ML073340847

Proposed Alternative Relief Request to Performance of System Pressure Test on Instrument Air Piping, Second Ten-Year Inservice Inspection Interval

ML073340847
Person / Time
Site:	Clinton
Issue date:	12/13/2007
From:	Russell Gibbs NRC/NRR/ADRO/DORL/LPLIII-2
To:	Pardee C Exelon Generation Co
Sands S,NRR/DORL, 415-3154
References
TAC MD4896
Download: ML073340847 (7)

Text

December 13, 2007 Mr. Charles G. Pardee Chief Nuclear Officer and Senior Vice President Exelon Generation Company, LLC 4300 Winfield Road Warrenville, IL 60555

SUBJECT:

CLINTON POWER STATION, UNIT NO. 1 - PROPOSED ALTERNATIVE RELIEF REQUEST TO PERFORMANCE OF SYSTEM PRESSURE TEST ON INSTRUMENT AIR PIPING, SECOND TEN-YEAR INSERVICE INSPECTION INTERVAL (TAC NO. MD4896)

Dear Mr. Pardee:

By letter dated December 15, 2006, to the Nuclear Regulatory Commission (NRC), AmerGen Energy Company, LLC (the licensee) requested approval of an alternative to the requirements of the American Society of Mechanical Engineers (ASME) Code,Section XI, with regard to pressure testing of all Class 2 and Class 3 instrument air piping supplying all safety relief valves and both feedwater containment outboard isolation check valves, for the Clinton Power Station, Unit No. 1 (CPS). As an alternative to the examination requirement of the Code, CPS proposed to perform a pressure decay test in lieu of the visual examination during pressure testing of each safety relief valve and the feedwater isolation check valve accumulators, including all associated piping. The NRC staff has evaluated the licensees request for relief from the 1989 Edition, ASME Code,Section XI, and authorizes the proposed alternative pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(a)3(i) for the second 10-year inservice inspection interval of CPS. The NRC staffs review of the licensees analysis in support of its request for relief is documented in the enclosed safety evaluation.

Sincerely,

/RA/

Russell Gibbs, Chief Plant Licensing Branch III-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-461

Enclosure:

Safety Evaluation cc w/encl: See next page

ML073340847 OFFICE LPL3-2/PM LPL3-2/LA DCI/CSGB/BC OGC* LPL3-2/BC NAME SSands EWhitt AHiser MSmith RGibbs DATE 12/4/2007 12/4/2007 12/5/2007 12/12/2007 12/13/2007 Clinton Power Station, Unit No. 1 cc:

Corporate Distribution Exelon Generation Company, LLC via e-mail Clinton Distribution Exelon Generation Company, LLC via e-mail Clinton Senior Resident Inspector U.S. Nuclear Regulatory Commission via e-mail Illinois Emergency Management Agency Division of Disaster Assistance &

Preparedness via e-mail J. W. Blattner, Project Manager Sargent & Lundy Engineers via e-mail Chairman of DeWitt County c/o County Clerks Office DeWitt County Courthouse Clinton, IL 61727

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUEST OF THE SECOND 10-YEAR INTERVAL INSERVICE INSPECTION PROGRAM PROPOSED ALTERNATIVE TO SYSTEM PRESSURE TEST CLINTON POWER STATION, UNIT NO. 1 AMERGEN ENERGY COMPANY, LLC DOCKET NO. 50-461

1.0 INTRODUCTION

By letter to the Nuclear Regulatory Commission (NRC) dated December 15, 2006 (Agencywide Documents Access and Management System Accession No. ML063620098), AmerGen Energy Company, LLC, (the licensee) requested approval of an alternative to the requirement of the American Society of Mechanical Engineers (ASME) Code,Section XI, with regard to pressure testing of all Class 2 and Class 3 instrument air (IA) piping supplying all safety relief valves (SRV) and both feedwater containment outboard isolation check valves, for Clinton Power Station, Unit No. 1 (CPS). As an alternative to the examination requirement of the Code, CPS proposed to perform a pressure decay test in lieu of the visual examination during pressure testing of each SRV and the feedwater isolation check valve accumulators, including all associated piping as required under surveillance procedure, CPS 9061.11, that verifies the operability of SRV and check valves in the IA supply lines to all 16 SRVs and both feedwater containment outboard isolation check valves in accordance with the CPS inservice test program.

