PLA-8074, Proposed Relief Request for the Fifth 10-Year Inservice Test Program Interval

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Proposed Relief Request for the Fifth 10-Year Inservice Test Program Interval
ML23215A173
Person / Time
Site: Susquehanna  Talen Energy icon.png
Issue date: 08/03/2023
From: Casulli E
Susquehanna, Talen Energy
To:
Office of Nuclear Reactor Regulation, Document Control Desk
References
PLA-8074
Download: ML23215A173 (1)
v • d • e


Text

TALEN~

Edward Casulli Susquehanna Nuclear, LLC Site Vice President 769 Salem Boulevard Berwick, PA 18603 ENERGY Tel. 570.542.3795 Fax 570.542.1504 Edward.Casulli@talenenergy.com August 3, 2023 Attn: Document Control Desk 10 CFR 50.55a U. S. Nuclear Regulatory Commission Washington, DC 20555-0001 SUSQUEHANNA STEAM ELECTRIC STATION PROPOSED RELIEF REQUEST FOR THE FIFTH 10-YEAR INSERVICE TEST PROGRAM INTERVAL Docket No. 50-387 PLA-8074 and 50-388

References:

1) Susquehanna letter to NRC, Revised Proposed Relief Requests for the Fourth Ten-Year Inservice Testing Interval for Susquehanna Units 1 and 2 (PLA-7120), dated December 12, 2013 (ADAMS Accession No. ML13347B233)
2) NRC letter to Susquehanna, Relief Requests for the Fourth 10-Year Inservice Testing Interval (TAC Nos. MF2905 through MF2912 and MF2915), dated May 22, 2014 (ADAMS Accession No. ML14122A197)

In accordance with 10 CFR 50.55a, "Codes and standards," paragraph (z)(1), Susquehanna Nuclear, LLC (Susquehanna), requests NRC approval of the enclosed relief requests associated with the Fifth 10-Year Inservice Test (IST) Program Interval for Susquehanna Steam Electric Station (SSES), Units 1 and 2. Relief Request RR01 proposes alternatives to the American Society of Mechanical Engineers (ASME) Operation and Maintenance of Nuclear Power Plants Code (OM Code) test requirements for excess flow check valves. Relief Request RR02 proposes alternatives to the ASME OM Code to revise pressure isolation valve leak test frequency consistent with 10 CFR 50, Appendix J, Option B, for the affected components.

The Fifth 10-Year IST Program Interval at SSES will begin on June 1, 2024, and is currently scheduled to end May 31, 2034. Accordingly, Susquehanna requests authorization of these requests by March 31, 2024, to support the Fifth 10-Year IST Program Interval. The Fifth 10-Year IST Program Interval will comply with the 2020 Edition of the ASME OM Code.

There are no new or revised commitments contained in this submittal.

Document Control Desk PLA-8074 Should you have any questions regarding this submittal, please contact Ms. Melisa Krick, Manager - Nuclear Regulatory Affairs, at (570) 542-1818.

E. Casulli

Enclosures:

1. Relief Request RR01
2. Relief Request RR02 Copy: NRC Region I Mr. C. Highley, NRC Senior Resident Inspector Ms. A. Klett, NRC Project Manager Mr. M. Shields, PA DEP/BRP to PLA-8074 Relief Request RR01

Enclosure 1 of PLA-8074 Page 1 of 9 10 CFR 50.55a Relief Request RR01 Proposed Alternative to the American Society of Mechanical Engineers (ASME) Operation and Maintenance of Nuclear Power Plants Code (OM Code) Test Requirements for Excess Flow Check Valves, in Accordance with 10 CFR 50.55a(z)(1), "Acceptable Level of Quality and Safety"

1. ASME Code Component(s) Affected Unit 1 Nuclear Boiler System Excess Flow Check Valves (EFCVs)

Valve Category Class Valve Cat Class Valve Cat Class (Cat)

XV141F009 C 1 XV142F043B C 1 XV142F059G C 1 XV141F070A C 1 XV142F045A C 1 XV142F059H C 1 XV141F070B C 1 XV142F045B C 1 XV142F059L C 1 XV141F070C C 1 XV142F047A C 1 XV142F059M C 1 XV141F070D C 1 XV142F047B C 1 XV142F059N C 1 XV141F071A C 1 XV142F051A C 1 XV142F059P C 1 XV141F071B C 1 XV142F051B C 1 XV142F059R C 1 XV141F071C C 1 XV142F051C C 1 XV142F059S C 1 XV141F071D C 1 XV142F051D C 1 XV142F059T C 1 XV141F072A C 1 XV142F053A C 1 XV142F059U C 1 XV141F072B C 1 XV142F053B C 1 XV142F061 C 1 XV141F072C C 1 XV142F053C C 1 XV141F072D C 1 XV142F053D C 1 XV141F073A C 1 XV142F055 C 1 XV141F073B C 1 XV142F057 C 1 XV141F073C C 1 XV142F059A C 1 XV141F073D C 1 XV142F059B C 1 XV14201 C 1 XV142F059C C 1 XV14202 C 1 XV142F059D C 1 XV142F041 C 1 XV142F059E C 1 XV142F043A C 1 XV142F059F C 1

