ML24215A188

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LLC, Response to SDAA Audit Question Number A-16.1.1-4
ML24215A188
Person / Time
Site: 05200050
Issue date: 08/02/2024
From:
NuScale
To:
Office of Nuclear Reactor Regulation
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Response to SDAA Audit Question Question Number: A-16.1.1-4 Receipt Date: 02/19/2024 Question:

In Section 1.1, the definition of PASSIVE COOLING method a. says one or more valves are open; for consistency, method b. should say one or more trains are in operation instead of saying one or more trains is in operation.

Response

The definition of PASSIVELY COOLED - PASSIVE COOLING was revised in response to Audit item A-16-9 to replace one or more with one of the two in Items a. and b. Based on those changes it is more appropriate to use is in place of are in Items a. and b.

NuScale revises Items a. and b. in the definition of PASSIVELY COOLED - PASSIVE COOLING in GTS Section 1.1 to read:

a. One of the two reactor vent valves is open and one of the two reactor recirculation valves is open, or
b. One of the two trains of DHRS is in operation, or Markups of the affected changes, as described in the response, are provided below:

NuScale Nonproprietary NuScale Nonproprietary

NuScale [US460]

ii Draft Revision 2 TABLE OF CONTENTS Revision B 3.0 LIMITING CONDITION FOR OPERATION AND SURVEILLANCE REQUIREMENTS (continued)

B 3.6 CONTAINMENT SYSTEMS B 3.6.1 Containment....................................................................................................... 2.0 B 3.6.2 Containment Isolation Valves............................................................................. 2.0 B 3.6.3 Containment Closure.......................................................................................... 2.0 B 3.7 PLANT SYSTEMS B 3.7.1 Main Steam Isolation Valves (MSIVs)................................................................ 2.0 B 3.7.2 Feedwater Isolation............................................................................................ 2.0 B 3.8 REFUELING OPERATIONS B 3.8.1 Nuclear Instrumentation..................................................................................... 2.0 B 3.8.2 Decay Time........................................................................................................ 2.0

Definitions 1.1 NuScale US460 1.1-5 Draft Revision 2 1.1 Definitions PASSIVELY COOLED -

PASSIVE COOLING A module is in PASSIVE COOLING or is being PASSIVELY COOLED when:

a. One of the twoor more reactor vent valves isare open and one of the twoor more reactor recirculation valves isare open, or
b. One of the twoor more trains of DHRS is in operation, or
c. Water level in the containment vessel is > 45 ft.

PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation. These tests are:

a. Described in Chapter 14, Initial Test Program and Inspections, Tests, Analyses, and Acceptance Criteria, of the FSAR;
b. Authorized under the provisions of 10 CFR 50.59; or
c. Otherwise approved by the Nuclear Regulatory Commission.

PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

The PTLR is the unit-specific document that provides the reactor vessel pressure and temperature limits, including heatup and cooldown rates, for the current reactor vessel fluence period. These pressure and temperature limits shall be determined for each fluence period in accordance with Specification 5.6.4.

RATED THERMAL POWER (RTP)

RTP shall be a total reactor core heat transfer rate to the reactor coolant of 250 MWt.

LCO Applicability 3.0 NuScale US460 3.0-1 Draft Revision 2 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY LCO 3.0.1 LCOs shall be met during the MODES or other specified conditions in the Applicability, except as provided in LCO 3.0.2, [and] LCO 3.0.7[,

and LCO 3.0.8].

LCO 3.0.2 Upon discovery of a failure to meet an LCO, the Required Actions of the associated Conditions shall be met, except as provided in LCO 3.0.5 and LCO 3.0.6.

If the LCO is met or is no longer applicable prior to expiration of the specified Completion Time(s), completion of the Required Action(s) is not required, unless otherwise stated.

LCO 3.0.3 When an LCO is not met and the associated ACTIONS are not met, an associated ACTION is not provided, or if directed by the associated ACTIONS, the unit shall be placed in a MODE or other specified condition in which the LCO is not applicable. Action shall be initiated within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to place the unit, as applicable, in:action shall be initiated to place the unit in a MODE or other specified condition in which the LCO is not applicable within:

a. MODE 2 within 7 hour8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />s1 hour; or
b. MODE 3 and PASSIVELY COOLED within 37 hours4.282407e-4 days <br />0.0103 hours <br />6.117725e-5 weeks <br />1.40785e-5 months <br />.24 hours if entry into LCO 3.0.3 is unplanned and risk is assessed and managed.

At the end of the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period:

a. Be in MODE 2 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; and
b. Be in MODE 3 and PASSIVELY COOLED within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

Exceptions to this Specification are stated in the individual Specifications.

WhereIf corrective measures are completed that permit operation in accordance with the LCO or ACTIONS, then completion of the actions required by LCO 3.0.3 is not required.

LCO 3.0.3 is only applicable in MODES 1 and 2, and in MODE 3 when not PASSIVELY COOLED. and 2, and in MODE 3 when not PASSIVELY COOLED.