The NRC staff has reviewed the licensees proposed alternative pursuant to Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Section 55a(a)(3)(i), that the alternative would provide an acceptable level of quality and safety.

2.0 REGULATORY REQUIREMENTS Section 50.55a(g) of 10 CFR requires that inservice inspection (ISI) of ASME Code Class 1, 2, and 3 components be performed in accordance with Section XI of the ASME Code and applicable addenda, except where specific written relief has been granted by the NRC, pursuant to 10 CFR 50.55a(g)(6)(i). According to 10 CFR 50.55a(a)(3), alternatives to the requirements of paragraph 50.55a(g) may be used, when authorized by the NRC, if an applicant demonstrates that the proposed alternatives would provide an acceptable level of quality and safety, or if the specified requirement would result in hardship, or unusual difficulty without a compensating increase in the level of quality and safety.

Pursuant to 10 CFR 50.55a(g)(4), ASME Code Class 1, 2, and 3 components (including supports) shall meet the requirements, except the design and access provisions and the preservice examination requirements, set forth in the ASME Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, to the extent practical within the limitations of design, geometry, and materials of construction of the components. The Enclosure

regulations require that ISI of components and system pressure tests conducted during the first 10-year interval and subsequent intervals, comply with the requirements in the latest edition and addenda of Section XI of the ASME Code incorporated by reference in 10 CFR 50.55a(b),

12 months prior to the start of the 120-month interval, subject to the limitations and modifications listed therein. The ISI Code of Record for the second 10-year ISI interval is the 1989 Edition of the ASME Code,Section XI.

3.0 TECHNICAL EVALUATION

3.1 System/Component(s) for Which Relief is Requested:

Class 2 IA piping and components between containment isolation valves.

Class 3 IA piping and components supplying to all SRVs and both feedwater containment outboard isolation check valves.

3.2 ASME Code Requirements The 1989 Edition of the ASME Code,Section XI, paragraphs IWC-2500 and IWD-2500, state that components shall be examined and pressure tested as specified in Tables IWC-2500-1 and IWD-2500-1. Tables IWC-2500-1 and IWD-2500-1 require performance of VT-2 visual examination during system pressure tests.

3.2.1 Examination Categories:

Class 2 - C-H, All Pressure Retaining Components.

Class 3 - D-A, Systems in Support of Reactor Shutdown Function.

3.2.2 Item Numbers:

Class 2 - C7.30, Piping and Pressure Retaining Components and C7.70, Valves and Pressure Retaining Components.

Class 3 - D1.10, Pressure Retaining Components.

3.3 Licensees Request for Relief The licensee requested relief from performance of system pressure tests and VT-2 visual examination requirements specified in Tables IWC-2500-1 and IWD-2500-1 for all Class 2 and Class 3 IA piping supplying all SRVs and both feedwater containment outboard isolation check valves.

3.4 Basis for Requesting Relief Pursuant to 10 CFR 50.55a(a)(3), relief is requested on the basis that the proposed alternatives provide an acceptable level of quality and safety. Surveillance procedure CPS 9061.11, Instrument Air Check Valve Operability and Pipe Pressure Test, verifies the operability of SRV actuation and check valves in the IA supply lines to all 16 SRVs and both feedwater containment outboard isolation check valves. This surveillance test is performed for each individual SRV and both feedwater containment outboard isolation check valves as a requirement of the CPS inservice test (IST) program. One specific test this surveillance

performs, is a pressure decay test of the accumulators of the SRV and the feedwater containment outboard isolation check valves as well as associated piping and valves. The pressure decay test is performed by isolating and pressurizing these accumulators and associated piping to the nominal operating pressure. The decay in pressure is then monitored through calibrated pressure measuring instrumentation. If any pressure decay acceptance criterion of the procedure is exceeded, the surveillance identifies appropriate troubleshooting steps to perform including application of soap solution to locate leakage.

The pressure decay test performed as part of CPS 9061.11 identifies any degradation of the Class 2 and 3 IA supply piping to automatic depressurization system (ADS), the SRVs and the feedwater containment outboard isolation check valve accumulators and associated piping.

The volume tested by this surveillance encompasses all piping and components requiring testing under the ASME Code,Section XI for these portions of the IA system. This surveillance is performed on a greater frequency than that required per Tables IWC-2500-1 and IWD-2500-1 and the test pressure is consistent with the pressure requirements of both tables. Thus, the testing performed during this surveillance will provide the same level of quality and safety as that of the pressure testing and the VT-2 examination requirements of Tables IWC-2500-1 and IWD-2500-1.