Enclosure 1 of PLA-8074 Page 2 of 9

1. ASME Code Component(s) Affected (Cont.)

Unit 1 Reactor Recirculation System EFCVs Valve Cat Class Valve Cat Class Valve Cat Class XV143F003A C 1 XV143F010B C 1 XV143F012C C 1 XV143F003B C 1 XV143F010C C 1 XV143F012D C 1 XV143F004A C 1 XV143F010D C 1 XV143F040A C 1 XV143F004B C 1 XV143F011A C 1 XV143F040B C 1 XV143F009A C 1 XV143F011B C 1 XV143F040C C 1 XV143F009B C 1 XV143F011C C 1 XV143F040D C 1 XV143F009C C 1 XV143F011D C 1 XV143F057A C 1 XV143F009D C 1 XV143F012A C 1 XV143F057B C 1 XV143F010A C 1 XV143F012B C 1 Unit 1 Other System EFCVs Valve System Cat Class Valve System Cat Class XV14411A Reactor Water C 1 XV155F024B High Pressure Coolant C 1 Cleanup Injection XV14411B Reactor Water C 1 XV155F024C High Pressure Coolant C 1 Cleanup Injection XV14411C Reactor Water C 1 XV155F024D High Pressure Coolant C 1 Cleanup Injection XV14411D Reactor Water C 1 XV15109A Residual Heat Removal C 1 Cleanup XV144F046 Reactor Water C 1 XV15109B Residual Heat Removal C 1 Cleanup XV149F044A Reactor Core C 1 XV15109C Residual Heat Removal C 1 Isolation Cooling XV149F044B Reactor Core C 1 XV15109D Residual Heat Removal C 1 Isolation Cooling XV149F044C Reactor Core C 1 XV152F018A Core Spray C 1 Isolation Cooling XV149F044D Reactor Core C 1 XV152F018B Core Spray C 1 Isolation Cooling XV155F024A High Pressure C 1 Coolant Injection

Enclosure 1 of PLA-8074 Page 3 of 9

1. ASME Code Component(s) Affected (Cont.)

Unit 2 Nuclear Boiler System EFCVs Valve Cat Class Valve Cat Class Valve Cat Class X241F009 C 1 XV242F043B C 1 XV242F059G C 1 XV241F070A C 1 XV242F045A C 1 XV242F059H C 1 XV241F070B C 1 XV242F045B C 1 XV242F059L C 1 XV241F070C C 1 XV242F047A C 1 XV242F059M C 1 XV241F070D C 1 XV242F047B C 1 XV242F059N C 1 XV241F071A C 1 XV242F051A C 1 XV242F059P C 1 XV241F071B C 1 XV242F051B C 1 XV242F059R C 1 XV241F071C C 1 XV242F051C C 1 XV242F059S C 1 XV241F071D C 1 XV242F051D C 1 XV242F059T C 1 XV241F072A C 1 XV242F053A C 1 XV242F059U C 1 XV241F072B C 1 XV242F053B C 1 XV242F061 C 1 XV241F072C C 1 XV242F053C C 1 XV241F072D C 1 XV242F053D C 1 XV241F073A C 1 XV242F055 C 1 XV241F073B C 1 XV242F057 C 1 XV241F073C C 1 XV242F059A C 1 XV241F073D C 1 XV242F059B C 1 XV24201 C 1 XV242F059C C 1 XV24202 C 1 XV242F059D C 1 XV242F041 C 1 XV242F059E C 1 XV242F043A C 1 XV242F059F C 1

Enclosure 1 of PLA-8074 Page 4 of 9

1. ASME Code Component(s) Affected (Cont.)

Unit 2 Reactor Recirculation System EFCVs Valve Cat Class Valve Cat Class Valve Cat Class XV243F003A C 1 XV243F010B C 1 XV243F012C C 1 XV243F003B C 1 XV243F010C C 1 XV243F012D C 1 XV243F004A C 1 XV243F010D C 1 XV243F040A C 1 XV243F004B C 1 XV243F011A C 1 XV243F040B C 1 XV243F009A C 1 XV243F011B C 1 XV243F040C C 1 XV243F009B C 1 XV243F011C C 1 XV243F040D C 1 XV243F009C C 1 XV243F011D C 1 XV243F057A C 1 XV243F009D C 1 XV243F012A C 1 XV243F057B C 1 XV243F010A C 1 XV243F012B C 1 Unit 2 Other System EFCVs Valve System Cat Class Valve System Cat Class XV24411A Reactor Water C 1 XV255F024B High Pressure Coolant C 1 Cleanup Injection XV24411B Reactor Water C 1 XV255F024C High Pressure Coolant C 1 Cleanup Injection XV24411C Reactor Water C 1 XV255F024D High Pressure Coolant C 1 Cleanup Injection XV24411D Reactor Water C 1 XV25109A Residual Heat Removal C 1 Cleanup XV244F046 Reactor Water C 1 XV25109B Residual Heat Removal C 1 Cleanup XV249F044A Reactor Core C 1 XV25109C Residual Heat Removal C 1 Isolation Cooling XV249F044B Reactor Core C 1 XV25109D Residual Heat Removal C 1 Isolation Cooling XV249F044C Reactor Core C 1 XV252F018A Core Spray C 1 Isolation Cooling XV249F044D Reactor Core C 1 XV252F018B Core Spray C 1 Isolation Cooling XV255F024A High Pressure C 1 Coolant Injection These valves are instrumentation line EFCVs provided in each instrument process line that penetrates primary containment in accordance with Regulatory Guide 1.11, Instrument Lines Penetrating the Primary Reactor Containment. The EFCVs are designed to close upon rupture of the instrument line downstream of the EFCV and otherwise remain open. The lines are sized and/or orificed such that the Control Room Habitability Envelope and off-site dose from an instrument line break will be substantially below 10 CFR 50.67 dose guideline limits.

Enclosure 1 of PLA-8074 Page 5 of 9

2. Applicable Code Edition and Addenda

The Susquehanna Steam Electric Station (SSES), Units 1 and 2, will start the Fifth 10-Year Inservice Test (IST) Program Interval on June 1, 2024, and is required to comply with 2020 Edition, no Addenda, of the ASME OM Code.

3. Applicable Code Requirement

ASME OM Code Subsection ISTC-3522, Category C Check Valves, subparagraph (c) states, that If exercising is not practicable during operation at power and cold shutdown outages, it shall be performed during refueling outages.

ASME OM Code Subsection ISTC- 3700, Position Verification Testing, states, in part, that Valves with remote position indicators shall be observed locally at least once every 2 yr to verify that valve operation is accurately indicated.

