ML050690239

Third 10-Year Interval Inservice Testing Program Plans

ML050690239
Person / Time
Site:	Susquehanna
Issue date:	03/10/2005
From:	Richard Laufer NRC/NRR/DLPM/LPD1
To:	Shriver B Susquehanna
Guzman R, NRR/DLPM 415-1030
References
TAC MC3382, TAC MC3383, TAC MC3384, TAC MC3385, TAC MC3386, TAC MC3387, TAC MC3388, TAC MC3389, TAC MC4421, TAC MC4422
Download: ML050690239 (34)

Text

March 10, 2005 Mr. Bryce L. Shriver President-PPL Generation and Chief Nuclear Officer PPL Susquehanna, LLC Two North Ninth Street, GENTW15 Allentown, PA 18101-1179

SUBJECT:

SUSQUEHANNA STEAM ELECTRIC STATION, UNITS 1 AND 2 - THIRD 10-YEAR INTERVAL INSERVICE TESTING (IST) PROGRAM PLANS (TAC NOS. MC3382, MC3383, MC3384, MC3385, MC3386, MC3387, MC3388, MC3389, MC4421, MC4422)

Dear Mr. Shriver:

By letter dated May 28, 2004, as supplemented on September 10 and December 6, 2004, PPL Susquehanna, LLC (PPL, the licensee), submitted relief requests (RR) RR-01, RR-02, RR-03, and RR-04, for Susquehanna Steam Electric Station, Units 1 and 2 (SSES 1 and 2), proposing alternatives to the requirements of Title 10 of the Code of Federal Regulations, Part 50, Section 55a (10 CFR 50.55a), concerning the requirements of the American Society of Mechanical Engineers, Boiler and Pressure Vessel Code (ASME Code), for its third 10-year interval inservice testing (IST) program plan. In addition, by letter dated September 10, 2004, PPL submitted RR-05, proposing an alternative to the testing frequency of its high-pressure coolant injection and reactor core isolation cooling check valves for SSES 1 and 2. PPL is currently in its third 10-year IST interval which began on June 1, 2004, and will end on May 31, 2014.

In a letter dated February 7, 2005, PPL withdrew RR-04 for SSES 1 and 2 to allow time for the ASME Code Committee to propose a code case on the subject of comprehensive pump testing.

The Nuclear Regulatory Commission (NRC) staff understands that PPL will monitor the ASME Code Committees activity on this issue and may submit a relief request in the future.

The NRC staff has reviewed PPLs regulatory and technical analysis in support of its requests for relief. Based on the information provided by PPL, the NRC staff has concluded that PPLs proposed alternatives for RR-01, RR-02, RR-03, and RR-05 provide an acceptable level of quality and safety. Therefore, pursuant to 10 CFR 50.55a(a)(3)(i), the NRC staff authorizes the proposed alternatives for RR-01, RR-02, RR-03, and RR-05 for SSES 1 and 2 for the third 10-year IST interval.

B. Shriver If you have any questions, please contact the project manager, Rich Guzman, at (301)415-1030.

Sincerely,

/RA/

Richard J. Laufer, Chief, Section 1 Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-387 and 50-388

Enclosure:

Safety Evaluation cc w/encl: See next page

B. Shriver If you have any questions, please contact the project manager, Rich Guzman, at (301)415-1030.

Sincerely,

/RA/

Richard J. Laufer, Chief, Section 1 Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-387 and 50-388

Enclosure:

Safety Evaluation cc w/encl: See next page DISTRIBUTION:

PUBLIC PDI-1 RF RLaufer DTerao ACRS MOBrien GMatakas, Rgn-1 GBedi DLPM DPR RGuzman CMiller, EDO, RGN-I ACCESSION NO.: ML050690239 *SE provided. No substantive changes made.

OFFICE PDI-1/PM PDI-2/LA EMEB/SC* OGC PDI-1/SC NAME RGuzman MO'Brien DTerao TSmith RLaufer DATE 2/28/05 3/10/05 2/14/05 SE DTD 3/8/05 3/10/05 OFFICIAL RECORD COPY

Susquehanna Steam Electric Station, Units 1 and 2 cc:

Britt T. McKinney Michael H. Crowthers Vice President - Nuclear Site Operations Supervising Engineer PPL Susquehanna, LLC Nuclear Regulatory Affairs 769 Salem Blvd., NUCSB3 PPL Susquehanna, LLC Berwick, PA 18603-0467 Two North Ninth Street, GENPL4 Allentown, PA 18101-1179 Robert A. Saccone General Manager - Nuclear Operations Dale F. Roth PPL Susquehanna, LLC Manager - Quality Assurance 769 Salem Blvd., NUCSB3 PPL Susquehanna, LLC Berwick, PA 18603-0467 769 Salem Blvd., NUCSB2 Berwick, PA 18603-0467 Aloysius J. Wrape, III General Manager - Performance Luis A. Ramos Improvement and Oversight Community Relations Manager, PPL Susquehanna, LLC Susquehanna Two North Ninth Street, GENPL4 PPL Susquehanna, LLC Allentown, PA 18101-1179 634 Salem Blvd., SSO Berwick, PA 18603-0467 Terry L. Harpster General Manager - Plant Support Bryan A. Snapp, Esq PPL Susquehanna, LLC Assoc. General Counsel 769 Salem Blvd., NUCSA4 PPL Services Corporation Berwick, PA 18603-0467 Two North Ninth Street, GENTW3 Allentown, PA 18101-1179 Gregory F. Ruppert General Manager - Nuclear Engineering Supervisor - Document Control Services PPL Susquehanna, LLC PPL Susquehanna, LLC 769 Salem Blvd., NUCSB3 Two North Ninth Street, GENPL4 Berwick, PA 18603-0467 Allentown, PA 18101-1179 Rocco R. Sgarro Richard W. Osborne Manager - Nuclear Regulatory Affairs Allegheny Electric Cooperative, Inc.

PPL Susquehanna, LLC 212 Locust Street Two North Ninth Street, GENPL4 P.O. Box 1266 Allentown, PA 18101-1179 Harrisburg, PA 17108-1266 Walter E. Morrissey Director - Bureau of Radiation Protection Supervising Engineer Pennsylvania Department of Nuclear Regulatory Affairs Environmental Protection PPL Susquehanna, LLC P.O. Box 8469 769 Salem Blvd., NUCSA4 Harrisburg, PA 17105-8469 Berwick, PA 18603-0467

Susquehanna Steam Electric Station, Units 1 and 2 cc:

Senior Resident Inspector U.S. Nuclear Regulatory Commission P.O. Box 35, NUCSA4 Berwick, PA 18603-0035 Regional Administrator, Region 1 U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Board of Supervisors Salem Township P.O. Box 405 Berwick, PA 18603-0035 Dr. Judith Johnsrud National Energy Committee Sierra Club 443 Orlando Avenue State College, PA 16803

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO RELIEF REQUEST NOS. RR-01, RR-02, RR-03, AND RR-05 FOR THE INSERVICE TESTING PROGRAM PLAN FOR THE THIRD 10-YEAR INSPECTION INTERVAL PER THE AMERICAN SOCIETY OF MECHANICAL ENGINEERS BOILER AND PRESSURE VESSEL CODE, SECTION XI, REQUIREMENTS PPL SUSQUEHANNA, LLC ALLEGHENY ELECTRIC COOPERATIVE, INC.

SUSQUEHANNA STEAM ELECTRIC STATION, UNITS 1 AND 2 DOCKET NOS. 50-387 AND 50-388

1.0 INTRODUCTION

By letter dated May 28, 2004, as supplemented on September 10 and December 6, 2004, PPL Susquehanna, LLC (PPL, the licensee), submitted relief requests (RR) RR-01, RR-02, RR-03, and RR-04, for Susquehanna Steam Electric Station, Units 1 and 2 (SSES 1 and 2), proposing alternatives to the requirements of Title 10 of the Code of Federal Regulations, Part 50, Section 55a (10 CFR 50.55a), concerning the requirements of the American Society of Mechanical Engineers, Boiler and Pressure Vessel Code (ASME Code), for its third 10-year interval inservice testing (IST) program plan. In addition, by letter dated September 10, 2004, PPL submitted RR-05, proposing an alternative to the testing frequency of its high-pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) check valves for SSES 1 and

2. PPL is currently in its third 10-year IST interval which began on June 1, 2004, and will end on May 31, 2014.

In a letter dated February 7, 2005, PPL withdrew RR-04 for SSES 1 and 2 to allow time for the ASME Code Committee to propose a code case on the subject of comprehensive pump testing.

