ML24255A315
| ML24255A315 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 10/22/2024 |
| From: | Markley M Plant Licensing Branch II |
| To: | Carr E Dominion Energy South Carolina |
| Miller, GE | |
| References | |
| EPID L-2023-LLR-0068 | |
| Download: ML24255A315 (14) | |
Text
October 22, 2024 Eric S. Carr Senior Vice President and Chief Nuclear Officer Innsbrook Technical Center 5000 Dominion Blvd.
Glen Allen, VA 23060-6711
SUBJECT:
VIRGIL C. SUMMER NUCLEAR STATION, UNIT NO. 1 - RE: RELIEF REQUEST RR-5-V2 REGARDING PRESSURE ISOLATION VALVES (EPID L-2023-LLR-0068)
Dear Eric Carr:
By letter dated December 21, 2023 (Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML23361A104), as supplemented by letter dated April 29, 2024 (ML24121A100), Dominion Energy South Carolina (Dominion, the licensee), submitted a request for the Virgil C. Summer Nuclear Station, Unit No. 1 (Summer), to the U.S. Nuclear Regulatory Commission (NRC or Commission) for a proposed alternative to certain American Society of Mechanical Engineers (ASME) Operations and Maintenance (OM) Code inservice inspection (IST) leakage testing requirements.
Specifically, Dominion requested relief from ASME OM Code Subsection ISTC-3630, Leakage Rate for Other Than Containment Isolation Valves, and Subsection ISTC-3630(a) on the basis that the proposed performance-based alternative provides an acceptable level of quality and safety.
The NRC staff has reviewed the alternative request and concludes, as set forth in the enclosed safety evaluation, that Dominion has adequately addressed all the regulatory requirements set forth in 10 CFR 50.55a(z)(1). Therefore, the NRC authorizes the proposed alternative for the Fifth IST Interval that begins January 1, 2025. All other ASME Code requirements for which relief was not specifically requested and approved remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.
If you have any questions, please contact the Project Manager, Ed Miller at 301-415-2481 or via e-mail at Ed.Miller@nrc.gov.
Sincerely, Michael T. Markley, Chief Plant Licensing Branch 2-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-395
Enclosure:
Safety Evaluation cc: Listserv MICHAEL MARKLEY Digitally signed by MICHAEL MARKLEY Date: 2024.10.22 12:45:52 -04'00'
Enclosure SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION PRESSURE ISOLATION VALVES RR-5-V2 DOMINION ENERGY SOUTH CAROLINA VIRGIL C. SUMMER NUCLEAR STATION, UNIT 1 DOCKET NO. 50-395
1.0 INTRODUCTION
By letter dated December 21, 2023 (Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML23361A104), as supplemented by letter dated April 29, 2024 (ML24121A100), Dominion Energy South Carolina (Dominion, the licensee), submitted a request for the Virgil C. Summer Nuclear Station, Unit No. 1 (Summer), to the U.S. Nuclear Regulatory Commission (NRC or Commission) for a proposed alternative to certain American Society of Mechanical Engineers (ASME) Operations and Maintenance (OM) Code inservice inspection (IST) leakage testing requirements.
Specifically, Dominion proposed a performance-based alternative from ASME OM Code Subsection ISTC-3630, Leakage Rate for Other Than Containment Isolation Valves, and Subsection ISTC-3630(a) on the basis that the proposed alternative provides an acceptable level of quality and safety.
2.0 REGULATORY EVALUATION
The NRC regulations in 10 CFR 50.55a(z), Alternatives to codes and standards requirements, state that alternatives to the requirements of 10 CFR 50.55a(b) through (h) or portions thereof may be used when authorized by the Director, Office of Nuclear Reactor Regulation. A proposed alternative must be submitted and authorized prior to implementation.
The applicant or licensee must demonstrate that:
(1) Acceptable level of quality and safety. The proposed alternative would provide an acceptable level of quality and safety; or (2) Hardship without a compensating increase in quality and safety. Compliance with the specified requirements of this section would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix J, Option B, Performance Based Requirements.
Regulatory Guide (RG) 1.163, Performance-Based Containment Leak-Test Program (ML003740058). Revision 1 to RG 1.163, June 2023.