10 CFR 50.55a(b)(3)(xi), OM condition: Valve Position Indication, states, When implementing paragraph ISTC-3700, "Position Verification Testing," in the ASME OM Code, 2012 Edition through the latest edition of the ASME OM Code incorporated by reference in paragraph (a)(1)(iv) of this section, licensees must verify that valve operation is accurately indicated by supplementing valve position indicating lights with other indications, such as flow meters or other suitable instrumentation to provide assurance of proper obturator position for valves with remote position indication within the scope of Subsection ISTC including its mandatory appendices and their verification methods and frequencies.

4. Reason for Request

Pursuant to 10 CFR 50.55a, Codes and standards, paragraph (z)(1), an alternative is proposed to the testing requirements of ASME OM Code ISTC-3522(c), ISTC-3700, and 10 CFR 50.55a(b)(3)(xi) for the affected components on the basis that the alternative testing would provide an acceptable level of quality and safety.

EFCVs are required to be tested in accordance with ISTC-3522, which requires exercising check valves nominally every three months to the positions required to perform their safety functions. ISTC-3522(c) permits deferral of this requirement to every reactor refueling outage. EFCVs are also required to be tested in accordance with ISTC-3700, which requires remote position indication verification at least once every 2-years. 10 CFR 50.55a(b)(3)(xi) requires supplementing the ISTC-3700 testing with other indications to ensure valve position indicating lights accurately reflect valve operation. Functional testing is also required by Technical Specification (TS) Surveillance Requirement (SR) 3.6.1.3.9.

Enclosure 1 of PLA-8074 Page 6 of 9 Testing the subject valves quarterly or during cold shutdown is not practicable based on plant conditions. Reactor pressure greater than 500 pounds per square inch gauge (psig) is needed for testing. System design does not include test taps upstream of the EFCV. For this reason, the EFCVs cannot be isolated and tested using a pressure source other than reactor pressure.

Major components of EFCVs are the poppet and spring. The spring holds the poppet open under static conditions. The valve will close upon sufficient differential pressure across the poppet. Functional testing of the valve is accomplished by venting the instrument side of the valve. The resultant increase in flow imposes a differential pressure across the poppet, which compresses the spring and decreases flow through the valve.

The testing described above requires removal of the associated instrument(s) from service.

Since these instruments are in use during plant operation, removal of any of these instruments from service may cause a spurious signal, which could result in a plant trip or an unnecessary challenge to safety systems. Additionally, process liquid will be contaminated to some degree, requiring special measures to collect flow from the vented instrument side and will contribute to an increase in personnel radiation exposure.

The EFCVs are containment isolation valves, but are classified as ASME OM Code Category C valves. These valves are excluded from 10 CFR 50, Appendix J Type C leak rate testing due to the size of the instrument lines and upstream orifices. Therefore, they have no safety-related seat leakage criterion.

Industry experience as documented in NEDO-32977-A, Excess Flow Check Valve Testing Relaxation (NEDO), indicates the ECFVs have a low failure rate. The NEDO indicates that many reported test failures were related to test methodologies and not actual EFCV failures. The SSES test history shows no evidence of common mode failure, and has demonstrated that EFCVs are highly reliable and that failures to isolate are very infrequent, which is consistent with the findings of NEDO-32977-A.

Although highly unlikely, in the event of a rupture, or failure to isolate, the lines are sized and/or orificed such that off-site dose will be substantially below 10 CFR 50.67 limits.

Thus, the testing of ECFVs at SSES is consistent with the industry, and has exhibited a high degree of reliability, availability, and provide an acceptable level of quality and safety.

In accordance with TS SR 3.6.1.3.9 bases and the IST Program Plan, EFCV failures will be evaluated to determine if additional testing in that test interval is warranted to ensure overall reliability. Adverse trends and EFCV performance are identified and disposition in the Corrective Action Program.

The appropriate time for performing EFCV test is during refueling outages in conjunction with vessel hydrostatic testing. Recent improvements in refueling outage schedules minimized the time that is planned for refueling and testing activities during the outages.

Enclosure 1 of PLA-8074 Page 7 of 9 As a result of shortened outages, decay heat levels during hydrostatic tests are higher than in the past. If the hydrostatic test were extended to test all EFCVs, the vessel could require depressurization several times to avoid exceeding the maximum bulk coolant temperature limit. This is an evolution that challenges the reactor operators and thermally cycles the reactor vessel. This evolution should be avoided if possible. Also, based on past experience, EFCV testing during hydrostatic testing becomes the outage critical path and could possibly extend the outage by two days if all EFCVs were to be tested during this time frame.

5. Proposed Alternative and Basis for Use

As an alternative to testing all EFCVs during a single refueling outage, a sampling plan will be implemented.

The basis for this alternative is that testing a sample of EFCVs each refueling outage provides a level of safety and quality equivalent to that of the ASME OM Code-required testing. Functional testing with verification that flow is checked will be performed per TS SR 3.6.1.3.9, either immediately preceding a planned refueling outage or during the refueling outage for certain EFCVs.

EFCVs will be tested on a representative sampling basis in accordance with TS SR 3.6.1.3.9, such that each EFCV will be tested at least once every ten years (nominal).

The EFCVs have position indication in the control room. Check valve remote position indication is excluded from Regulatory Guide 1.97, Criteria for Accident Monitoring Instrumentation for Nuclear Power Plants, as a required parameter for evaluating containment isolation. The remote position indication will be verified in the closed direction on a representative sampling basis to coincide with the associated exercise test.

