale
Issue date:	07/22/2019
From:	Rad Z NuScale
To:	Document Control Desk, Office of New Reactors
References
RAIO-0719-66377
Download: ML19203A315 (5)

Text

RAIO-0719-66377 July 22, 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. 325 (eRAI No. 9268) on the NuScale Design Certification Application

REFERENCES:

1. U.S. Nuclear Regulatory Commission, "Request for Additional Information No. 325 (eRAI No. 9268)," dated January 08, 2018

2. NuScale Power, LLC Response to NRC "Request for Additional Information No. 325 (eRAI No.9268)," dated April 24, 2018

3. NuScale Power, LLC Supplemental Response to "NRC Request for Additional Information No. 325 (eRAI No. 9268)" dated June 17, 2019 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. 9268:

12.02-11 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. 9268 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvalis, Oregon 97330, Office: 541.360.0500, Fax: 541.207.3928 www.nuscalepower.com

RAIO-0719-66377 :

NuScale Supplemental Response to NRC Request for Additional Information eRAI No. 9268 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.: 9268 Date of RAI Issue: 01/08/2018 NRC Question No.: 12.02-11 Regulatory Basis 10 CFR 52.47(a)(5) requires applicants to identify the kinds and quantities of radioactive materials expected to be produced in the operation and the means for controlling and limiting radiation exposures within the limits of 10 CFR Part 20.

Appendix A to Part 50General Design Criteria for Nuclear Power Plants, Criterion 61Fuel storage and handling and radioactivity control, requires systems which may contain radioactivity to be designed with suitable shielding for radiation protection and with appropriate containment, confinement, and filtering systems.

10 CFR 20.1101(b) and 10 CFR 20.1003, require the use of engineering controls to maintain exposures to radiation as far below the dose limits in 10 CFR Part 20 as is practical. NuScale DSRS section 12.2 Radiation Source, regarding the identification of isotopes and the methods, models and assumptions used to determine dose rates. NuScale DSRS section 12.3 Radiation Protection Design Feature, states in the specific acceptance criteria that areas inside the plant structures should be subdivided into radiation zones, with maximum design dose rate zones and the criteria used in selecting maximum dose rates identified.

10 CFR 52.47(a)(8) requires that the final safety analysis report provide the information necessary to demonstrate compliance with any technically relevant portions of the Three Mile Island requirements set forth in 10 CFR 50.34(f), except paragraphs (f)(1)(xii), (f)(2)(ix), and (f)

(3)(v).

10 CFR 50.34(f)(2)(vii) requires that applicants perform radiation and shielding design reviews of spaces around systems that may, as a result of an accident, contain accident source term NuScale Nonproprietary

radioactive materials, and design as necessary to permit adequate access to important areas and to protect safety equipment from the radiation environment.

10 CFR 50.34(f)(2)(viii) requires that applicants provide a capability to promptly obtain and analyze samples from the reactor coolant system and containment that may contain accident source term radioactive materials without radiation exposures to any individual exceeding 5 rems to the whole body or 50 rems to the extremities.

10 CFR 50.49 and 10 CFR Part 50, Appendix A, Criterion 4 require that certain components important to safety be designed to withstand environmental conditions, including the effects of radiation, associated with design basis events, including normal operation, anticipated operational occurrences, and design basis accidents.

Background

DCD Tier 2 Revision 0 Section 12.2.1.13 Post-Accident Sources, discusses post-accident sources and points to several tables in DCD Section 12.2 for additional information. Table 12.2-28: Post-Accident Source Term Input Assumptions, identifies the assumptions used to derive the post-accident sources of radiation. DCD Table 12.2-29: Post-Accident Core Inventory Release Fractions, provides a listing of the radionuclide groups and the associated release fractions. DCD Table 12.2-30: Post-Accident Containment Aerosol Removal Rates, describes how some radionuclides are removed from the contain atmosphere following an accident. DCD Table 12.2-31: Post-Accident Integrated Energy Deposition and Integrated Dose, provides the Integrated MeV energy deposition, and the integrated dose at specific time intervals during post-accident conditions.

Key Issue:

Because DCD Tier 2 Revision 0 Section 12.2 does not contain a listing of the isotopic inventory (i.e., isotope identification and concentration) during post-accident conditions, the staff is unable to determine how the assumptions listed in Tables 12.2-28 and 12.2-29 have been applied. The post-accident isotopic concentrations in the containment (CNV) air volume liquid are used to determine the post-accident radiation levels in a variety of areas, including but not limited to; areas above the reactor building (RXB) pool due to shine from the air volume in the CNV, the area above the CNV but below the bioshield, areas adjacent to the bioshield subject to radiation penetrating shielding or streaming through opening. The post-accident isotopic concentrations in the reactor coolant system (RCS) liquid are used to determine the post-accident radiation levels in a variety of areas, including but not limited to; in areas with pipes containing RCS NuScale Nonproprietary

liquids (e.g., chemical and volume control system (CVCS), Plant Sample System (PSS) and liquid radioactive waste system (LWRS).

The staff uses the calculated radiation levels to compare design features described in the DCD to the acceptance criteria in DSRS 12.2, 12.3 and 3.11.

Question Please provide, as a revision in the DCD (Section 12.2) a listing of radionuclide concentrations in the CNV air volume for post- accident conditions, and provide in the DCD a listing of radionuclide concentrations in the RCS liquid volume for post-accident conditions.

Or, Provide the specific alternative approaches used and the associated justification.

NuScale Response:

During a face-to-face AST Audit meeting on June 18, 2019, at the NuScale office in Rockville, MD, NuScale agreed to clarify a sentence provided in the supplemental response (Letter #

RAIO-0619-65974 submitted on June 17, 2019; ML19168A244) to facilitate the staff's understanding of the methods used to develop NuScale's accident source term under the bioshield. To clarify, the sentence that reads, "...largest design basis accident radionuclide release into the bioshield envelope..." means that the design basis accident that leads to the largest primary coolant activity release into the bioshield envelope is modeled, which is currently a small line (CVCS) carrying primary coolant outside of containment. This amount of primary coolant released into the bioshield envelope is released instantaneously and homogeneously.

The remainder of the primary coolant source term is assumed to be in the upper CNV as a vapor, from where it will shine and be available to leak out into the bioshield envelope at the CNV design basis leak rate.

Impact on DCA:

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

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