The VT-2 visual examination described in Tables IWC-2500-1 and IWD-2500-1 and performed once per inspection period, would not provide an increase in safety, system reliability, or structural integrity. In addition, performance of a VT-2 visual examination would require applying a leak detection solution to a large amount of piping and components, many of which are in high radiation areas with limited access. The VT-2 examination would result in additional radiation exposure (estimated 2 roentgen equivalent man) and industrial safety challenges to personnel with no added benefit in the level of quality and safety. The visual examination would not be consistent with radiation exposure of practices As Low As Reasonably Achievable.

3.5 Licensees Proposed Alternate Examination As an alternative to the examination requirements of Tables IWC-2500-1 and IWD-2500-1, CPS will perform pressure decay testing on all Class 2 and Class 3 IA piping supplying all 16 SRVs and both feedwater containment outboard isolation check valves as required under CPS Inservice Test program outlined in Surveillance procedure CPS 9061.11.

4.0 STAFF EVALUATION The ADS utilizes selected SRVs for depressurization of the reactor. Each of the SRV utilized for automatic depressurization is equipped with an air accumulator and check valve arrangement. The accumulators assure that the valves can be held open following failure of the air supply to the accumulators.

The 1989 ASME Code,Section XI, requires a system leakage test of all pressure retaining components of the ADS and SRV accumulators including their associated piping once every 40 months, and a VT-2 visual examination is required to be conducted to detect evidence of leakage from the pressure retaining components. The ASME Code further states that the contained fluid in the system shall serve as the pressurizing medium. As an alternative to the

VT-2 visual examination requirements of the ASME Code, the licensee proposes to take credit for the technical specifications (TS) surveillance performed which states that each ADS SRV shall be determined operable automatically and manually on a 24-month frequency. In addition, the TS requires that these valves are to be surveillance tested in accordance with the IST program during each refueling outage. The licensees surveillance is governed by procedure CPS 9061.1, which verifies proper operation of the SRV IA supply system, actuator, solenoids, and if necessary, the proper valve stroke distance. This procedure requires performance of a pressure decay test by isolating and pressurizing the ADS and SRV accumulators including their associated piping to the nominal operating pressure and monitoring the pressure decay to ensure leak-tight integrity of the components. The decay in pressure of 1.5 psig within 60 to 108 minutes for larger volume components such as the accumulator headers for SRVs and 26 to 31 minutes for smaller volume components such as the accumulator header for the feedwater check valves as stated in the above surveillance procedure is an acceptable limit beyond which the pressure boundary is subjected to investigation for location of any leakage. The NRC staff considers this leakage criterion based on pressure decay to be an acceptable alternative to the ASME Code-required VT-2 visual examination of the pressure boundary. Even the ASME Code,Section XI, allows rate of pressure loss as an alternative to the VT-2 visual examination during system leakage test of buried components that are isolable by means of valves.

Nevertheless, assuming that the pressure decay provides an adequate assurance of a leak-tight integrity of the components during the surveillance, the proposed alternative offers further conservatism as reflected in the frequency of surveillance being once every 24 months as opposed to the ASME Code-required VT-2 visual examination frequency of once every 40 months. The NRC staff, believes that it is redundant to conduct a VT-2 visual examination in accordance with Tables IWC-2500-1 or IWD-2500-1 for Class 2 or Class 3 components of the ASME Code,Section XI, over and above the surveillance required under TS and the inservice tests to ensure leak-tight integrity of the pressure boundary of the ADS SRV accumulators and their associated piping. Therefore, the NRC staff has determined that the licensees proposed alternative would provide an acceptable level of quality and safety.

5.0 CONCLUSION

The NRC staff concludes that the licensees proposed alternative to perform pressure decay test for the ADS SRV accumulators and both feedwater containment outboard isolation check valves including associated piping every refueling outage in accordance with surveillance procedure CPS 9061.11, in lieu of the ASME Code-required VT-2 visual examination during each inspection period, would provide an acceptable level of quality and safety. Therefore, the proposed alternative is authorized pursuant to 10 CFR 50.55a(a)(3)(i) for CPS, for the second 10-year ISI interval. All other requirements of the ASME Code,Section XI for which relief has not been specifically requested remain applicable, including a third party review by the Authorized Nuclear Inservice Inspector.

Principal Contributor: P. Patnaik, NRR Date: December 13, 2007