The Nuclear Regulatory Commission (NRC) staff understands that PPL will monitor the ASME Code Committee activity on this issue and may submit a relief request in the future.

2.0 REGULATORY EVALUATION

Section 50.55a requires that IST of certain ASME Code, Class 1, 2, and 3 pumps and valves be performed at 120-month (10-year) IST program intervals in accordance with the ASME Code for Operation and Maintenance of Nuclear Power Plants (OM Code) and applicable addenda, except where alternatives have been authorized or relief has been requested by the licensee and granted by the Commission pursuant to paragraphs (a)(3)(i), (a)(3)(ii), or (f)(6)(i) of 10 CFR 50.55a. In accordance with 10 CFR 50.55a(f)(4)(ii), licensees are required to comply with the requirements of the latest edition and addenda of the ASME Code incorporated by reference in the regulations 12 months prior to the start of each 120-month IST program interval. In accordance with 50.55a(f)(4)(iv), the IST of pumps and valves may meet the requirements set forth in subsequent editions and addenda that are incorporated by reference in 10 CFR 50.55a(b), subject to Commission approval.

Portions of editions or addenda may be used provided that all related requirements of the respective editions and addenda are met. In proposing alternatives or requesting relief, PPL must demonstrate that: (1) the proposed alternatives provide an acceptable level of quality and safety; (2) compliance would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety; or (3) conformance is impractical for the facility.

Section 50.55a authorizes the Commission to approve alternatives and to grant relief from ASME Code requirements upon making necessary findings.

NRC guidance contained in Generic Letter (GL) 89-04, Guidance on Developing Acceptable Inservice Testing Programs, provides alternatives to ASME Code requirements which are acceptable to the NRC staff. Further guidance is given in GL 89-04, Supplement 1, and NUREG-1482, Guidance for Inservice Testing at Nuclear Power Plants.

The SSES 1 and 2 third 10-year IST interval commenced June 1, 2004, and is scheduled to be completed by May 31, 2014. The third 10-year interval IST programs were developed to meet the requirements of the ASME OM Code 1998 edition through the 2000 addenda pursuant to 10 CFR 50.55a(f)(4)(ii).

3.0 TECHNICAL EVALUATION

PPLs regulatory and technical analyses in support of its requests for relief from the ASME OM Code IST requirements are described in PPLs letter dated May 28, 2004, as supplemented on September 10 and December 6, 2004. PPL submitted two separate sets of relief requests 1RR-01, 1RR-02, 1RR-03, 1RR-04, 1RR-05 and 2RR-01, 2RR-02, 2RR-03, 2RR-04, 2RR-05 for SSES Units 1 and 2, respectively; and on February 7, 2005, PPL withdrew 1RR-04 and 2RR-04 to allow time for the ASME Code Committee to propose a code case on the subject of comprehensive pump testing. In this safety evaluation (SE), both Unit 1 and Unit 2 relief requests have been combined as RR-01, RR-02, RR-03, and RR-05. A description of the relief request and the NRC staffs evaluation follows.

3.1 Valve Relief Request No. RR-01 ASME OM Code Requirements OM Code, paragraph ISTC-5221(c)(3), Valve Obturator Movement, states that at least one valve from each group shall be disassembled and examined at each refueling outage; all valves in each group shall be disassembled and examined at least once every 8 years.

Specific Relief Requested PPL has requested relief for the various check valves listed in groups in Table 1-1 through Table 1-23 from the ASME Code requirements of Paragraph ISTC-5221(c)(3) of the ASME OM Code 1998 edition through the 2000 addenda.

Check-Valve Groups and Check-Valve Numbers Table 1-1, Check-Valve Groups CV02 and CV03 (Unit 1)

Check-Valve Valve System Size Category Safety Group Number Inch Class CV02 011033 Emergency Service Water 8 C 3 011034 Emergency Service Water 8 C 3 011035 Emergency Service Water 8 C 3 011036 Emergency Service Water 8 C 3 011037 Emergency Service Water 8 C 3 011038 Emergency Service Water 8 C 3 011039 Emergency Service Water 8 C 3 011040 Emergency Service Water 8 C 3 CV03 011513 Emergency Service Water 10 C 3 011514 Emergency Service Water 10 C 3 These check valves are in the cooling water lines to the emergency diesel generators (EDGs).

They have an open safety function to provide emergency service water (ESW) to the EDG jacket water coolers, lube oil coolers and intercoolers. They have a close safety function to prevent backflow when the cooling is being supplied by the opposite loop of the ESW. These valves have no containment isolation function. The open and close safety functions of these valves are currently verified by valve disassembly. These valves are part stroked open during quarterly ESW flow verification (inservice pump test).

Table 1-2, Check-Valve Group CV05 (Unit 1)

Check-Valve Valve System Size Category Safety Group Number Inch Class CV05 149F011 Reactor Core Isolation Cooling 6 C 2 149F030 Reactor Core Isolation Cooling 6 C 2 These check valves are in the RCIC pump suction lines. They have an open safety function to provide a flow path for the RCIC pump while taking suction from the condensate storage tank or the suppression pool. They have a close safety function to prevent a diversion of RCIC flow when the alternate suction path is being used. These valves have no containment isolation function. The open and close safety functions of these valves are currently verified by valve disassembly.

Table 1-3: Check-Valve Group CV09 (Unit 1 and 2)

Check-Valve Valve System Size Category Safety Group Number Inch Class CV09 149F021 Reactor Core Isolation Cooling 2 C 2 249F021 Reactor Core Isolation Cooling 2 C 2 These check valves are in the RCIC pump minimum flow line. They have an open safety function to provide a minimum flow path for protection of the pump. This valve has a containment isolation function although it is not Appendix J-tested. This line terminates below the minimum suppression pool level, which provides a water seal. The open and close safety functions of these valves are currently verified by valve disassembly. These valves are part stroked open during quarterly RCIC flow verification (inservice pump test).

Table 1-4: Check-Valve Group CV24 (Unit 1 and 2)

Check-Valve Valve System Size Category Safety Group Number Inch Class CV24 150F047 Reactor Core Isolation Cooling 2 C 2 156F052 High-Pressure Coolant Injection 2 C 2 250F047 Reactor Core Isolation Cooling 2 C 2 256F052 High-Pressure Coolant Injection 2 C 2 These check valves are located in the discharge of the RCIC and High-Pressure Coolant Injection (HPCI) vacuum condenser pumps and provide the ASME Code boundary between the RCIC/HPCI pump suction and the discharge of the vacuum tank condenser pump. They have a

close safety function to maintain RCIC/HPCI water inventory in the event of a line break of the non-ASME Code piping. Per paragraph ISTC-5221(a)(3), these check valves are also required to be verified to partially open. These valves have no containment isolation function. The close safety function of these valves is currently verified by valve disassembly.

Table 1-5: Check-Valve Group CV08 (Unit 1)

Check-Valve Valve System Size Category Safety Group Number Inch Class CV08 151F046A Residual Heat Removal 4 C 2 151F046B Residual Heat Removal 4 C 2 151F046C Residual Heat Removal 4 C 2 151F046D Residual Heat Removal 4 C 2 These check valves are in residual heat removal (RHR) pump minimum flow lines. They have an open safety function to provide a minimum flow path for pump operation. Per paragraph ISTC-5221(a)(2), these check valves are also required to be verified for closure. These valves have no containment isolation function. The open safety function of these valves is currently verified by valve disassembly. These valves are part stroked open during quarterly RHR flow verification (inservice pump test).

Table 1-6: Check-Valve Group CV10 (Unit 1 and 2)

Check-Valve Valve System Size Category Safety Group Number Inch Class CV10 152005 Core Spray 3 C 2 252005 Core Spray 3 C 2 These check valves are located in the suppression pool fill lines. These valves have a safety function to close if the line is being used for filling the suppression pool (manual upstream valve 152028 open) to maintain core spray (CS) water inventory. These valves have no containment isolation function. The close safety function of this valve is currently verified by valve disassembly.

Table 1-7: Check-Valve Group CV22 (Unit 1)

Check-Valve Valve System Size Category Safety Group Number Inch Class CV22 152F029A Core Spray 2 C 2 152F029B Core Spray 2 C 2 152F030A Core Spray 2 C 2 152F030B Core Spray 2 C 2

These check valves are in the keep fill lines for the CS system. They have a close safety function to prevent loss of inventory during CS system operation. Per paragraph ISTC- 5221(a)(3), these check valves are also required to be verified to partially open. These valves have no containment isolation function. The close safety function of these valves is currently verified by valve disassembly. The open function of these valves is continually verified during plant operation by proper operation of the keep fill system.

