ML24215A202

LLC, Response to SDAA Audit Question Number A-16.5.6.4-1

ML24215A202
Person / Time
Site:	05200050
Issue date:	08/02/2024
From:	NuScale
To:	Office of Nuclear Reactor Regulation
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Download: ML24215A202 (1)

Text

Response to SDAA Audit Question Question Number: A-16.5.6.4-1 Receipt Date: 09/18/2023 Question:

US460 SDAA Specification 5.6.4.c uses the phrase fluency period (also used in US600 Specification 5.6.4.c), while the Westinghouse STS 5.6.4 uses fluence period, which is used by both GTS and STS in the second sentence of the PTLR definition in Section 1.1. Recommend changing the generic TS to conform to the W-STS and the PTLR definition phrasing of fluence period. (Looks like Word auto correct feature may have introduced this variation.)

Response

NuScale revises SDAA Technical Specification Section 5.6.4.c to read:

The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.

Markups of the affected changes, as described in the response, are provided below:

NuScale Nonproprietary NuScale Nonproprietary

Reporting Requirements 5.6 NuScale US460 5.6-5 Draft Revision 2 5.6 Reporting Requirements 5.6.4 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) a.

RCS pressure and temperature limits for heat up, cooldown, low temperature operation, criticality, and hydrostatic testing as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

3.3.1, Module Protection System (MPS) Instrumentation; 3.3.3, Engineered Safety Features Actuation System (ESFAS)

Logic and Actuation; 3.3.4, Manual Actuation Functions; 3.4.3, RCS Pressure and Temperature (P/T) Limits; and 3.4.4, Reactor Safety Valves (RSVs).

b.

The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the following document:

TR-130877-P, "Pressure and Temperature Limits Methodology,"

[Revision 2, December 2022.]

c.

The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluencey period and for any revision or supplement thereto.

5.6.5 Steam Generator Tube Inspection Report A report shall be submitted within 180 days after the initial entry into MODE 3 following completion of an inspection performed in accordance with the Specification 5.5.4, "Steam Generator (SG) Program." The report shall include:

a.

The scope of inspections performed on each SG; b.

The nondestructive examination techniques utilized for tubes with increased degradation susceptibility; c.

For each degradation mechanism found:

1.

The nondestructive examination techniques utilized;