NUREG/CR-5928, [Inter System Loss of Coolant Accident] ISLOCA Research Program (ML072430731).
3.0 TECHNICAL EVALUATION
ASME OM Code Components Affected The following plant equipment within the scope of this request are:
Table 1 Valve Component ID Component Description Code Class Code Category 1
XVG08701A-RH RH Header A Isolation Valve (IRC) 1 A
2 XVG08701B-RH RH Header B Isolation Valve (IRC) 1 A
3 XVG08702A-RH RH Inlet Header A Isolation Valve 1
A 4
XVG08702B-RH RH Inlet Header B Isolation Valve 1
A 5
XVC08703A-RH RH Header A Bypass Check Valve (IRC) 2 A/C 6
XVC08703B-RH RH Header B Bypass Check Valve (IRC) 2 A/C 7
XVC08948A-SI SI Loop A Outlet Header Check Valve 1
A/C 8
XVC08948B-SI SI Loop B Outlet Header Check Valve 1
A/C 9
XVC08948C-SI SI Loop C Outlet Header Check Valve 1
A/C 10 XVC08956A-SI SI Accum A Disch Header Check Valve 1
A/C 11 XVC08956B-SI SI Accum B Disch Header Check Valve 1
A/C 12 XVC08956C-SI SI Accum C Disch Header Check Valve 1
A/C 13 XVC08973A-SI RCS Loop A Cold Leg Inlet Hdr Check Valve 1
A/C 14 XVC08973B-SI RCS Loop B Cold Leg Inlet Hdr Check Valve 1
A/C
15 XVC08973C-SI RCS Loop C Cold Leg Inlet Hdr Check Valve 1
A/C 16 XVC08974A-SI SI Header A Check Valve (IRC) 2 A/C 17 XVC08974B-SI SI Header B Check Valve (IRC) 2 A/C 18 XVC08988A-SI RHR Supply Header Check Valve 1
A/C 19 XVC08988B-SI RHR Supply Header Check Valve 1
A/C 20 XVC08990A-SI Loop A Low Head Hot Leg Check Valve 1
A/C 21 XVC08990B-SI Loop B Low Head Hot Leg Check Valve 1
A/C 22 XVC08990C-SI Loop C Low Head Hot Leg Check Valve 1
A/C 23 XVC08992A-SI Loop A High Head Hot Leg Check Valve 1
A/C 24 XVC08992B-SI Loop B High Head Hot Leg Check Valve 1
A/C 25 XVC08992C-SI Loop C High Head Hot Leg Check Valve 1
A/C 26 XVC08993A-SI Loop A High Head Hot Leg Hdr Check Valve 1
A/C 27 XVC08993B-SI Loop B High Head Hot Leg Hdr Check Valve 1
A/C 28 XVC08993C-SI Loop C High Head Hot Leg Hdr Check Valve 1
A/C 29 XVC08995A-SI Loop A High Head Cold Leg Check Valve 1
A/C 30 XVC08995B-SI Loop B High Head Cold Leg Check Valve 1
A/C 31 XVC08995C-SI Loop C High Head Cold Leg Check Valve 1
A/C 32 XVC08997A-SI Loop A Low Head Cold Leg Check Valve 1
A/C 33 XVC08997B-SI Loop B Low Head Cold Leg Check Valve 1
A/C 34 XVC08997C-SI Loop C Low Head Cold Leg Check Valve 1
A/C 35 XVC08998A-SI Loop A Low Head Cold Leg Check Valve 1
A/C 36 XVC08998B-SI Loop B Low Head Cold Leg Check Valve 1
A/C 37 XVC08998C-SI Loop C Low Head Cold Leg Check Valve 1
A/C
The licensee stated:
Thirty-seven (37) valves, as shown in Table 1 below, are included in this alternative request (AR). With the exception of valves No. 5 and 6, thirty-five (35) of these valves are Reactor Coolant System (RCS) Pressure Isolation Valves (PIVs) with functions to provide reactor coolant pressure boundary isolation by separating the high-pressure RCS from an attached lower-pressure system, and prevent excessive PIV leakage which could lead to overpressure of the low-pressure piping or components, potentially resulting in a loss of coolant accident (LOCA) outside of containment.