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;

NuScale [US460]

ii Draft Revision 2 TABLE OF CONTENTS Revision 3.0 LIMITING CONDITION FOR OPERATION AND SURVEILLANCE REQUIREMENTS (continued) 3.5 PASSIVE CORE COOLING SYSTEMS (PCCS) 3.5.1 Emergency Core Cooling System (ECCS)......................................................... 2.0 3.5.2 Decay Heat Removal System (DHRS)............................................................... 2.0 3.5.3 Ultimate Heat Sink.............................................................................................. 2.0 3.5.4 Emergency Core Cooling System Supplemental Boron (ESB).......................... 2.0 3.6 CONTAINMENT SYSTEMS 3.6.1 Containment....................................................................................................... 2.0 3.6.2 Containment Isolation Valves............................................................................. 2.0 3.6.3 Containment Closure.......................................................................................... 2.0 3.7 PLANT SYSTEMS 3.7.1 Main Steam Isolation Valves (MSIVs)................................................................ 2.0 3.7.2 Feedwater Isolation............................................................................................ 2.0 3.8 REFUELING OPERATIONS 3.8.1 Nuclear Instrumentation..................................................................................... 2.0 3.8.2 Decay Time........................................................................................................ 2.0 4.0 DESIGN FEATURES 4.1 Site Location............................................................................................................. 2.0 4.2 Reactor Core............................................................................................................ 2.0 4.3 Fuel Storage............................................................................................................. 2.0 5.0 ADMINISTRATIVE CONTROLS 5.1 Responsibility........................................................................................................... 2.0 5.2 Organization............................................................................................................. 2.0 5.3 Facility Staff Qualifications....................................................................................... 2.0 5.4 Procedures............................................................................................................... 2.0 5.5 Programs and Manuals............................................................................................ 2.0 5.6 Reporting Requirements.......................................................................................... 2.0 5.7 High Radiation Area................................................................................................. 2.0

LCO Applicability B 3.0 NuScale US460 B 3.0-7 Draft Revision 2 BASES LCO 3.0.4 (continued)

Therefore, in such cases, entry into a MODE or other specified condition in the Applicability may be made and the Required Actions followed after entry into the Applicability.

For example, LCO 3.0.4.a may be used when the Required Action to be entered states that an inoperable instrument channel must be placed in the trip condition within the Completion Time. Transition into a MODE or other specified condition in the Applicability may be made in accordance with LCO 3.0.4 and the channel is subsequently placed in the tripped condition within the Completion Time, which begins when the Applicability is entered. If the instrument channel cannot be placed in the tripped condition and the subsequent default ACTION ("Required Action and associated Completion Time not met") allows the OPERABLE train to be placed in operation, use of LCO 3.0.4.a is acceptable because the subsequent ACTIONS to be entered following entry into the MODE include ACTIONS (place the OPERABLE train in operation) that permit safe unit operation for an unlimited period of time in the MODE or other specified condition to be entered.

LCO 3.0.4.b allows entry into a MODE or other specified condition in the Applicability with the LCO not met after performance of a risk assessment addressing inoperable systems and components, consideration of the results, determination of the acceptability of entering the MODE or other specified condition in the Applicability, and establishment of risk management actions, if appropriate.

The risk assessment may use quantitative, qualitative, or blended approaches, and the risk assessment will be conducted using the plant program, procedures, and criteria in place to implement 10 CFR 50.65(a)(4), which requires that risk impacts of maintenance activities to be assessed and managed. The risk assessment, for the purposes of LCO 3.0.4.b, must take into account all inoperable Technical Specification equipment regardless of whether the equipment is included in the normal 10 CFR 50.65(a)(4) risk assessment scope.

The risk assessments will be conducted using the procedures and guidance endorsed by Regulatory Guide 1.160, Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, Revision 4.

Regulatory Guide 1.160 endorses the guidance in Section 11 of NUMARC 93-01, Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants. These documents address general guidance for conduct of the risk assessment, quantitative and qualitative guidelines for establishing risk management actions, and example risk management actions. These include actions to plan and conduct other activities in a manner that controls overall risk, increased risk awareness by shift and management personnel, actions to reduce

SR Applicability B 3.0 NuScale US460 B 3.0-20 Draft Revision 2 BASES SR 3.0.3 (continued) of the SR included the relay contact; the adjacent, physically connected relay contacts were tested during the SR performance; the subject relay contact has been tested by another SR; or historical operation of the subject relay contact has been successful. It is not sufficient to infer the behavior of the associated equipment from the performance of similar equipment. The rigor of determining whether there is a reasonable expectation a Surveillance will be met when performed should increase based on the length of time since the last performance of the Surveillance. If the Surveillance has been performed recently, a review of the Surveillance history and equipment performance may be sufficient to support a reasonable expectation that the Surveillance will be met when performed. For Surveillances that have not been performed for a long period or that have never been performed, a rigorous evaluation based on objective evidence should provide a high degree of confidence that the equipment is OPERABLE. The evaluation should be documented in sufficient detail to allow a knowledgeable individual to understand the basis for the determination.

Failure to comply with specified Frequencies for SRs is expected to be an infrequent occurrence. Use of the delay period established by SR 3.0.3 is a flexibility which is not intended to be used repeatedly to extend Surveillance intervals. While up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or the limit of the specified Frequency is provided to perform the missed Surveillance, it is expected that the Surveillance will be performed at the first reasonable opportunity. The determination of the first reasonable opportunity should include consideration of the impact on plant risk (from delaying the Surveillance as well as any plant configuration changes required or shutting the unit down to perform the Surveillance) and impact on any analysis assumptions, in addition to unit conditions, planning, availability of personnel, and the time required to perform the Surveillance. This risk impact should be managed through the program in place to implement 10 CFR 50.65(a)(4) and its implementation guidance, NRC Regulatory Guide 1.160, "Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, Revision 4." This Regulatory Guide addresses consideration of temporary and aggregate risk impacts, determination of risk management action thresholds, and risk management action up to and including unit shutdown.

The missed Surveillance should be treated as an emergent condition as discussed in the Regulatory Guide. The risk evaluation may use quantitative, qualitative, or blended methods. The degree of depth and rigor of the evaluation should be commensurate with the importance of the component. Missed Surveillances for important components should be analyzed quantitatively. If the results of the risk evaluation determine

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