Table 1-8: Check-Valve Group CV11 (Unit 1)

Check-Valve Valve System Size Category Safety Group Number Inch Class CV11 152F036A Core Spray 3 C 2 152F036B Core Spray 3 C 2 152F036C Core Spray 3 C 2 152F036D Core Spray 3 C 2 These check valves are in the CS pump minimum flow lines. They have an open safety function to provide a minimum flow path for pump protection. Per paragraph ISTC-5221(a)(2),

these check valves are also required to be verified for closure. These valves have no containment isolation function. The open safety function of these valves is currently verified by valve disassembly. These valves are part stroked open during quarterly CS flow verification (inservice pump test).

Table 1-9: Check-Valve Group CV14 (Unit 1 and 2)

Check-Valve Valve System Size Category Safety Group Number Inch Class CV14 153071A Fuel Pool Cooling & Cleanup 8 C 3 153071B Fuel Pool Cooling & Cleanup 8 C 3 253071A Fuel Pool Cooling & Cleanup 8 C 2 253071B Fuel Pool Cooling & Cleanup 8 C 2 These check valves are in the alternate flow path to the fuel storage pool. They have an open safety function to provide fuel storage pool cooling. Under paragraph ISTC-5221(a)(2), it is required that these check valves also be verified for closure. These valves have no containment isolation function. The open safety functions of these valves are currently verified by valve disassembly. These valves are part stroked open during periodic pressure testing as required by the ASME Code,Section XI, inservice inspection requirements.

Table 1-10: Check-Valve Group 06 (Unit 1)

Check-Valve Valve System Size Category Safety Group Number Inch Class CV06 155F019 High-Pressure Coolant Injection 16 C 2 155F045 High-Pressure Coolant Injection 16 C 2 These check valves are in the HPCI pump suction lines. They have an open safety function to provide a flow path for the HPCI pump while taking suction from the condensate storage tank or the suppression pool. They have a close safety function to prevent a diversion of HPCI flow when the alternate suction path is being used. These valves have no containment isolation function. The open and close safety functions of these valves are currently verified by valve disassembly.

Table 1-11: Check-Valve Group 13 (Unit 1 and 2)

Check-Valve Valve System Size Category Safety Group Number Inch Class CV13 155F046 High-Pressure Coolant Injection 4 C 2 255F046 High-Pressure Coolant Injection 4 C 2 These check valves are in the HPCI pump minimum flow lines. They have an open safety function to provide a minimum flow path for protection of the pump. These valves have a containment isolation function although they are not Appendix J-tested. These lines terminate below the minimum suppression pool level, which provides a water seal. The open and close safety functions of these valves are currently verified by valve disassembly. This valve is part stroked open during the quarterly HPCI flow verification (inservice pump test).

Table 1-12: Check-Valve Group CV25 (Unit 1)

Check-Valve Valve System Size Category Safety Group Number Inch Class CV25 011193A Emergency Service Water 1 C 2 011193B Emergency Service Water 1 C 2 012807A RHR Service Water 1 C 2 012807B RHR Service Water 1 C 2 These check valves are in the ESW and RHR service water biocide injection lines, and provide the ASME boundary between ESW and RHR service water biocide injection line. They have a close safety function to provide and maintain ESW and RHR service water inventory in the event of a line break in the non-ASME Code piping. Per paragraph ISTC-5221(a)(3), these

valves are also required to be verified to partially open. These valves have no containment isolation function. The close safety function of these valves had been verified by checking a telltale drain valve upstream of the check valves.

Table 1-13: Check-Valve Group CV12 (Unit 1)

Check-Valve Valve System Size Category Safety Group Number Inch Class CV12 149F063 Reactor Core Isolation Cooling 3 C 2 149F064 Reactor Core Isolation Cooling 3 C 2 155F076 High-Pressure Coolant Injection 3 C 2 155F077 High-Pressure Coolant Injection 3 C 2 These check valves are in RCIC and HPCI turbine exhaust lines. They have an open safety function to prevent a vacuum relief path for the turbine exhaust line. They have a close safety function to prevent steam flow into the suppression chamber. These valves have no containment isolation function. The open and close safety functions of these valves are currently verified by valve disassembly.

Table 1-14: Check-Valve Group CV04 (Unit 1 and 2)

Check-Valve Valve System Size Category Safety Group Number Inch Class CV04 149016 Reactor Core Isolation Cooling 2 C 2 155013 High-Pressure Coolant Injection 2 C 2 249016 Reactor Core Isolation Cooling 2 C 2 255013 High-Pressure Coolant Injection 2 C 2 These check valves are located in the keep fill lines for the RCIC and HPCI systems. They have a close safety function to prevent loss of inventory during RCIC and HPCI system operation. Per paragraph ISTC-5221(a)(3), these check valves are also required to be verified to partially open. These valves have no containment function. The close functions of these valves are currently verified by valve disassembly. The open function of these valves is continually verified during plant operation by proper operation of the keep fill system.

Table 1-15: Check-Valve Group CV07 (Unit 1 and 2)

Check-Valve Valve System Size Category Safety Group Number Inch Class CV07 156F048 High-Pressure Coolant Injection 2 C 2

156F057 High-Pressure Coolant Injection 2 C 2 256F048 High-Pressure Coolant Injection 2 C 2 256F057 High-Pressure Coolant Injection 2 C 2 These check valves are located on the outlet of the HPCI Lube Oil Cooler and in the HPCI Lube Oil Cooler return line to the HPCI Booster pump. They have an open function to provide a flow path for cooling water from the turbine lube oil cooler. Per paragraph ISTC-5221(a)(2), these check valves are also required to be verified for closure. These valves have no containment isolation function. The open safety function of these valves is currently verified by valve disassembly.

Table 1-16: Check-Valve Group CV17 (Unit 2)

Check-Valve Valve System Size Category Safety Group Number Inch Class CV17 249F011 Reactor Core Isolation Cooling 6 C 2 249F030 Reactor Core Isolation Cooling 6 C 2 These check valves are in RCIC pump suction lines. They have an open safety function to provide a flow path for the RCIC pump while taking suction from the condensate storage tank or the suppression pool. They have a close safety function to prevent a diversion of RCIC flow when the alternate suction path is being used. These valves have no containment isolation function. The open and close safety functions of these valves are currently verified by valve disassembly.

Table 1-17: Check-Valve Group CV15 (Unit 2)

Check-Valve Valve System Size Category Safety Group Number Inch Class CV15 251F046A Residual Heat Removal 4 C 2 251F046B Residual Heat Removal 4 C 2 251F046C Residual Heat Removal 4 C 2 251F046D Residual Heat Removal 4 C 2 These check valves are in RHR pump minimum flow lines. They have an open safety function to provide a minimum flow path for pump operation. Per paragraph ISTC-5221(a)(2), these check valves are also required be verified for closure. These valves have no containment isolation function. The open and close safety functions of these valves are currently verified by valve disassembly. These valves are part stroked open during quarterly RHR flow verification (inservice pump test).

Table 1-18: Check-Valve Group CV23 (Unit 2)

Check-Valve Valve System Size Category Safety Group Number Inch Class CV23 252F029A Core Spray 2 C 2 252F029B Core Spray 2 C 2 252F030A Core Spray 2 C 2 252F030B Core Spray 2 C 2 These check valves are in the keep fill lines for the CS system. They have a close safety function to prevent loss of inventory during CS system operation. Per paragraph ISTC- 5221(a)(3), these check valves are also required to be verified to partially open. These valves have no containment isolation function. The close safety function of these valves is currently verified by valve disassembly. The open function of these valves is continually verified during plant operation by proper operation of the keep fill system.

Table 1-19: Check-Valve Group CV16, (Unit 2)

Check-Valve Valve System Size Category Safety Group Number Class CV16 252F036A Core Spray 3 C 2 252F036B Core Spray 3 C 2 252F036C Core Spray 3 C 2 252F036D Core Spray 3 C 2 These check valves are in the CS pump minimum flow lines. They have an open safety function to provide a minimum flow path for pump protection. Per Paragraph ISTC-5221(a)(2),

these check valves are also required to be verified for closure. These valves have no containment isolation function. The open safety functions of these valves are currently verified by valve disassembly. These valves are part stroked open during quarterly CS flow verification (inservice pump test).

Table 1-20: Check-Valve Group CV19 (Unit 2)

Check-Valve Valve System Size Category Safety Group Number Class CV19 255F019 High-Pressure Coolant Injection 16 C 2 255F045 High-Pressure Coolant Injection 16 C 2

These check valves are in the HPCI pump suction lines. They have an open safety function to provide a flow path for the HPCI pump while taking suction from the condensate storage tank or the suppression pool. They have a close safety function to prevent a diversion of HPCI flow when the alternate suction path is being used. These valves have no containment isolation function. The open and close safety functions of these valves are currently verified by valve disassembly.