Valves No. 5 and 6 (XVC08703A/B-RH) in Table 1, are not identified as PIVs in Technical Specifications (TS) Table 3.4-1. These valves are Residual Heat Removal (RHR) check valves located in bypass lines around the inner RHR Inlet Isolation motor operated valves (MOVs) (XVG08702A/B-RH, valves No. 3 and 4) and perform an active safety function in the open position to provide overpressure protection due to thermal buildup occurring in the interconnecting piping between the RHR Inlet Isolation MOVs (XVG08701A/B-RH, valves No. 1 and 2, and XVG08702A/B-RH, valves No. 3 and 4).
These two valves (valves No. 5 and 6) perform an active safety function in the closed position by providing high to low pressure boundary isolation.
Applicable ASME Code Edition and Addenda The applicable ASME OM Code is the 2020 Edition.
Applicable ASME OM Code Requirements The IST requirements in the ASME OM Code as incorporated by reference in 10 CFR 50.55a related to Request RR-5-V2 are as follows:
ASME OM Code, Subsection ISTC, paragraph ISTC-3630, Leakage Rate for Other Than Containment Isolation Valves, states, in part, that:
Category A valves with a leakage requirement not based on an Owners 10 CFR 50, Appendix J program, shall be tested to verify their seat leakages within acceptable limits. Valve closure before seat leakage testing shall be by using the valve operator with no additional closing force applied.
ASME OM Code, Subsection ISTC, paragraph ISTC-3630, subparagraph (a), Frequency, states, in part, that:
Tests shall be conducted at least once every 2 yr [years].
Licensees Proposed Alternative for Request RR-5-V2 In its letter dated December 21, 2023, the licensee stated that:
VCSNS proposes to perform testing of the PIVs and check valves XVC08703A/B-RH at intervals ranging from every refueling outage (RFO) to every fourth refueling outage. The specific interval for each valve would be a function of its performance and would be established in a manner consistent with the CIV extended test eligibility process guidance under 10 CFR 50, Appendix J, Option B. Performance-based scheduling of
PIVs and the check valves testing will be controlled in a manner similar to the methods described in NEI 94-01, Revision 3-A. PIV test performances would occur at a nominal frequency ranging from every refueling outage to every fourth refueling outage, subject to acceptable valve performance. Valves that have demonstrated good performance for two consecutive cycles may have their test interval extended up to 75-months, with a permissible extension (for non-routine emergent conditions) of nine months (84 months total).
A conservative control will be established such that if any valve fails the PIV test, the test interval will be reduced consistent with Appendix J, Option B, requirements. Leakage rates less than the leakage limits found in TS and VCSNS procedure STP-215.008, SI and RHR System Valve Leakage Test, shall be considered acceptable. Any PIV leakage test failure would require the associated component to return to the initial test interval of every RFO or two years until acceptable performance is re-established.
=
Reason for Request===
In its letter dated December 21, 2023, the licensee stated that:
ASME OM Code Subsection ISTC-3630, paragraph (a) requires that leakage rate testing for Category A valves with a leakage requirement not based on an owner's 10 CFR 50, Appendix J program be performed at least once every two years. Since PIVs may or may not be containment isolation valves (CIVs), they are not necessarily included in scope for performance-based testing as provided in 10 CFR 50, Appendix J, Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors, Option B, Performance Based-Requirements.
The concept behind the Option B alternative for CIVs is that licensees should be allowed to adopt cost effective methods for complying with regulatory requirements. Additionally, NEI 94-01, Revision 3-A describes the risk-informed basis for the extended test intervals under Option B. The discussion concludes that CIVs which have demonstrated good performance by the successful completion of two consecutive leakage rate tests over two consecutive cycles, licensees may increase their frequencies. NEI 94-01, Revision 3-A also presents the results of a comprehensive risk analysis, including the conclusion that "the risk impact associated with increasing [leak rate] test intervals is negligible (i.e.,
less than 0.1 percent of total risk)."
