ML18355A921

LLC Response to NRC Request for Additional Information No. 459 (Erai No. 9523) on the NuScale Design Certification Application

ML18355A921
Person / Time
Site:	NuScale
Issue date:	12/21/2018
From:	Rad Z NuScale
To:	Document Control Desk, Office of New Reactors
References
RAIO-1218-63922
Download: ML18355A921 (5)

Text

RAIO-1218-63922 December 21, 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.

459 (eRAI No. 9523) on the NuScale Design Certification Application

REFERENCE:

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

459 (eRAI No. 9523)," 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. 9523:

15-14 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. 9523 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvalis, Oregon 97330, Office: 541.360.0500, Fax: 541.207.3928 www.nuscalepower.com

RAIO-1218-63922 :

NuScale Response to NRC Request for Additional Information eRAI No. 9523 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.: 9523 Date of RAI Issue: 05/02/2018 NRC Question No.: 15-14 10 CFR Part 50 Appendix A, General Design Criterion (GDC) 34, "Residual heat removal," and NuScale's Primary Design Criterion (PDC) 34, in FSAR Section 3.1.4.5, state:

A system to remove residual heat shall be provided. The system safety function shall be to transfer fission product decay heat and other residual heat from the reactor core at a rate such that specified acceptable fuel design limits and the design conditions of the reactor coolant pressure boundary are not exceeded.

The long-term cooling technical report, TR-0916-51299, supports Final Safety Analysis Report (FSAR) Section 15.0.5, "Long Term Decay and Residual Heat Removal," when the emergency core cooling system (ECCS) is used for long term decay heat removal following either a non-loss of coolant accident (LOCA) or LOCA event up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The primary acceptance criteria for the analysis are 1) Collapsed liquid level is maintained above the active fuel and 2) fuel cladding temperature is maintained at an acceptable level such that the specified acceptable fuel design limits (SAFDLs) are preserved.

The staff notes that in Section 5.3 of TR-0916-51299, "Demonstration of Results," that no non-LOCA cases for minimum cooldown appear to have been evaluated as was the case for LOCAs. The staff notes that reactor coolant system (RCS) conditions following a non-LOCA transient such as turbine trip or loss of normal feedwater would yield different RCS conditions than a LOCA upon reaching the inadvertent actuation block (IAB) release threshold. Therefore, the staff is requesting additional information or justification as to why a non-LOCA minimum cooldown case was not evaluated. If one has been evaluated, the staff is requesting the applicant update TR-0916-51299 as appropriate. If non-LOCA minimum cooldown case(s) has been evaluated or need to be evaluated, the applicant should address the effect of IAB setpoint uncertainty on the resulting long term cooling acceptance criteria.

NuScale Nonproprietary

NuScale Response:

Section 5 of TR-0916-51299 was revised to include discussion of non-LOCA events, including the loss of normal feedwater (LOFW), as indicated in the response to RAI 9516, question 15-26, submitted in NuScale letter RAIO-1218-63931, dated December 21, 2018. As stated in Table 5-1 of TR-0916-51299, non-LOCA events are always evaluated with DHRS enabled. Under minimum cooldown conditions, the LOCA events bound the non-LOCA events in terms of maximum core temperatures since DHRS operation was conservatively disabled. Relative to other non-LOCA events, the steam generator tube failure (SGTF) event was bounding for maximum temperature since DHRS performance was degraded. Long term results for the LOFW event trended closely with results for the DHRS cooldown event where the reactor was tripped at initiation. Long term, the onset of recirculation flow and therefore the start of LTC occurs after CNV and RPV conditions are nearly equalized in terms of liquid inventory and pressure. At this point, the particular initiating event is not very influential to the overall LTC response, except for some variance in decay heat due to transient progression and event dependent differences in RCS inventory available for ECCS cooling. Table 1 provides long term module conditions for a variety of minimum cooldown events to support the above discussion.

The DHRS cooldown and LOFW cases with ECCS actuation after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> show similar module conditions at the end of the transient simulation.

Table 1. Selection of minimum cooldown scenario results Time of Riser Maximum Core Inlet Reported Collapsed Clad Pressure at Case Description Temperature Results Level above Temperature RVV (psia)

(°F)

(hour) TAF (ft) (°F) 100% Injection Line Break 12.5 8.93 292.8 303.2 62.0 DHRS cooldown (loss of AC and DC power at 12.5 9.67 252.3 267.2 31.8 time zero, ECCS actuation on IAB)

SGTF (loss of AC and DC power at 12.5 9.30 270.0 276.7 38.6 time zero, ECCS actuation on IAB)

DHRS cooldown (loss of AC power at time 36.0 9.57 249.0 261.9 28.7 zero, ECCS actuation at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />)

LOFW (loss of AC power at reactor 36.0 9.56 247.9 260.6 27.9 trip, ECCS actuation at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />)

NuScale Nonproprietary

The inadvertent actuation block (IAB) setpoint uncertainty does not significantly impact long term module conditions, and any influence on final conditions simply results from changing the total time under ECCS cooling before the transient simulation end time is reached. However, the IAB setpoint does influence the shorter term LTC collapsed level response. Higher decay heat is limiting for the level response in the hours following ECCS actuation. Earlier ECCS actuation will maximize decay heat at the time collapsed level is most challenged. For this reason, and because actuation timing is not significant for long term conditions, all LTC evaluations are performed with the maximum IAB release setpoint (1200 psid).

Impact on DCA:

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

NuScale Nonproprietary