Table-1-21: Check-Valve Group CV20 (Unit 2)

Check-Valve Valve System Size Category Safety Group Number Class CV20 211132 Emergency Service Water 4 C 3 211134 Emergency Service Water 4 C 3 Check valve 211132 is in the ESW supply line to the emergency switchgear and load center room A cooler. Check valve 211134 is in the ESW supply line to the emergency switchgear and load center room B cooler. The open safety functions of these valves are currently verified by valve disassembly.

Table-1-22: Check-Valve Group CV21 (Unit 2)

Check-Valve Valve System Size Category Safety Group Number Class CV21 211133 Emergency Service Water 4 C 3 211135 Emergency Service Water 4 C 3 Check valve 211133 is in the ESW return line from the emergency switchgear and load center room A cooler. Check valve 211135 is in the ESW return line from the emergency switchgear and load center room B cooler. The open safety functions of these valves are currently verified by valve disassembly.

Table 1-23: Check-Valve Group CV18 (Unit 2)

Check-Valve Valve System Size Category Safety Group Number Class CV18 249F063 Reactor Core Isolation Cooling 3 C 2 249F064 Reactor Core Isolation Cooling 3 C 2 255F076 High-Pressure Coolant Injection 3 C 2 255F077 High-Pressure Coolant Injection 3 C 2

These check valves are in RCIC and HPCI turbine exhaust lines. They have an open safety function to prevent a vacuum relief path for the turbine exhaust line. They have close safety function to prevent steam flow into the suppression chamber. These valves have no containment isolation function. The open and close safety functions of these valves are currently verified by valve disassembly.

PPLs Basis for Relief PPL states that the components listed above are check valves with no external means for exercising and no external position indication. Due to a lack of installed flow or pressure indication and a lack of test connections, it is not possible to use other means to verify the open and/or close exercising of these check valves. Disassembly of the valves is the most feasible method to verify operability and can be accomplished during system outages, which may be conducted on-line (during power operation). The check valves have been grouped by valve manufacturer, design, service, size, materials of construction, and orientation as required by ASME OM Code 1998 through 2000 Addenda, paragraph ISTC-5221(c)(1).

Prior to performing a system outage on-line, its effect on the risk is evaluated in accordance with the requirements of 10 CFR 50.65, Requirement for monitoring the effectiveness of maintenance at nuclear power plant. This requirement, 50.65(a)(4), states in part that, Before performing maintenance activities (including but not limited to surveillance, post-maintenance testing, and corrective and preventive maintenance), the licensee shall assess and manage the increase in risk that may result from the proposed maintenance activities.

PPL complies with the requirements of 10 CFR 50.65(a)(4) via application of a program governing maintenance scheduling. The program dictates the requirements for risk evaluations as well as the necessary levels of action required for risk management in each case. The program also controls operation of the on-line risk monitor system, which is based on probabilistic risk assessment (PRA). With the use of risk evaluation for various aspects of plant operations, PPL has initiated efforts to perform additional maintenance, surveillance, and testing activities during normal operation. Planned activities are evaluated to maintain associated safety margins. Individual system components, a system train, or a complete system may be planned to be out of service to allow maintenance, or other activities, during normal operation.

Disassembly and inspection may involve a system breach. However, the valves are isolated and associated sections of piping drained during disassembly. Thus, the system breach does not increase the risk due to an internal flooding or internal system loss-of-coolant accident (LOCA). The risk associated with these activities would be bounded by the risk experienced due to the system outage. Therefore, PPL believes that disassembly and testing of these valves during a scheduled system outage while on-line would have no additional impact on core damage frequency (CDF).

As more system outages are performed on-line, it is evident that selected refueling outage inservice testing activities, e. g. valve exercising and disassembly, could be performed during these system outage windows without sacrificing the level of quality or safety. IST performed on a refueling outage frequency is currently acceptable in accordance with the ASME OM Code 1998 edition through the 2000 addenda. By specifying testing activities on a frequency commensurate with each refueling outage, the ASME OM Code 1998 edition through the 2000

addenda, establishes an acceptable time period between testing. Historically, the refueling outage has provided a convenient and defined time period in which testing activities could be safely and efficiently performed. However, an acceptable testing frequency can be maintained separately without being tied directly to a refueling outage. IST performed on a frequency that maintains the acceptable time period between testing activities during the operating cycle is consistent with the intent of ASME OM Code 1998 edition through the 2000 addenda.

Over time, approximately the same number of tests will be performed using the proposed operating cycle frequency as would be performed using the current refueling outage frequency.

Thus, IST activities performed during the proposed operating cycle test frequency provide an equivalent level of quality and safety.

PPLs Proposed Alternative Testing PPL proposes an alternative testing frequency for performing IST of the valves identified above.

At least one valve from each group will be tested on a frequency of once each operating cycle in lieu of once each refueling outage as currently required by the ASME OM Code 1998 edition through the 2000 addenda, paragraph ISTC-5221(c)(3), Valve Obturator Movement. All valves in each group will be tested at least once every 8 years as required by the ASME OM Code 1998 edition through the 2000 addenda, paragraph ISTC-5221(c)(3).

Check-valve groups CV04, CV07, CV09, CV10, CV13, CV14, and CV24 include identical Unit 1 and Unit 2 valves. Check-valve groups CV01, CV03, CV05, CV06, CV08, CV11, CV12, CV22, CV25 only include valves of Unit 1, and check-valve groups CV15, CV16, CV17, CV18, CV19, CV20, CV21, and CV23 only include valves of Unit 2. One check valve from each group will be disassembled and inspected during an operating cycle every 24 months, and all check valves in the group will be disassembled and inspected at least once every 8 years.

NRC Staffs Evaluation of Relief Request No. RR-01 OM Code paragraph ISTC-5221(c)(3), Valve Obturator Movement, states, in part, that at least one valve from each group shall be disassembled and examined at each refueling outage; all valves in each group shall be disassembled and examined at least once every 8 years.

PPL proposes an alternative testing frequency for performing IST of the valves identified in the groups in Tables 1-1 through 1-23, with at least one valve from each group being tested on a frequency of once each operating cycle, i.e., 24 months, in lieu of once each refueling outage as currently required by the ASME OM Code 1998 edition through the 2000 addenda, paragraph ISTC-5221(c)(3). All valves in each group will be tested at least once every 8 years as required by the ASME OM Code 1998 edition through the 2000 addenda, paragraph ISTC-5221(c)(3).

PPL states that all the check valves listed above have no external means for exercising and no external position indication. Due to a lack of installed flow or pressure indication and a lack of test connections, it is not possible to use other means to verify the open and/or close exercising of these check valves. Disassembly of the valves is the most feasible method to verify operability and can be accomplished during system outages which will occur while the plant is on-line.

PPL proposes, as an alternative, to perform IST disassembly and inspection activities during normal plant operation, in conjunction with appropriate system outages, or during refueling outages. In any case, disassembly, inspection, and manual exercising will be performed at least once each operating cycle, i.e., 24 months. Check valve disassembly and inspection testing during normal plant operation will be managed in accordance with the requirements of 10 CFR 50.65(a)(4) in conjunction with system isolation as described above.

The NRC staff finds that disassembly and inspection of system check valves are the appropriate methods to verify operability and can be accomplished during system outages when the plant is on-line or during refueling outages. The NRC staffs finding is based on the following considerations:

! PPL states that all check valves listed in Tables 1-1 through 1-23 above have been grouped by valve manufacturer, design, service, size, materials of construction, and orientation as required by paragraph ISTC-5221(c)(1) of ASME OM Code 1998 through 2000 Addenda.

! Some of the check valves of SSES 1 and 2 that meet the grouping criteria of the OM Code are grouped accordingly. This is consistent with the guidelines of NUREG-1482, Section 4.1.

! ASME OM Code, paragraph ISTC-5221(c)(3), states, At least one valve from each group shall be disassembled and examined at each refueling outage, and all valves in each group shall be disassembled and examined at least once every 8 years. By specifying testing activities on a frequency commensurate with each refueling outage, the ASME OM Code recognizes and establishes an acceptable time period between testing. The refueling outages have provided a practical and definitive time period in which testing activities can be safely and effectively performed. An acceptable testing frequency can be maintained separately without being tied directly to a refueling outage.