After the close position test, the valve will be reset, and the remote open position indication will be verified. 10 CFR 50.55a(b)(3)(xi) requires obturator verification at the same frequency as ISTC-3700 (i.e., 2-years). Obturator verification for EFCVs equates to the position of the poppet assembly which is verified by secession of flow during testing, and detected by change to the status of indicating lights. The current methodology used for functional testing of the EFCVs will be credited to meet the obturator verification requirements. However, this obturator verification for the EFCVs will be performed on the same representative sampling basis as the functional testing prescribed in TS SR 3.6.1.3.9. Although inadvertent actuation of an EFCV during operation is highly unlikely due to the spring poppet design, SSES monitors EFCVs in the control room as part of normal operator rounds. EFCV test failures and EFCVs with abnormal position indication displays will be entered into the Corrective Action Program.

In summary, considering the extremely low failure rate along with plant safety concerns to perform testing, the proposed alternative to perform EFCV testing on a sampling basis will

Enclosure 1 of PLA-8074 Page 8 of 9 continue to provide assurance of the EFCVs operational readiness and provides an acceptable level of quality and safety pursuant to 10 CFR 50.55a(z)(1).

6. Duration of Proposed Alternative

The proposed alternative will be utilized for the entire Fifth 10-Year IST Program Interval, which is scheduled to begin on June 1, 2024, and end on May 31, 2034.

7. Precedent
1. Letter from M. Markley (NRC) to C. R. Pierce (Southern Nuclear Operating Company, Inc.), "Edwin I. Hatch Nuclear Plant, Units 1 and 2 - Inservice Testing Program Relief Request and Alternatives for Pumps and Valves - Fifth Ten-Year Interval (CAC Nos. MF6238, MF6239, MF6240, MF6241, MF6242, MF6243, MF6244, MF6245, MF6246, and MF6247)," dated December 30, 2015 (ADAMS Accession No. ML15310A406)
2. Letter from D. J. Wrona (NRC) to B. C. Hanson (Exelon Generation Company, LLC),

LaSalle County Station, Units 1 and 2 - Relief from the Requirements of the ASME Code for Operation and Maintenance of Nuclear Power Plants (EPID L-2018-LLR-0004), dated July 3, 2018 (ADAMS Accession No. ML18163A054)

3. Letter from L. M. Regner (NRC) to P. Fessler (DTE Electric Company), "Fermi 2 -

Proposed Alternative to the Required Examination Associated with Valves (EPID L-2019-LLR-0047 and EPID L-2019-LLR-0050)," dated October 3, 2019 (ADAMS Accession No. ML19248C707)

4. Letter from J. G. Danna (NRC) to D. P Rhoades (Constellation Energy Generation, LLC), Nine Mile Point Nuclear Station, Unit 2 - Relief Request Associated with Excess Flow Check Valves (EPID L-2021-LLR-0066), dated March 11, 2022 (ADAMS Accession No. ML22061A040)

Enclosure 1 of PLA-8074 Page 9 of 9

8. References
1. Letter from S. A. Richards (NRC) to W. G. Warren (BWR Owners Group), Safety Evaluation of General Electric Nuclear Energy Topical Report B21-00658-01, Excess Flow Check Valve Testing Relaxation (TAC Nos. MA7884 and M84809), dated March 14, 2000 (ADAMS Accession No. ML003691722)
2. GE Nuclear Energy Topical Report NEDO-32977-A, "Excess Flow Check Valve Testing Relaxation," dated June 2000 (ADAMS Accession No. ML003729011)
3. Letter from M. K. Khanna (NRC) to T. S. Rausch (PPL Susquehanna, LLC),

Susquehanna Steam Electric Station, Units 1 and 2 - Relief Requests for the Fourth 10-Year Inservice Testing Interval (TAC Nos. MF2905 through MF2912 and MF2915), dated May 22, 2014 (ADAMS Accession No. ML14122A197)

4. Regulatory Guide 1.11, Instrument Lines Penetrating the Primary Reactor Containment.
5. Regulatory Guide 1.97, Criteria for Accident Monitoring Instrumentation for Nuclear Power Plants.

to PLA-8074 Relief Request RR02

Enclosure 2 of PLA-8074 Page 1 of 16 10 CFR 50.55a Relief Request RR02 Proposed Alternative to the American Society of Mechanical Engineers (ASME) Operation and Maintenance of Nuclear Power Plants Code (OM Code) to Revise Pressure Isolation Valve Leak Test Frequency Consistent with 10 CFR 50, Appendix J, Option B, in Accordance with 10 CFR 50.55a(z)(1), "Acceptable Level of Quality and Safety"

1. ASME Code Component(s) Affected Component Number Valve CIV, System Code Category Unit Type PIV, Class Both HV151F008 MO Gate Both RHR 1 A 1 HV151F009 MO Gate Both RHR 1 A 1 HV151F015A/B MO Gate Both RHR 1 A 1 HV151F022 MO Gate Both RHR 1 A 1 HV151F023 MO Globe Both RHR 2 A 1 HV151F050A/B Check PIV RHR 1 A/C 1 HV151F122A/B AO Globe PIV RHR 1 A 1 151130 Check PIV RHR 1 A/C 1 HV152F005A/B MO Gate Both CS 1 A 1 HV152F006A/B Check Both CS 1 A/C 1 HV152F037A/B AO Globe Both CS 1 A 1 HV251F008 MO Gate Both RHR 1 A 2 HV251F009 MO Gate Both RHR 1 A 2 HV251F015A/B MO Gate Both RHR 1 A 2 HV251F022 MO Gate Both RHR 1 A 2 HV251F023 MO Globe Both RHR 2 A 2 HV251F050A/B Check PIV RHR 1 A/C 2 HV251F122A/B AO Globe PIV RHR 1 A 2 251130 Check PIV RHR 1 A/C 2 HV252F005A/B MO Gate Both CS 1 A 2 HV252F006A/B Check Both CS 1 A/C 2 HV252F037A/B AO Globe Both CS 1 A 2 AO = Air Operated; MO = Motor Operated; CIV = containment isolation valve; PIV = pressure injection valve; RHR= Residual Heat Removal System; CS = Core Spray These valves are the Category A and A/C PIVs for RHR and CS for Susquehanna Steam Electric Station (SSES). They provide isolation and prevent over pressurization of the low-pressure piping between the Emergency Core Cooling System and Reactor Coolant System boundaries.