The valves identified in this request are all in water applications. Testing is performed with water pressurized to slightly below or at the function maximum pressure differential; however, where necessary the observed leakage is adjusted to the function maximum pressure differential value in accordance with ASME OM Code Subsection ISTC-3630, paragraph (b) Differential Test Pressure, item (4). Testing of the PIVs is performed during plant startup following a refueling shutdown. The testing is performed by applying test pressure to the Reactor Coolant System (RCS) side of the disk by using the RCS as the pressure source or the Charging System via the Emergency Core Cooling System (ECCS) test header and the associated flow meters. The purpose of the test is to perform Category A, seat leakage testing of the PIVs. Although the testing of the PIVs includes a limit on allowable PIV leakage rate, the main purpose of this limit is to prevent overpressure failure of the low-pressure portions of connecting systems. The allowable leakage limit provides a standard against which the PIV leakage can be compared to determine if the component is degraded or degrading. Excessive PIV leakage (i.e.,
greater than the allowable leakage limit) could lead to over-pressurization of the low-pressure piping or components, potentially resulting in a loss of coolant accident (LOCA) outside of containment.
The RHR Header Bypass Check Valves (XVC08703A/B-RH) are 0.75-inch check valves located in the bypass lines around the inner RHR Inlet Isolation valves (XVG08702A/BRH). The purpose of these check valves is to open and provide over-pressure relief due to thermal buildup in the piping between the RHR Inlet Isolation valves (XVG08701A/BRH and XVG08702A/BRH) and to close to prevent pressure/flow from bypassing XVG08702A/B-RH. Check valves XVC08703A/B-RH are not identified in Technical Specifications (TS) Table 3.4-1 as PIVs; however, XVC08703A/B-RH are leak tested in parallel with XVG08702A/B-RH using the same VCSNS procedure, STP-215.008, SI and RHR System Valve Leakage Test, and under the same plant conditions.
If leakage through both valves is unacceptable, XVC08703A/B-RH can be isolated so that XVG08702A/B-RH can be tested independently to determine the source of the leakage. XVC08703A/B-RH cannot be tested independently.
This proposed alternative is intended to provide for a performance-based scheduling of PIV tests at VCSNS and align the testing frequency of the two RHR Header Bypass Check Valves with the performance-based testing of the PIVs. The primary reason for requesting this alternative is to eliminate unnecessary thermal cycles in the RCS Cold Leg Safety Injection (SI) piping. A periodic thermal transient was identified in the RCS Cold Leg SI piping after post-refueling heat-up since approximately 1999. These transients coincide with the testing of the RCS PIVs, which causes the inlet check valves (XVC08998A-SI, XVC08998B-SI, and XVC08998C-SI) to open during this portion of testing, allowing cooler Volume Control Tank (VCT) temperature water into the SI piping.
These thermal transients, identified by the plant thermal cycle counting software, are counted against allowable fatigue usage totals for the affected piping system. For the RCS Cold Leg SI lines, the approximate fatigue usage is at 70% of the allowable. As a result of the high cumulative usage factor, additional ultrasonic inspections of the welds and elbows of the RCS Cold Leg SI lines A, B, and C in areas susceptible to thermal stratification were performed during refueling outage RF-21 in April 2014, with acceptable exam results. The proposed extended test intervals would reduce the frequency and, therefore, the impact of injecting ECCS water into the RCS during testing.
An additional reason for requesting this alternative is dose reduction to conform with Nuclear Regulatory Commission (NRC) and industry As Low As Reasonably Achievable (ALARA) radiation dose principles. A review of recent historical data identified that PIV testing results in a total personnel dose of approximately 0.3 Roentgen Equivalent Man (REM) each RFO.
NUREG-0933, Resolution of Generic Safety Issues, Section 3, Issue 105, Interfacing Systems LOCA at LWRs, discussed the need for PIV leak rate testing based primarily on three pre-1985 historical failures of applicable valves industry-wide. These failures all involved human errors in either operations or maintenance. None of these failures involved inservice equipment degradation. The performance of PIV leak rate testing provides assurance of acceptable seat leakage with the valve in a closed condition.
For check valves, functional testing is accomplished in accordance with ASME OM Code Mandatory Appendix 11, Check Valve Condition Monitoring Program. For power-operated valves, full stroke functional testing is accomplished in accordance with the ASME OM Code paragraph ISTC-3521, Category A and Category B Valves. The functional testing of the PIV check valves will be monitored through a Condition Monitoring Plan in accordance with ISTC-5222, Condition -Monitoring Program, and Mandatory Appendix II, Check Valve Condition Monitoring Program. Performance of the separate two-year PIV leak rate testing does not contribute any additional assurance of functional capability but rather provides added assurance of valve integrity, thereby reducing the probability of gross valve failure and consequent intersystem LOCA.