IST performed on a frequency of 24 months that maintains the acceptable time period between testing activities during the operating cycle, i.e., 24 months, is consistent with the intent of the ASME OM Code.

! Over time, approximately the same number of tests will be performed using the proposed operating-cycle test frequency as would be performed using the current refueling-outage frequency. Thus, IST activities performed during the proposed operating cycle, a 24-month test frequency, provide an equivalent level of quality and safety as IST performed at a refueling outage frequency.

! PPL states that the on-line check valves inspections will be performed by system outage and by system isolation. The valves used to provide the isolation boundary for the disassembly and inspection of the check valves have excellent history of providing adequate isolation. Also, all activities, including check valves disassembly and inspection, would not proceed if adequate isolation could not be established and maintained. Once adequate isolation is confirmed, it is maintained by passive isolation valves or valves made passive, e. g. , de-energized motor-operated valves, that are controlled in accordance with SSES 1 and 2 energy control process.

! PPL states that a typical on-line system-outage window lasts from 72 to 96 hours4 days <br />0.571 weeks <br />0.132 months <br />, < 50 percent, of the allowable limiting condition for operation (LCO) time limit. Based on a review of maintenance history, disassembly, inspection, and reassembly of check valves, completion of check-valve disassembly and inspection takes between 6 and 24 hours1 days <br />0.143 weeks <br />0.0329 months <br />, depending upon size and location of the valve. The IST activity would be conducted simultaneously with other maintenance activities scoped into the system-outage window. This provides adequate margin to complete disassembly and inspection activities in an orderly manner.

! PPL states that performing IST activity on-line would not change the duration of the on-line system-outage window or the core damage probability (CDP) associated with the existing on-line activities. Therefore, the risk/CDP over the entire operating/shutdown spectrum would remain unchanged.

! PPL is using equipment out of service (EOOS) software to evaluate plant configuration risk when the plant is in Modes 1, 2, or 3. The EOOS processes the activities through a SSES 1 and 2 specific model and calculates overall effects on nuclear safety. EOOS determines that the increase in CDF and large early release frequency (LERF), which provides the risk impact in performing disassembly and inspection of these check valves while the plant is on-line, is insignificant.

! There are no technical barriers to performing these IST activities during either the refueling outage or the operating cycle.

! PPL states that the potential risk impact in performing disassembly and inspection of these check valves while the plant is on-line is insignificant.

On the basis of these considerations, the NRC staff finds that the proposed alternative provides an acceptable level of quality and safety.

Conclusion Based on its review of the information provided for relief request RR-01, the NRC staff finds that PPLs proposed alternative provides an acceptable level of quality and safety. Therefore, the NRC staff concludes that the proposed alternative to disassemble and inspect the check valves once every operating cycle, in lieu of once during each refueling outage, is authorized pursuant to 10 CFR 50.55a(a)(3)(i).

3.2 Valve Relief Request No. RR-02 ASME OM Code Requirements ASME OM Code Appendix I, paragraph I-1330(a), Test Frequencies, Class 1 Pressure Relief Valves, requires that Class 1 pressure relief valves shall be tested at least once every 5 years, starting with initial electric power generation. No maximum limit is specified for the number of valves to be tested within each interval; however, a minimum of 20% of the valves from each valve group shall be tested within any 24-month interval. This 20% shall consist of valves that have not been tested during the current 5-year interval, if they exist. The test interval for any individual valve shall not exceed 5 years.

Specific Relief Requested PPL has requested relief for several main steam safety/relief valves (MSRVs) listed in Table 2-1 from the OM Code requirements of paragraph I-1330(a).

Table 2-1: Main Steam Safety/Relief Valves Valve System Category Class PSV141F013A Nuclear Boiler C 1 PSV141F013B Nuclear Boiler C 1 PSV141F013C Nuclear Boiler C 1 PSV141F013D Nuclear Boiler C 1 PSV141F013E Nuclear Boiler C 1 PSV141F013F Nuclear Boiler C 1 PSV141F013G Nuclear Boiler C 1 PSV141F013H Nuclear Boiler C 1 PSV141F013J Nuclear Boiler C 1 PSV141F013K Nuclear Boiler C 1 PSV141F013L Nuclear Boiler C 1 PSV141F013M Nuclear Boiler C 1 PSV141F013N Nuclear Boiler C 1 PSV141F013P Nuclear Boiler C 1 PSV141F013R Nuclear Boiler C 1 PSV141F013S Nuclear Boiler C 1 These valves are main steam safety/relief valves. They provide overpressure protection for the reactor coolant pressure boundary to prevent unacceptable radioactive release and exposure to plant personnel.

PPLs Basis for Relief Pursuant to 10 CFR 50.55a(a)(3), PPL requested authorization of an alternative to the requirements of ASME OM Code, Appendix I, paragraph I-1330(a), on the basis that the proposed alternative would provide an acceptable level of quality and safety.

During the second 10-year IST interval, PPL removed and tested 8 of the 16 MSRVs during each refueling outage. This methodology meets the ASME Code criteria of testing previously untested valves and permits the removal and replacement of weeping valves detected during the previous operating cycle. Weeping MSRVs are detected by monitoring tailpipe temperatures. If the tailpipe temperature exceeds 200 degrees Fahrenheit, then the relief valve is viewed as a weeper.

Without Code relief for 24-month fuel cycles, strict Code compliance would restrict SSES 1 and 2's operating philosophy to not operate with weeping MSRVs as ASME Code testing would be required to be completed within 5 years. This testing strategy does not account for any leaking valves that may need to be refurbished. Since PPLs philosophy is to share spare valves between both units, (the valves that are removed from one unit are installed in the other units next refueling outage), this testing strategy is less than adequate. This strategy could only be

accomplished if a large population of MSRVs are tested each outage or additional spare valves are purchased. More than 8 valves would need to be sent to the offsite testing facility during a refueling outage. The testing and return of these valves would have to be completed expeditiously in order to not impact the refueling outage schedule duration. For this reason, additional expenditures would be incurred to purchase and test a greater number of valves each outage. Without Code relief, the additional outage work would be contrary to the principles of as low as reasonably achievable (ALARA) and could compromise radiation safety.

Because of the location of certain MSRVs in the containment, interferences exist that would require the removal of more valves and piping for those valves that must be removed for the sample testing. This results in more radiation exposure to the maintenance personnel than is desirable.

With Code relief, the 16 MSRVs per unit can be tested within 6 years to complete the ASME Code-required testing for the total population and accommodate any weeping MSRVs. The increased testing over only 2 refuel cycles would result in no additional safety benefit to the plant. SSES 1 and 2 has had excellent performance with MSRVs over the last 10 years. Since 1987, SSES 1 and 2 has imposed a more conservative, as-left leakage criterion on the testing facility than was specified in the General Electric (GE) Specification and incorporated in the PPL Specification for testing Crosby-style relief valves. The criterion imposed on the test lab is 0 ml/5 minutes (via the purchase order) compared to a GE Specification as-left leakage criterion of 38 ml/5 minutes.

Additionally, a review of the setpoint testing results (for both units) for the time period from initial operation to March 2004 (which comprises 255 data points) shows that the average of the setpoint drift percentages is -0.705%. This indicates that, in general, the MSRVs tend to drift slightly downward; not upward. The calculated standard deviation from the average for the data was determined to be 1.43%.

Also, the testing history shows that since commercial operation, SSES 1 and 2 has had only two as-found set pressure test acceptance criteria failures (above +3%) of the tested valves, which required additional MSRVs to be tested.

PPLs Proposed Alternative Testing For the third 10-year interval, PPL proposes to remove at least 20% of the 16 MSRVs plus any weeping valves detected during the previous operating cycle and any valves required to be removed to access scheduled or weeping valves up to a maximum of 8 valves during each refueling outage.

Additional valves above the ASME Code-required minimum 20% will be tested if the as-found setpoint exceeds +3% of the setpoint pressure on the nameplate. No additional valves will be tested if the as-found setpoint is below the nameplate setpoint. The additional valves tested will be from the initial population removed that are in excess of the 20% ASME Code-required minimum. If one of these valves fails, then all MSRVs would be removed and tested.

Completion of ASME Code testing will be accomplished over a period of 3 refuel cycles or 6 years. This approach results in maintenance and operational flexibility with the following benefits:

! provides the ability to both test the ASME Code-required valves out of the population not yet tested and replace any weeping MSRVs

! maintains relatively leak-free MSRVs, thus minimizing the necessary run time of emergency core cooling systems that provide suppression pool cooling

! ensures consistent application of ALARA principles

! enhances equipment reliability

! results in minimal impact on outage durations The MSRVs will be tested such that a minimum of 20% of the valves (previously untested, if they exist) are tested every 24 months, such that all the valves will be tested within 3 refueling cycles.