Enclosure 2 of PLA-8074 Page 2 of 16

2. Applicable Code Edition and Addenda

SSES, Units 1 and 2, will start the Fifth 10-Year Inservice Test (IST) Program Interval on June 1, 2024, and is required to comply with 2020 Edition, no Addenda, of the ASME OM Code.

3. Applicable Code Requirement

ASME OM Code Subsection ISTC-3630, Leakage Rate for Other Than Containment Isolation Valves, states, Category A valves with a leakage requirement not based on an Owner's 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-3630(a), Frequency, states, Tests shall be conducted at least once every 2 yr.

4. Reason for Request

Pursuant to 10 CFR 50.55a, Codes and standards, paragraph (z)(1), an alternative is proposed to the testing requirements of ASME OM Code ISTC-3630(a) for the affected components on the basis that the alternative testing would provide an acceptable level of quality and safety.

ISTC-3630(a) requires that leakage rate testing for PIVs be performed at least once every two years. PIVs are not specifically included in the scope for performance-based testing as provided for in 10 CFR 50, Appendix J, Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors, Option B, Performance-Based Requirements, (referred to hereafter as Option B).

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, Nuclear Energy Institute (NEI) 94-01, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, describes the risk-informed basis for the extended test intervals under Option B. That justification shows that valves, which have demonstrated good performance by the successful completion of two consecutive leak rate tests for two consecutive cycles may increase their test interval. Furthermore, it states that if the component does not fail within two operating cycles, further failures appear to be governed by the random failure rate of the component. NEI 94-01 also presents the results of a comprehensive risk analysis, including the statement that "the risk impact associated with increasing [leak rate] test intervals are negligible (less than 0.1 percent of total risk)."

Enclosure 2 of PLA-8074 Page 3 of 16 The valves identified in this request are installed in water applications. The PIV testing is performed with water pressurized to normal plant operating pressures in accordance with ISTC-3630.

This relief request is intended to provide for performance-based scheduling of PIV leakage tests at SSES. The reason for proposing this alternative request is dose reduction/industry As Low as Reasonably Achievable (ALARA) radiation dose principles. Recent historical data was used to identify that PIV testing alone incurs a total dose of approximately 500 millirem each refueling outage. Assuming the affected PIVs continue to remain classified as good performers, the extended test intervals would provide for a savings of approximately 1.0 Rem over a 4 1/2 year period (a bounding timeframe encompassing two refueling outages).

NUREG-0933, "Resolution of Generic Safety Issues," Issue 105, Interfacing Systems LOCA at LWRs, discussed the need for PIV leak rate testing based primarily on three pre-1980 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.

It has been shown that Interfacing Systems Loss of Coolant Accident (ISLOCA) represents a small risk impact to Boiling Water Reactors (BWR) such as SSES.

NUREG/CR-5928, "ISLOCA Research Program," evaluated the likelihood and potential severity of ISLOCA events in BWRs and Pressurized Water Reactors (PWR). The BWR design used as a reference for this analysis was a BWR-4 with Mark I containment. SSES is listed as a similar plant. The BWR systems were individually analyzed and in each case the report concluded that the system was "judged to not be an important consideration with respect to ISLOCA risk." Section 4.3 of the report concluded the BWR portion of the analysis by saying "ISLOCA is not a risk concern for the BWR plant examined here."

Typical PIV testing does not identify functional problems, which may inhibit the valves ability to reposition from open to closed. For check valves, such functional testing is accomplished per ASME OM Code paragraphs ISTC-3522, Category C Check Valves, and ISTC-3520, Exercising Requirements. Power-operated valves are routinely full stroke tested per ASME OM Code to ensure their functional capabilities.

At SSES, the functional tests for motor operated PIVs are performed on a 2-year frequency in accordance with ASME OM Code, Division 1, Mandatory Appendix III, Preservice and Inservice Testing of Active Electric Motor-Operated Valve Assemblies in Water-Cooled Reactor Nuclear Power Plants. Such testing is not performed online in order to prevent any possibility of an inadvertent ISLOCA condition. The functional

Enclosure 2 of PLA-8074 Page 4 of 16 testing of the PIVs is adequate to identify any abnormal condition that might affect closure capability.

The functional testing of the PIV check valves is performed in accordance with ISTC-5221, Valve Obturator Movement. Performance of the separate 2-year PIV leak rate testing does not contribute any additional assurance of functional capability.

The above tests provide reasonable assurance of the valves' operational readiness. PIV testing only determines the seat tightness of the closed valves, does not confirm movement (open/closed).

5. Proposed Alternative and Basis for Use

SSES proposes to perform PIV leak rate testing at intervals ranging from every refueling outage to every third 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 process under Option B program guidance. The test frequency will be established such that if any PIV test fail, the test interval will be reduced to once every 24 months until the valve can be re-classified as a good performer per the performance evaluation requirements of Option B. The test intervals for the valves with a PIV-only function will be determined in a similar manner as is done for CIV testing under Option B. The test interval may be extended upon completion of two consecutive periodic PIV tests with results within prescribed acceptance criteria. Any PIV test failure will require a return to the initial interval until good performance can again be established.

The primary basis for this proposed alternative is the historically good performance of the PIVs. The tables in Attachment RR02-1 present test data that demonstrate acceptable historical PIV performance for the RHR and CS systems.

The extension of test frequencies will be consistent with the guidance provided for Appendix J Type C leak rate tests as detailed in paragraph 10.2.3.2, "Extended Test Interval," of NEI 94-01, 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 [of NEI 94-01].