The use of a Condition Monitoring Plan is intended to align the frequency for the closure exercise testing with the pressure isolation valve test. By use of a Condition Monitoring Plan, the check valve closure test, based on performance, would be verified concurrently with the PIV seat leakage test. The frequency of the check valve closure test would then be the same as the PIV seat leakage test since closure performance and seat leakage performance are linked. The PIV seat leakage test would not pass if the valve failed to close. Pursuant to 10 CFR 50.55a, Codes and standards, paragraph 50.55a(z)(1 ), an alternative to the requirement of ASME OM Code Subsection ISTC-3630(a) is requested. The basis of the request is that the proposed alternative would provide an acceptable level of quality and safety.
Licensee Assessment of Request RR-5-V2 In its letter dated December 21, 2023, the licensee stated that:
The primary justification provided by the licensee for this proposed alternative is the historically good performance of the PIVs and the check valves.
Tables 2 and 3 below present historical test data that demonstrates acceptable PIV performance for the Residual Heat Removal (RHR) and Safety Injection (SI) systems.
Individual testing of PIVs is performed for the purposes of identifying leakage and when troubleshooting is required. Group testing of PIVs is performed where such capability exists. Group testing is more conservative, in that, the same limit is applied when testing a single valve or group consisting of multiple valves. The comments sections in the tables below delineate the manner of testing performed.
Table 4 presents historical data that demonstrates acceptable check valve performance for the RHR Header Bypass Check Valves XVC08703A/B-RH. These check valves are tested in parallel with the inner RHR Inlet Isolation valves XVG08702A/B-RH. If leakage through both valves is unacceptable, XVC08703A/B-RH can be isolated so that XVG08702A/B-RH can be tested independently to determine the source of the leakage.
XVC08703A/B-RH cannot be tested independently.
The functional capability of the motor-operated valves (MOVs) included within this alternative request is demonstrated by the full exercise test and stroke time testing (both the open and close directions) performed at a cold shutdown frequency in accordance with ASME OM Code. Additionally, the MOVs are position indication tested at a biennial frequency in accordance with ASME OM Code, paragraph ISTC-3700, and satisfying the requirements of 10 CFR 50.55a(b)(3)(xi). These tests are separate and distinct from the leakage determining function, which PIV testing demonstrates.
For the check valves, the functional capability included within this alternative request is demonstrated by the open and close exercise performed in accordance with ASME OM Code, Appendix II, paragraph ll-4000(b), Optimization of Condition-Monitoring Frequencies. The Condition Monitoring Plans for the check valves currently include open verification testing performed at least every third RFO. Additionally, leak testing is currently performed at least every third RFO and credited with confirming closure of the check valves.
Note that NEI 94-01, Revision 3-A, is not the sole basis for this alternative request, given that NEI 94-01, Revision 3-A, does not address seat leakage testing with water. Instead, NEI 94-01 was cited as an approach similar to the requested alternative method and provides reasonable assurance of continued PIV operational readiness. If the proposed alternative is authorized and the valves exhibit good performance, the PIV test frequency will be controlled similar to the method described in NEI 94-01, Revision 3-A, so that testing of these PIVs and check valves would not be required each refueling outage.
The extension of test frequencies proposed is consistent with the guidance provided in 10 CFR Part 50, Appendix J, Type C leak rate tests as detailed in NEI 94-01, Revision 3-A, paragraph 10.2.3.2, Extended Test Interval, which states:
Test intervals for Type C valves may be increased based upon completion of two consecutive periodic as-found Type C tests where the result of each test is within a licensee's allowable administrative limits.
Elapsed time between the first and last tests in a series of consecutive passing tests used to determine performance shall be 24 months or the nominal test interval (e.g., refueling cycle) for the valve prior to implementing Option B to Appendix J. Intervals for Type C testing may be increased to a specific value in a range of frequencies from 30 months up to a maximum of 75 months. Test intervals for Type C valves should be determined by a licensee in accordance with Section 11.0.