This proposal utilizes the same maintenance and testing approach that was applied in 18-month refueling cycles. This alternative frequency will continue to provide assurance of the valve operational readiness, and, thus, provides an acceptable level of quality and safety.

Additionally, any failures, either seat leakage or pressure setpoint, occurring at the test facility, as well as weeping MSRVs that develop during the operating cycle will be documented by the corrective action program, evaluated, and dispositioned accordingly.

NRC Staffs Evaluation of Relief Request No. RR-02 ASME OM Code, Appendix I, paragraph I-1330(a), Test Frequencies, Class 1 Pressure Relief Valves, requires that Class 1 pressure relief valves shall be tested at least once every 5 years, starting with initial electric power generation. The ASME OM Code also requires that a minimum of 20% of the valves from each valve group shall be tested within any 24-month interval. This 20% population shall consist of valves that have not been tested during the current 5-year interval, if they exist. The test interval for any individual valve shall not exceed 5 years.

PPL proposes to test the MSRVs such that a minimum of 20% of the valves (previously untested, if they exist) will be tested every 24 months and that all valves will be tested within 3 refueling cycles (6 years). Similar relief requests have been authorized for SSES 1 and 2 in the NRC staffs SEs dated April 7, 1998, and December 16, 1998. Later, revised relief requests for SSES 1 and 2 resulting from changes to the MSRVs setpoint tolerance from +/-1% to +/-3% as specified in the technical specifications (TSs) was authorized by the NRC on March 6, 2002 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML020560602). The previously approved relief requests were based on the ASME Code,Section XI, 1989 edition, whereas, this relief request is based on the ASME OM Code 1998 edition through the 2000 addenda.

The NRC staff has reviewed PPLs safety relief valve (SRV) test results to determine if PPLs proposed alternative testing provides an acceptable level of quality and safety. The SRV setpoint test results for the time period from initial operation to March 2004 included 255 data points. The average of the setpoint drift percentages is -0.705% which indicates that in general, the SRV setpoints tend to drift slightly downwards; not upwards. From an

overpressure protection standpoint, a setpoint drift in the downwards direction is conservative because the valve would tend to open sooner than required. PPL also calculated the standard deviation from the average for the data set and determined it to be 1.43%. The TS value of MSRV tolerances is +/-3%. Therefore, all the drift values will be within the tolerance of the TS value. PPL states that the testing history shows that, since commercial operation, there were only two as found valves above the +3% tolerance, which requires testing of additional MSRVs.

Also, the MSRVs will be tested such that a minimum of 20% of the valves (previously untested if they exist) are tested every 24 months such that all the valves will be tested within 3 refueling cycles (6 years). The additional time beyond that required by the ASME Code will not impair the valves operational readiness based on the past performance of these valves.

In addition, additional valves above the ASME Code-required minimum 20% will be tested if the as-found setpoint exceeds +3% of the nameplate setpoint. The additional valves tested will be selected from the initial population removed that are in excess of the 20% ASME Code-required minimum. If one of these valves fail, then all the MSRVs would be removed and tested.

Conclusion Based on a review of the information provided by PPL as discussed above, the NRC staff concludes that PPLs proposed alternative as specified in relief request RR-02, is authorized pursuant to 10 CFR 50.55a(a)(3)(i) on the basis that the alternative provides an acceptable level of quality and safety.

3.3 Valve Relief Request No. RR-03 ASME OM Code Requirements Paragraph ISTC-3510, Exercising Test Frequency, requires that Active Category A, Category B, and Category C check valves shall be exercised nominally every 3 months, except as provided by paragraphs ISTC-3520, ISTC-3540, ISTC-3550, ISTC-3560, ISTC-5221, and ISTC-5222.

Paragraph ISTC-3522(c), Category C Check Valves, requires that if exercising is not practicable during operation at power and cold shutdown, it shall be performed during refueling outages.

Paragraph ISTC-3700, Position Verification Testing, requires that valves with remote position indicators shall be observed locally at least once every 2 years to verify that valve operation is accurately indicated.

Specific Relief Requested PPL requested relief from paragraphs ISTC 3522(c) and ISTC-3700, and proposed reducing the number of excess flow check valves (EFCVs) tested every refueling outage from each to a representative sample every refueling outage (nominally once every 24 months) for EFCVs listed in Tables 3-1 and 3-2. The representative sample is based on approximately 20% of the valves each 2-year cycle such that each valve is tested at least every 10 years (nominal).

Table 3-1 Valve Number Valve Number System OM Safety Unit 1 Unit 2 Category Class XV141F009 XV241F009 Nuclear Boiler C 1 XV141F070A/B/C/D XV241F070A/B/C/D Nuclear Boiler C 1 XV141F071A/B/C/D XV241F071A/B/C/D Nuclear Boiler C 1 XV141F072A/B/C/D XV241F072A/B/C/D Nuclear Boiler C 1 XV141F073A/B/C/D XV241F073A/B/C/D Recirc. Pump C 1 XV14201 XV24201 Nuclear Boiler C 1 XV14202 XV24202 Nuclear Boiler C 1 XV142F041 XV242F041 Nuclear Boiler C 1 XV142F061 XV242F061 Nuclear Boiler C 1 XV142F043A/B XV242F043A/B Nuclear Boiler C 1 XV142F045A/B XV242F045A/B Nuclear Boiler C 1 XV142F047A/B XV242F047A/B Nuclear Boiler C 1 XV142F051A/B/C/D XV242F051A/B/C/D Nuclear Boiler C 1 Table 3-2 Valve Number Valve Number System OM Safety Unit 1 Unit 2 Category Class XV142F053A/B/C/D XV242F053A/B/C/D Nuclear Boiler C 1 XV142F055 XV242F055 Nuclear Boiler C 1 XV142F057 XV242F057 Nuclear Boiler C 1 XV142F059A/B/C/D/ XV242F059A/B/C/D/ Nuclear Boiler C 1 E/F/G/H E/F/G/H XV142F059L/M/N/P/ XV242F059L/M/N/P/ Nuclear Boiler C 1 R/S/T/U R/S/T/U XV143F003A/B XV243F003A/B Reactor C 1 Recirculation XV143F004A/B XV243F004A/B Reactor C 1 Recirculation XV143F009A/B/C/D XV243F009A/B/C/D Reactor C 1 Recirculation

XV143F010A/B/C/D XV243F010A/B/C/D Reactor C 1 Recirculation XV143F011A/B/C/D XV243F011A/B/C/D Reactor C 1 Recirculation XV143F012A/B/C/D XV243F012A/B/C/D Reactor 1 Recirculation C XV143F040A/B/C/D XV243F040A/B/C/D Reactor C 1 Recirculation XV143F057A/B XV243F057A/B Reactor C 1 Recirculation XV14411A/B/C/D XV24411A/B/C/D Reactor Water C 1 Cleanup XV144F046 XV244F046 Reactor Water C 1 Cleanup XV149F044A/B/C/D XV249F044A/B/C/D Reactor Core C 1 Isolation Cooling XV155F024A/B/C/D XV255F024A/B/C/D High-Pressure C 1 Coolant Injection XV15109A/B/C/D XV25109A/B/C/D Residual Heat C 1 Removal XV152F018A/B XV252F018A/B Core Spray C 1 These valves are instrumentation-line EFCVs provided in each instrument line process line that penetrates primary containment in accordance with Regulatory Guide (RG) 1.11. The EFCVs are designed to close upon rupture of the instrument line downstream of the EFCVs and otherwise remain open. The lines are sized and/or orificed such that off-site dose will be substantially below 10 CFR Part 100 limits in the event of a rupture.

PPLs Basis for Relief Pursuant to 10 CFR 50.55a(a)(3), PPL requested authorization of an alternative to the requirements of ASME OM Code, paragraphs ISTC-3522(c) and ISTC-3700. The basis of the relief request is that the proposed alternative would provide an acceptable level of quality and safety.

Testing the subject valves quarterly or during cold shutdown is not practicable, based on plant conditions. These valves have been successfully tested throughout the life of SSES 1 and 2, and they have shown no degradation or other signs of aging.

The technology for testing these valves is simple and has been demonstrated effectively during the operating history of SSES 1 and 2. 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 Code-required testing.

EFCVs are required to be tested in accordance with paragraph ISTC-3522, which requires exercising check valves nominally every 3 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 verification at least once every 2 years.