Enclosure 2 of PLA-8074 Page 5 of 16 Note that NEI 94-01 is not the sole basis for this request given NEI 94-01 does not address seat leakage testing with water. This document was cited as an approach similar to the requested alternative method.

Additional basis for this request is provided below:

Separate functional testing of motor operated valve (MOV) PIVs and Check Valve PIVs per ASME OM Code.

Low likelihood of valve mispositioning during power operations (e.g.,

procedures, interlocks).

Relief valves in the low-pressure piping - provide over pressure protection to the low-pressure piping.

Alarms that identify high pressure to low-pressure leakage - Operators are highly trained to recognize symptoms of a present or incipient ISLOCA and to take appropriate actions.

Therefore, the proposed alternative to perform PIV testing at the specified intervals will continue to provide assurance of the PIVs operational readiness and provides an acceptable level of quality and safety pursuant to 10 CFR 50.55a(z)(1).

6. Duration of Proposed Alternative

The proposed alternative will be utilized for the entire Fifth 10-Year IST Program Interval, which is scheduled to begin on June 1, 2024, and end on May 31, 2034.

7. Precedent
1. Letter from T. L. Tate (NRC) to B. Hanson (Exelon Generation Company, LLC),

"Dresden Nuclear Power Station, Units 2 and 3 - Relief Request to Use an Alternative from the American Society of Mechanical Engineers Code Requirements (CAC Nos.

MF5089 and MF5090)," dated October 27, 2015 (ADAMS Accession No. ML15174A303)

2. Letter from J. G. Danna (NRC) to B. C. Hanson (Exelon Generation Company, LLC),

"Peach Bottom Atomic Power Station, Units 2 and 3 - Safety Evaluation of Relief Request GVRR-2 Regarding the Fifth 10-Year Interval of the lnservice Testing Program (EPID No. L-2017-LLR-0094)," dated May 30, 2018 (ADAMS Accession No. ML18141A600)

3. Letter from J. G. Danna (NRC) to B. C. Hanson (Exelon Generation Company, LLC),

"Nine Mile Point Nuclear Station, Units 1 and 2 - Relief from the Requirements of the

Enclosure 2 of PLA-8074 Page 6 of 16 ASME Code (EPID L-2017-LLR-0145 through EPID L-2017-LLR-0152)," dated November 13, 2018 (ADAMS Accession No. ML18275A139)

4. Letter from L. M. Regner (NRC) to B. C. Hanson (Exelon Generation Company, LLC), "Lasalle County Station, Units 1 and 2 - Request from the Requirements of the ASME Code Related to Pressure Isolation Valve Testing Frequency (EPID L-2019-LLR-0062)," dated September 10, 2019 (ADAMS Accession No. ML19217A306)
5. Letter from J. G. Danna (NRC) to B. C. Hanson (Exelon Generation Company, LLC),

"Limerick Generating Station, Units 1 and 2 - Safety Evaluation of Relief Requests GVRR-8, 11-PRR-1, 90-PRR-1 and 47-VRR-2 Regarding the Fourth 10-Year Interval of the lnservice Testing Program (EPID L-2018-LLR-0384, EPID L-2018-LLR-0385, EPID L-2018-LLR-0386, and EPID L-2018-LLR-0387)," dated October 28, 2019 (ADAMS Accession No. ML19228A195)

6. Letter from J. L. Dixon-Herrity (NRC) to Vice President, Operations (Entergy Operations, Inc.), Grand Gulf Nuclear Station, Unit 1 - Inservice Testing Program Relief Request VRR-GGNS-2021-1, Alternative Request for Pressure Isolation Valve Testing Frequency (EPID L-2021-LLR-0040), dated October 28, 2021 (ADAMS Accession No. ML21294A067)
7. Letter from N. L. Salgado (NRC) to D. P. Rhoades (Constellation Energy Generation, LLC), "Quad Cities Nuclear Power Station, Units 1 and 2 - Proposed Alternative to the Requirements of the ASME OM Code (EPID L-2022-LLR-0014)," dated September 29, 2022 (ADAMS Accession No. ML22256A115)
8. References
1. NEI 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," Revision 3-A, dated July 2012
2. NUREG-0933, "Resolution of Generic Safety Issues, Issue 105, Interfacing Systems LOCA at LWRs, dated December 2011
3. NUREG/CR-5928, "ISLOCA Research Program," dated July 1993 (ADAMS Accession No. ML072430731)
4. Letter from M. K. Khanna (NRC) to T. S. Rausch (PPL Susquehanna, LLC),

Susquehanna Steam Electric Station, Units 1 and 2 - Relief Requests for the Fourth 10-Year Inservice Testing Interval (TAC Nos. MF2905 through MF2912 and MF2915), dated May 22, 2014 (ADAMS Accession No. ML14122A197)

Enclosure 2 of PLA-8074 Page 7 of 16 Attachment RR02-1: Leakage History of SSES, Unit 1 and 2 PIVs The tables below summarize the leakage history for the SSES Unit 1 and 2 RHR and CS systems PIVs for a minimum of the last 10-years.

Unit 1/2 RHR Suction PIVs Measured Value Leakage Valve Number Test Date Comments (gpm) Limit (gpm)