Additional justifications for NRC approval of this proposed alternative are:
Separate functional testing of MOV PIVs, check valve PIVs, and check valves will continue to be conducted per the ASME OM Code.
There is a low likelihood of valve mispositioning during power operations (e.g., alignment and valve position verification procedures, interlocks).
Relief valves in the low pressure (LP) piping might not provide inter-system LOCA (ISLOCA) mitigation for inadvertent PIV mispositioning, but their relief capacity can accommodate conservative PIV seat leakage rates.
Alarms are provided that identify high pressure (HP) to LP leakage. Operators are trained to recognize symptoms of a present ISLOCA and to take appropriate actions.
If this proposed alternative is authorized and the PIVs and the two check valves XVC08703A/B-RH continue to exhibit good performance, the test frequency for these valves could be extended such that testing would not be required each RFO. Instead, testing would be conducted at an interval not to exceed every 75-months, with a permissible extension (for non-routine emergent conditions) of nine months (84 months total).
Based on valve performance history, there is continued assurance of valve operational readiness, as required by ASME OM-2020 Code, paragraph ISTC-3630. Therefore, this proposed alternative to extend the testing frequency will continue to provide assurance of the valves' operational readiness and provides an acceptable level of quality and safety pursuant to 10 CFR 50.55a(z)(1).
NRC Staff Evaluation of Request RR-5-V2 In nuclear power plants, PIVs are defined as two valves in series within the reactor coolant pressure boundary which separate the high-pressure RCS from an attached lower pressure system. Failure of a PIV could result in an overpressurization event which could lead to a system rupture and possible release of fission products to the environment. This type of failure event was analyzed under NUREG/CR-5928, [Inter System Loss of Coolant Accident] ISLOCA Research Program (ML072430731). The purpose of NUREG/CR-5928 was to quantify the risk associated with an ISLOCA event for boiling water reactor (BWR) and pressurized water reactor (PWR) nuclear power plant designs.
The ASME OM Code, as incorporated by reference in 10 CFR 50.55a, requires in Subsection ISTC, paragraph ISTC-3630, that Category A valves with a leakage requirement not based on the licensees 10 CFR Part 50, Appendix J program, shall be tested to verify that their seat leakage is within acceptable limits. Seat leakage testing shall be performed by using the valve operator with no additional closing force applied. Subparagraph (a) of ASME OM Code, Subsection ISTC, paragraph ISTC-3630, requires that the tests shall be conducted at least once every 2 years.
The NRC regulations in 10 CFR Part 50, Appendix J, Option B, allow a performance-based leakage test program for CIVs in nuclear power plants. Guidance for implementation of acceptable leakage rate test methods, procedures, and analyses is provided in Regulatory Guide (RG) 1.163, Performance-Based Containment Leak-Test Program (ML003740058).
Revision 1 to RG 1.163, June 2023, endorses NEI 94-01, Revision 3-A, dated July 2012 (ML12221A202), with regulatory positions.
As an alternative to the IST requirements in the ASME OM Code, Subsection ISTC, paragraph ISTC-3630, for the specified valves, the licensee proposes in RR-5-V2 to perform testing of the specified PIVs and check valves XVC08703A/B-RH at intervals ranging from every RFO to every fourth RFO. The specific interval for each valve would be a function of its performance and would be established in a manner consistent with the CIV extended test eligibility process guidance under 10 CFR Part 50, Appendix J, Option B. Performance-based scheduling of PIVs and the check valves testing will be controlled in a manner similar to the methods described in NEI 94-01, Revision 3-A, as accepted in RG 1.163. PIV test performances would occur at a nominal frequency ranging from every RFO to every fourth RFO, subject to acceptable valve performance. Valves that have demonstrated good performance for two consecutive cycles may have their test interval extended up to 75 months, with a permissible extension (for non-routine emergent conditions) of nine months (84 months total). The licensee states that a conservative
control will be established such that if any valve fails the PIV test, the test interval will be reduced consistent with 10 CFR Part 50, Appendix J, Option B, requirements. The licensee states that leakage rates less than the leakage limits found in TS and Summer, Unit 1, procedure STP-215.008 will be considered acceptable. Under Request RR-5-V2, any PIV leakage test failure would require the associated component to return to the initial test interval of every RFO or 2 years until acceptable performance is re-established.