The EFCVs are classified as ASME Code, Category C and are also containment isolation valves. However, these valves are excluded from 10 CFR Part 50, Appendix J, Section III.C.,

leakage rate testing due to the size of the instrument lines and upstream orificing. Therefore, they have no safety-related seat leakage criterion.

The EFCV is a simple device. The major components are a 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.

Functional testing is required by TS Surveillance Requirement (SR) 3.6.1.3.9. System design does not include test taps upstream of the EFCVs. For this reason, the EFCVs cannot be isolated and tested using a pressure source other than reactor pressure.

The testing described above requires removal of the associated instrument or instruments 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 also will contribute to an increase in personnel radiation exposure.

Industry experience, as documented in Topical Report NEDO-32977-A, Excess Flow Check Valve Testing Relaxation, dated June 2000, indicates the ECFVs have a very low failure rate.

At SSES 1 and 2, the failure rate has been approximately 1%. Only half of these failures have resulted in replacement of the EFCVs. The SSES 1 and 2 test history shows no evidence of common mode failure. This SSES 1 and 2 test experience is consistent with the findings of NEDO-32977-A. NEDO-32977-A indicates similarly that many reported test failures at other plants were related to test methodologies and not actual EFCV failures. Thus, the ECFVs at SSES 1 and 2, consistent with the industry, have exhibited a high degree of reliability and availability, and provide an acceptable level of quality and safety.

Testing on a cold shutdown frequency is impractical considering the large number of valves to be tested and the condition that reactor pressure greater than 500 psig is needed for testing. In this instance, considering the number of valves to be tested and the conditions required for testing, it is also a hardship to test all these valves during refueling outages. Recent improvements in refueling-outage schedules minimized the time that is planned for refueling and testing activities during the outages.

The appropriate time for performing EFCV testing is during refueling outages in conjunction with the vessel hydrostatic testing. As a result of shorted 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.

PPLs Proposed Alternative Examination As an alternative to testing all EFCVs during the refueling outage, a sampling plan will be implemented. This plan will test certain EFCVs immediately preceding the refueling outage while the reactor is at power, while also instituting the appropriate conditions for testing (reactor pressure greater than 500 psig). Performance of this EFCV testing prior to the outage will be scheduled such that, in the event of a failure, the resulting action statement and LCO will encompass the planned shutdown for the refueling outage. Using this strategy, unplanned, unnecessary plant shutdowns, as a result of EFCV testing, will be avoided.

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. For those valves tested prior to the refueling outage, appropriate administrative and scheduling controls will be established.

SR 3.6.1.3.9 allows a representative sample of EFCVs to be tested every 24 months, such that each EFCVs will be tested at least once every 10 years (nominal).

The EFCVs have position indication in the control room. Check valve remote position indication is excluded from RG 1.97 as a required parameter for evaluating containment isolation. The remote position indication will be verified in the close direction at the same frequency as the exercise test, which will be performed at the frequency prescribed in TS SR 3.6.1.3.9. After the close-position test, the valve will be reset, and the remote open-position indication will be verified. Although inadvertent actuation of an EFCV during operation is highly unlikely due to the spring poppet design, PPL verifies the EFCVs indicate open in the control room at a frequency greater than once every two years.

In summary, considering the extremely low failure rate along with personnel and plant safety concerns to perform testing, the alternative sampling plan proposed provides an acceptable level of quality and safety.

NRC Staffs Evaluation of Relief Request No. RR-03 EFCVs are installed on boiling water reactor instrument lines to limit the release of fluid in the event of an instrument-line break. The EFCVs are installed in the nuclear boiler, reactor recirculation, reactor water cleanup, RCIC, HPCI, RHR, and CS systems. EFCVs are not required to close in response to a containment isolation signal and are not required to operate under post-LOCA conditions.

EFCVs are required to be tested in accordance with paragraph ISTC-3522, which requires exercising check valves nominally every 3 months to the positions required to perform their safety functions. Paragraph ISTC-3522(c) permits deferral of this requirement to every reactor refueling outage. EFCVs are also required to be tested in accordance with paragraph ISTC-3700, which requires remote position verification at least once every 2 years.

The current SSES 1 and 2 TS SR 3.6.1.3.9 states, Verify a representative sample of reactor instrumentation line EFCVs actuate to check flow on a simulationed instrument line break every 24 months. The TS Bases SR, 3.6.1.3.9, states that the representative sample consists of an equal number of EFCVs such that each EFCV is tested at least once every 10 years (nominal).

This revised TS SR was previously approved by the NRC staff on April 11, 2001 (ADAMS Accession No. ML010960024).

The previously approved TS SR was based on GEs Nuclear Energy Topical Report, NEDO-32977-A, Excess Flow Check Valve Testing Relaxation, dated June 2000. The topical report provided: (1) an estimate of steam release frequency (into the reactor building) due to a break in an instrument line concurrent with an EFCVs failure to close, and (2) an assessment of the radiological consequences of such a release. The NRC staff reviewed the GE topical report and issued its SE on March 14, 2000 (ADAMS Accession No. ML003691722). In its evaluation, the NRC staff agreed that the test interval may be extended up to a maximum of 10 years. In conjunction with this finding, the NRC staff noted that each licensee adopting the relaxed test interval program for EFCVs must have a failure feedback mechanism and corrective action program to ensure EFCV performance continues to be bounded by the topical reports results. Also, each licensee should perform a plant-specific radiological dose assessment, an EFCV failure analysis, and a release frequency analysis to confirm that they are bounded by the generic analyses of the topical report.

In the SE of previously approved TS Amendments Nos. 193 and 168 for SSES 1 and 2, respectively, dated April 11, 2001 (ADAMS Accession No. ML010960024), the NRC staff reviewed PPLs proposal for its applicability to GE Topical Report NEDO-32977-A and conformance with approved NRC staff guidance regarding radiological dose assessment, EFCV failure rate and release frequency, and the proposed failure feedback mechanism and corrective action program. Based on its review, the NRC staff concluded that the radiological consequences of an EFCV failure are sufficiently low and acceptable, and that the alternative testing in conjunction with the corrective action plan provides a high degree of valve reliability and operability. In its letter dated April 5, 2001 (ADAMS Accession No. ML011010179), PPL also committed to the following corrective actions:

1. Should a test failure occur, PPL will test an additional representative sample.

2. Should a test failure occur in the additional representative sample, PPL will test all the remaining representation valves.

3. For each test failure, PPL will retest the affected valve during the subsequent test interval. This test will be in addition to the number of tests required to be performed.

PPL states that, a representative sample of EFCVs with position indicators shall be observed locally once every 24 months with all EFCVs being observed at least once every 10 years (nominal).

Additionally, an orifice is installed upstream of the EFCVs to limit reactor water leakage in the event of rupture. The orifice limits leakage to a level where the integrity and functional performance of secondary containment and associated safety systems are maintained.

Therefore, the NRC staff finds that PPLs proposed test alternative provides an acceptable level of quality and safety.

Conclusion Based on the above evaluation, the NRC staff finds the proposed alternative which would allow a representative sample of EFCVs to be tested every 24 months with all EFCVs being tested at least once every 10 years (nominal) provides an acceptable level of quality and safety.

Therefore, PPLs proposed alternative to the ASME Code testing requirements is authorized pursuant to 10 CFR50.55a(a)(3)(i).

3.4 Valve Relief Request No. RR-05 ASME OM Code Requirements Paragraph ISTC-3522(c), Category C Check Valves requires that if exercising is not practical during operation at power and cold shutdowns, it shall be performed during refueling outages.

Specific Relief Requested PPL requested relief from Paragraph ISTC-3522(c) for the valves listed in the Table 5-1.

Table 5-1 Valve Number System Size Category Safety Inch Class Unit 1 Unit 2 149F028 249F028 Reactor Core Isolation 2 A/C 2 Cooling 149F040 249F040 Reactor Core Isolation 10 A/C 2 Cooling 155F049 255F049 High-Pressure Coolant 20 A/C 2 Injection Check valve 149F028/249 F028 is the RCIC vacuum pump discharge check valve to the suppression pool. It has a close safety function for containment isolation. Check valve 149F040/249F040 is the RCIC turbine exhaust check valve to the suppression pool. It has an open safety function to provide a flow path from the RCIC turbine exhaust to the suppression pool and a close safety function for containment isolation. Check valve 155F049/255F049 is the HPCI turbine exhaust check valve to the suppression pool. It has an open safety function to provide a flow path from the HPCI turbine exhaust to the suppression pool and a close safety function for containment isolation.

PPLs Basis for Relief Pursuant to 10 CFR 50.55a(a)(3), PPL requested authorization of an alternative to the requirements of ASME OM Code, paragraph ISTC-3522(c). The basis for the request is that the proposed alternative would provide an acceptable level of quality and safety.