HV151F008 03/23/2010 0.196 5 HV151F008 05/03/2012 0.210 5 HV151F008 05/10/2014 0.251 5 HV151F008 04/06/2016 0.182 5 HV151F008 04/18/2018 0.232 5 HV151F008 04/21/2018 0.160 5 HV151F008 04/15/2020 0.172 5 HV151F008 04/18/2022 0.170 5 HV151F009 03/22/2010 0.184 5 HV151F009 05/04/2012 0.330 5 HV151F009 05/10/2014 0.282 5 HV151F009 04/06/2016 0.214 5 HV151F009 04/18/2018 0.235 5 HV151F009 04/14/2020 0.243 5 HV151F009 04/16/2020 0.160 5 HV151F009 04/16/2022 0.211 5 HV251F008 03/28/2003 0.147 5 HV251F008 03/08/2005 0.188 5 HV251F008 03/26/2007 0.260 5 HV251F008 05/06/2009 0.270 5 HV251F008 04/23/2011 0.320 5 HV251F008 05/09/2013 0.288 5 HV251F008 05/10/2015 0.224 5 HV251F008 04/12/2019 0.514 5 HV251F008 04/12/2021 0.430 5 HV251F009 03/28/2003 0.132 5 HV251F009 03/08/2005 0.120 5 HV251F009 03/26/2007 0.160 5 HV251F009 05/06/2009 0.200 5 HV251F009 04/24/2011 0.200 5 HV251F009 05/09/2013 0.243 5 HV251F009 05/10/2015 0.282 5 HV251F009 04/12/2019 0.305 5 HV251F009 04/09/2021 0.329 5

Enclosure 2 of PLA-8074 Page 8 of 16 Attachment RR02-1: Leakage History of SSES, Units 1 and 2 PIVs Unit 1/2 RHR Suction PIVs Measured Value Leakage Valve Number Test Date Comments (gpm) Limit (gpm) 151130 04/07/2016 0.002 0.5 151130 04/19/2018 0.000 0.5 151130 04/14/2020 0.032 0.5 151130 04/16/2022 Would not pressurize (WNP) 0.5 WNP 151130 04/19/2022 0.000 0.5 251130 05/07/2009 0.000 0.5 251130 04/24/2011 0.000 0.5 251130 05/09/2013 WNP 0.5 WNP 251130 05/11/2013 0.000 0.5 251130 05/10/2015 0.000 0.5 251130 04/12/2019 0.012 0.5 251130 04/10/2021 0.000 0.5

Enclosure 2 of PLA-8074 Page 9 of 16 Unit 1 RHR Injection PIVs Measured Value Leakage Valve Number Test Date Comments (gpm) Limit (gpm)

HV151F015A 03/28/2010 0.000 5 HV151F015A 05/03/2012 0.100 5 HV151F015A 05/03/2014 0.024 5 HV151F015A 04/09/2016 0.042 5 HV151F015A 04/04/2018 0.058 5 HV151F015A 04/28/2018 0.049 5 HV151F015A 04/01/2020 0.059 5 HV151F015A 04/08/2020 0.045 5 HV151F050A /

03/11/2010 0.031 0.5 HV151F122A HV151F050A /

04/06/2012 0.035 0.5 HV151F122A HV151F050A /

05/03/2012 0.100 0.5 HV151F122A HV151F050A /

05/15/2014 0.043 0.5 HV151F122A HV151F050A /

04/10/2016 0.009 0.5 HV151F122A HV151F050A /

04/04/2018 0.035 0.5 HV151F122A HV151F050A /

04/01/2020 0.027 0.5 HV151F122A

Enclosure 2 of PLA-8074 Page 10 of 16 Attachment RR02-1: Leakage History of SSES, Units 1 and 2 PIVs Unit 1 RHR Injection PIVs Measured Value Leakage Valve Number Test Date Comments (gpm) Limit (gpm)

HV151F015B 04/03/2010 0.008 5 HV151F015B 05/05/2012 0.009 5 HV151F015B 04/27/2014 0.006 5 HV151F015B 03/21/2016 0.000 5 HV151F015B 04/16/2018 0.004 5 HV151F015B 04/01/2022 0.000 5 HV151F050B /

HV151F122B 03/13/2010 0.012 0.5 HV151F050B /

HV151F122B 04/03/2010 0.016 0.5 HV151F050B /

HV151F122B 05/05/2012 0.006 0.5 HV151F050B /

HV151F122B 04/29/2014 0.008 0.5 HV151F050B /

HV151F122B 03/21/2016 0.000 0.5 HV151F050B /

HV151F122B 04/16/2018 0.004 0.5 HV151F050B /

HV151F122B 04/03/2022 0.000 0.5

Enclosure 2 of PLA-8074 Page 11 of 16 Attachment RR02-1: Leakage History of SSES, Units 1 and 2 PIVs Unit 2 RHR Injection PIVs Measured Value Leakage Valve Number Test Date Comments (gpm) Limit (gpm)

HV251F015A 04/20/2009 WNP 5 WNP HV251F015A 04/21/2009 0.580 5 HV251F015A 04/23/2011 0.082 5 HV251F015A 04/18/2013 0.024 5 HV251F015A 05/02/2015 0.004 5 HV251F015A 03/16/2017 0.030 5 HV251F015A 03/26/2021 0.008 5 HV251F050A /

HV251F122A 04/10/2009 0.110 0.5 HV251F050A /

HV251F122A 04/9/2011 0.332 0.5 HV251F050A /

HV251F122A 04/20/2013 0.078 0.5 HV251F050A /

HV251F122A 05/03/2015 0.409 0.5 HV251F050A /

HV251F122A 03/16/2017 0.450 0.5 HV251F050A /

HV251F122A 04/06/2017 0.0254 0.5 HV251F050A /

HV251F122A 04/05/2019 0.036 0.5 HV251F050A /

HV251F122A 03/26/2021 0.012 0.5

Enclosure 2 of PLA-8074 Page 12 of 16 Attachment RR02-1: Leakage History of SSES, Units 1 and 2 PIVs Unit 2 RHR Injection PIVs Measured Value Leakage Valve Number Test Date Comments (gpm) Limit (gpm)

HV251F015B 03/21/2007 0.020 5 HV251F015B 04/29/2009 0.020 5 HV251F015B 04/25/2011 0.055 5 HV251F015B 04/26/2013 0.027 5 HV251F015B 04/22/2015 0.006 5 HV251F015B 03/26/2019 0.031 5 HV251F050B /