During its review, the NRC staff noted that the leakage data provided in RR-5-V2 for the applicable valves show a wide variation in the measured leakage rates over many years. In a request for additional information (RAI), the NRC staff requested the licensee to describe the representative sampling of the subject valves (e.g., plans to obtain test data during outages to support an extended test interval program) within the scope of the request during each RFO for leakage testing when implementing Request RR-5-V2. The NRC staff also requested the licensee to clarify how the historical data of the leakage rates will be used to determine an initial and successive sampling program to ensure that valves with increasing trends do not exceed their leakage limits before their next scheduled test.
In its RAI response dated April 29, 2024, the licensee described the use of performance-based testing intervals for seat leakage testing of the PIVs and closure testing of check valves XVC08703A/B-RH that would be comparable to the performance-based Option B in Appendix J for local leak rate testing. The licensee stated they will specify the leakage testing of these PIVs in the Check Valve Condition Monitoring (CVCM) plans. The licensee stated they will use trending and evaluation of existing data as the bases to justify the time interval between tests.
To support the leakage rate testing intervals, and they will identify applicable preventive maintenance activities and associated intervals required to maintain continued acceptable performance of the valves. Leakage testing of the subject valves will occur at a nominal frequency ranging from every RFO to every fourth RFO, subject to acceptable valve performance. Valves that have demonstrated good performance for two prior consecutive surveillance tests not exceeding the acceptance criteria may have their test interval extended. If a valve exceeds the acceptance criteria, the test interval will be reduced to the initial interval of every RFO, or 2 years, until good performance is re-established. Adequate margin will be trended such that leakage will not be expected to exceed the acceptance criteria prior to the next scheduled test. The licensee stated that they will schedule valve testing such that all of the valves are tested throughout the interval. A minimum of 20 percent of the PIVs will be tested within any 24-month interval consisting of valves that have not been tested during the 75-month interval (with a permissible 9-month extension for nonroutine emergent conditions for a total of 84 months). Group testing of PIVs will be performed with the most limiting acceptance criteria applied for the group of multiple valves. Individual testing of PIVs will be performed for the purpose of identifying leakage and troubleshooting when required. Check valves XVC08703A/B-RH will continue to be tested in parallel with PIVs XVG08702A/B-RH. Separate functional testing of the PIVs and check valves XVC08703A/B-RH will continue to be conducted in accordance with the ASME OM Code.
Based on its the above, the NRC staff finds that the valves within the scope of RR-5-V2 have maintained a history of overall good performance. The NRC staff also finds that extending the leakage test interval based on good performance of specific valves and the low risk factor supports the basis for a performance-based leakage test program. Therefore, the NRC staff concludes that the licensee's proposed alternative provides an acceptable level of quality and safety and would satisfy 10 CFR 50.55a(z)(1) and is, therefore, acceptable.
4.0 CONCLUSION
Based on the above, the NRC staff concludes that the proposed performance-based alternative in RR-5-V2 provides an acceptable level of quality and safety and that that Dominion has adequately addressed all the regulatory requirements set forth in 10 CFR 50.55a(z)(1). Further, the proposed alternative provides reasonable assurance that the applicable components at Summer, Unit 1, will be available to perform their safety functions. Therefore, the NRC staff authorizes the proposed alternative RR-5-V2 for the Fifth Interval IST Program at Summer, Unit 1, that begins on January 1, 2025.
All other ASME Code requirements for which relief has not been specifically requested and approved in this relief request remain applicable, including third party review by the Authorized Nuclear Inservice Inspector.
Principal Contributor: M. Breach, NRR T. Scarbrough, NRR Date: October 22, 2024
- Via SE Input OFFICE NRR/DORL/LPL2-1/PM NRR/DORL/LPL2-1/LA NRR/DEX/EMIB/BC NRR/DORL/LPL2-1/BC NAME GEMiller KGoldstein TScarbrough (for SBailey)* MMarkley DATE 9/6/2024 9/12/2024 8/30/2024 10/22/2024 OFFICE NRR/DORL/LPL2-1/PM NAME GEMiller DATE 10/22/2024