The check valves listed in Table 5-1 are with no external means for exercising and no external position indication. The only practical means to verify closure is by Appendix J local leakage rate testing. This involves the setup of test equipment and system configuration changes that are impractical on a quarterly or cold shutdown basis. The Appendix J testing can be performed at intervals other than refueling outages such as during system outage windows.

Prior to performing a system outage on-line, its effect on risk is evaluated in accordance with the requirements of 10 CFR 50.65(a)(4), Requirements for Monitoring the Effectiveness of Maintenance at Nuclear power Plants. This requirement states in part that before performing maintenance activities (including but not limited to surveillance, post-maintenance testing, and corrective and preventive maintenance) PPL shall assess and manage the increase in risk that may result from the proposed maintenance activities.

PPL complies with the requirements of 10 CFR 50.65(a)(4) via application of a program governing maintenance scheduling. The program dictates the requirements for risk evaluations as well as the necessary levels of action required for risk management in each case. The program also controls operation of the on-line risk monitoring system, which is based on PRA.

With the use of risk evaluation for various aspects of plant operations, PPL has initiated efforts to perform additional maintenance, surveillance, and testing activities during normal operation.

Planned activities are evaluated utilizing risk insights to determine the impact on safe operation of the plant and the ability to maintain associated safety margins. Individual system components, a system train, or a complete system may be planned to be out of service to allow maintenance, or other activities, during normal operation.

Appendix J testing may involve a system breach, if required to repair a failed valve. However, during the disassembly process to perform maintenance, the subject valve is isolated and the associated section of piping drained. Thus, the system breach does not increase the risk due to internal flooding or internal system loss-of-coolant accident. The risk associated with these activities would be bounded by the risk experienced due to the system outage. Therefore, closure testing of these valves by Appendix J during schedule system outages while on-line would have no additional impact on core damage frequency.

As more system outages are performed on-line, it is evident that selected refueling outage inservice testing activities (e.g., closure testing by leak testing) could be performed during these system outage windows without sacrificing the level of quality or safety. Inservice testing performed on a refueling outage frequency is currently acceptable in accordance with the ASME OM Code 1998 edition through the 2000 addenda. By specifying testing activities on a frequency commensurate with each refueling outage, the ASME OM Code 1998 edition through the 2000 addenda, establishes an acceptable time period between testing. Historically, the refueling outage has provided a convenient and defined time period in which testing activities could be safely and efficiently performed. However, an acceptable testing frequency can be maintained separately without being tied directly to a refueling outage. Inservice testing performed on a frequency that maintains the acceptable time period between testing activities during the operating cycle is consistent with the intent of the ASME OM Code 1998 edition

through the 2000 addenda.

Over time, approximately the same number of tests will be performed using the proposed operating cycle frequency as would be performed using the current refueling outage frequency.

Thus, inservice testing activities performed during the proposed operating cycle test frequency provide an equivalent level of quality and safety.

PPLs Proposed Alternative Examination Pursuant to 10 CFR 50.55a(a)(3)(i), PPL proposes an alternative testing frequency for performing inservice testing of the valves identified above. The valves will be closure tested by Appendix J on a frequency of at least once per operating cycle in lieu of once each refueling outage as currently allowed by the ASME OM Code 1998 edition through the 2000 addenda, paragraph ISTC-3522(c), Category C Check Valves. The open safety function of check valve 149F040/249F040 will be demonstrated quarterly in conjunction with the RCIC flow verification (inservice pump test). The open safety function of check valve 155F049/255F049 will be demonstrated quarterly in conjunction with the HPCI flow verification (inservice pump test). As required by Paragraph ISTC-5221(a)(3), the open function of check valve 149F028/249F028 will be demonstrated quarterly in conjunction with the RCIC flow verification (inservice pump test).

NRC Staffs Evaluation of Relief Request No. RR-05 Paragraph ISTC-3522(c), Category C Check Valves, requires that if exercising is not practical during operation at power and cold shutdown, it shall be performed during refueling outages.

In this relief request RR-05, PPL proposes an alternative testing frequency for performing inservice testing of the RCIC and HPCI check valves. These valves will be closure tested by Appendix J on a frequency of at least once per operating cycle in lieu of once each refueling outage. The open safety function of check valves will be demonstrated quarterly in conjunction with the RCIC or HPCI flow verification (inservice pump tests).

PPL states that the RCIC and HPCI check valves listed in Table 5-1 are with no external means for exercising and no external position indication. The only means to verify closure is by Appendix J local leakage rate testing. This involves set-up of test equipment and system configuration changes that are a hardship without a compensating increase in quality or safety on a quarterly or cold shutdown basis.

The check valve 149F028/249F028 is the RCIC vacuum pump discharge check valve to the suppression pool. It has a close safety function for containment isolation. Check Valve 149F040/249F040 is the RCIC turbine exhaust check valve to the suppression pool. It has an open safety function to provide a flow path from the RCIC turbine exhaust to the suppression pool and a close safety function for containment isolation. Check valve 155F049/255F049 is the HPCI turbine exhaust check valve to the suppression pool. It has an open safety function to provide a flow path from the HPCI turbine exhaust to the suppression pool and a close safety function for containment isolation.

The NRC staff finds that the proposed Appendix J test for closure of check valves on a frequency of at least once per operating cycle, in lieu of once each refueling outage, and open safety function check valves will be demonstrated quarterly in conjunction with the RCIC or

HPCI flow verification (inservice pump test), are the appropriate methods to verify operability, and can be accomplished during system outages when the plant is on-line or during refueling outages. The NRC staffs finding is based on the following considerations:

! PPL proposes to verify the check valves closure by Appendix J local leakage rate testing. The Appendix J method of testing is accepted by the ASME OM Code paragraph ISTC-3620. This involves set-up of test equipment and system configuration changes that are a hardship without a compensating increase in quality or safety on a quarterly or cold shutdown basis.

! There are no technical barriers to performing these IST activities during either the refueling outage or the operating cycle.

! PPL states that a typical on-line system outage window (SOW) lasts utilizing 96 hours4 days <br />0.571 weeks <br />0.132 months <br /> (<

50%) of the allowable LCO time limit. Based on the review of maintenance history, disassembly, inspection, and reassembly of the check valves, it takes between 6 and 24 hours1 days <br />0.143 weeks <br />0.0329 months <br />, depending upon size and location. The IST activity would be conducted simultaneously with other maintenance activities scoped into the SOW. This provides adequate margin to complete disassembly and inspection activities in an orderly manner.

! PPL states that performing IST activity on-line would not change the duration of the on-line SOW or the CDP associated with the existing on-line activities. Therefore, the risk/CDP over the entire operating/shutdown spectrum would remain unchanged.

! PPL is using EOOS software to evaluate plant configuration risk when the SSES 1 and 2 is in Mode 1, 2, or 3. The EOOS processes the activities through an SSES 1 and 2 specific model and calculates overall effects on nuclear safety. EOOS determines that the increase in CDF and LERF, which provides the risk impact in performing disassembly and inspection of these check valves while the plant is on-line, is insignificant.

! The refueling outages have provided a practical and definitive time period in which inspection testing activities can be safely and effectively performed. An acceptable testing frequency can be maintained separately without being tied directly to a refueling outage. IST performed on a frequency (24 months) that maintains the acceptable time period between testing activities during the operating cycle (i.e., 24 months) is consistent with the intent of the ASME OM Code and GL-89-04.

! Over time, approximately the same number of tests will be performed using the proposed operating cycle test frequency as would be performed using the current refueling outage frequency. Thus, inservice testing activities performed during the proposed operating cycle, i.e., 24 months, test frequency provide an equivalent level of quality and safety as IST performed at a refueling outage frequency.

On the basis of these considerations, the NRC staff finds that the proposed alternative provides an acceptable level of quality and safety.

Conclusion The NRC staff concludes that PPLs proposed alternative provides an acceptable level of quality and safety. Therefore, the proposed alternative to disassemble and inspect the check valves once every operating cycle in lieu of once during each refueling outage is authorized pursuant to 10 CFR 50.55a(a)(3)(i).

4.0 CONCLUSION

S Based on the review of the information provided in relief requests RR-01, RR-02, RR-03, and RR-05, the NRC staff concludes that PPLs proposed alternatives will provide an acceptable level of quality and safety. Therefore, the proposed alternatives RR-01, RR-02, RR-03, and RR-05 are authorized pursuant to 10 CFR 50.55a(a)(3)(i).

Principal Contributor: G. Bedi Date: March 10, 2005