HV251F122B 03/09/2007 0.020 0.5 HV251F050B /

HV251F122B 04/16/2009 0.018 0.5 HV251F050B /

HV251F122B 04/30/2009 0.006 0.5 HV251F050B /

HV251F122B 04/14/2011 0.063 0.5 HV251F050B /

HV251F122B 04/25/2011 0.000 0.5 HV251F050B /

HV251F122B 04/26/2013 0.020 0.5 HV251F050B /

HV251F122B 04/22/2015 0.000 0.5 HV251F050B /

HV251F122B 03/26/2019 0.0156 0.5

Enclosure 2 of PLA-8074 Page 13 of 16 Attachment RR02-1: Leakage History of SSES, Units 1 and 2 PIVs Unit 1/2 RHR PIVs Measured Value Leakage Valve Number Test Date Comments (gpm) Limit (gpm)

HV151F022 04/08/2010 0.000 3 HV151F022 05/17/2012 0.000 3 HV151F022 05/25/2014 0.000 3 HV151F022 04/14/2016 0.000 3 HV151F022 03/29/2022 0.000 3 HV151F023 04/08/2010 0.000 3 HV151F023 05/17/2012 0.002 3 HV151F023 05/25/2014 0.000 3 HV151F023 04/14/2016 0.000 3 HV151F023 03/29/2022 0.000 3 HV251F022 04/05/2007 0.000 3 HV251F022 05/10/2009 0.000 3 HV251F022 05/07/2011 0.002 3 HV251F022 05/19/2013 0.000 3 HV251F022 05/18/2015 0.000 3 HV251F022 04/01/2017 0.000 3 HV251F022 04/16/2019 0.000 3 HV251F022 04/14/2021 0.000 3 HV251F023 04/05/2007 0.000 3 HV251F023 05/10/2009 0.000 3 HV251F023 05/07/2011 0.002 3 HV251F023 05/19/2013 0.000 3 HV251F023 05/18/2015 0.000 3 HV251F023 04/01/2017 0.000 3 HV251F023 04/16/2019 0.000 3 HV251F023 04/14/2021 0.000 3

Enclosure 2 of PLA-8074 Page 14 of 16 Attachment RR02-1: Leakage History of SSES, Units 1 and 2 PIVs Unit 1 CS Injection PIVs Measured Value Leakage Valve Number Test Date Comments (gpm) Limit (gpm)

HV152F005A 03/22/2008 0.009 5 HV152F005A 03/08/2010 0.000 5 HV152F005A 04/04/2012 0.000 5 HV152F005A 04/20/2014 0.000 5 HV152F005A 03/24/2016 0.000 5 HV152F005A 04/01/2020 0.000 5 HV152F006A /

HV152F037A 03/22/2008 0.006 0.5 HV152F006A /

HV152F037A 03/08/2010 0.000 0.5 HV152F006A /

HV152F037A 04/04/2012 0.004 0.5 HV152F006A /

HV152F037A 04/20/2014 0.000 0.5 HV152F006A /

HV152F037A 03/24/2016 0.000 0.5 HV152F006A /

HV152F037A 04/02/2020 0.000 0.5 HV152F005B 03/22/2010 0.000 5 HV152F005B 04/24/2012 0.000 5 HV152F005B 04/27/2014 0.008 5 HV152F005B 04/17/2018 0.000 5 HV152F005B 03/31/2022 0.000 5 HV152F006B /

HV152F037B 03/22/2010 0.009 0.5 HV152F006B /

HV152F037B 04/24/2012 0.012 0.5 HV152F006B /

HV152F037B 04/28/2014 0.008 0.5 HV152F006B /

HV152F037B 04/17/2018 0.000 0.5 HV152F006B /

HV152F037B 03/31/2022 0.000 0.5

Enclosure 2 of PLA-8074 Page 15 of 16 Attachment RR02-1: Leakage History of SSES, Units 1 and 2 PIVs Unit 2 CS Injection PIVs Measured Value Leakage Valve Number Test Date Comments (gpm) Limit (gpm)

HV252F005A 04/19/2009 0.000 5 HV252F005A 04/12/2011 0.000 5 HV252F005A 01/30/2013 0.000 5 HV252F005A 04/16/2015 0.000 5 HV252F005A 03/16/2017 0.000 5 HV252F005A 03/25/2021 0.000 5 HV252F005A 04/01/2021 0.000 5 HV252F006A /

HV252F037A 04/19/2009 0.000 0.5 HV252F006A /

HV252F037A 04/12/2011 0.000 0.5 HV252F006A /

HV252F037A 04/14/2011 0.000 0.5 HV252F006A /

HV252F037A 04/30/2013 0.000 0.5 HV252F006A /

HV252F037A 04/16/2015 0.000 0.5 HV252F006A /

HV252F037A 03/16/2017 0.000 0.5 HV252F006A /

HV252F037A 03/29/2017 0.000 0.5 HV252F006A /

HV252F037A 03/26/2021 0.000 0.5

Enclosure 2 of PLA-8074 Page 16 of 16 Attachment RR02-1: Leakage History of SSES, Units 1 and 2 PIVs Unit 2 CS Injection PIVs Measured Value Leakage Valve Number Test Date Comments (gpm) Limit (gpm)

HV252F005B 03/18/2007 0.006 5 HV252F005B 04/29/2009 0.023 5 HV252F005B 04/28/2011 0.032 5 HV252F005B 05/06/2013 0.000 5 HV252F005B 04/25/2015 0.000 5 HV252F005B 03/26/2019 0.000 5 HV252F005B 04/02/2019 0.000 5 HV252F006B /

HV252F037B 03/18/2007 0.010 0.5 HV252F006B /

HV252F037B 04/29/2009 0.003 0.5 HV252F006B /

HV252F037B 04/29/2011 0.000 0.5 HV252F006B /

HV252F037B 05/06/2013 0.000 0.5 HV252F006B /

HV252F037B 04/25/2015 0.002 0.5 HV252F006B /

HV252F037B 03/26/2019 0.000 0.5

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