ML14142A089

Surry Power Station, Units No. 1 and 2, Fourth Interval Inservice Inspection (ISI) Program, System Pressure Testing (Spt), SPT-003 and SPT-002 (TAC MF3232 and MF3233)

ML14142A089
Person / Time
Site:	Surry
Issue date:	05/28/2014
From:	Pascarelli R J Plant Licensing Branch II
To:	Heacock D A Virginia Electric & Power Co (VEPCO)
Pascarelli R J
References
TAC MF3232, TAC MF3233
Download: ML14142A089 (9)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Mr. David A. Heacock President and Chief Nuclear Officer Virginia Electric and Power Company lnnsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711 May 28, 2014

SUBJECT:

SURRY POWER STATION, UNITS 1 AND 2-FOURTH INTERVAL INSERVICE INSPECTION (lSI) PROGRAM, SYSTEM PRESSURE TESTING (SPT), SPT-003 AND SPT-002 (TAC NOS. MF3232 AND MF3233) .

Dear Mr. Heacock:

By letter dated December 9, 2013 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 13350A 1 09), as supplemented by letter dated March 7, 2014 (ADAMS Accession No. ML 14072A009), Virginia Electric and Power Company -Dominion (the licensee) submitted for the U.S. Nuclear Regulatory Commission (NRC) approval requests for alternative (RFA) SPT-003, Revision 1, and RFA SPT-002, Revision 1. The licensee proposed an alternative to a certain requirement of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI. The requests relate to the inservice inspection (lSI) requirement of IWB-5222(b) when the licensee conducts system leakage tests of the reactor vessel head vent (RVHV) lines at or near the end of the inspection interval.

The licensee submitted RFA SPT-003, Revision 1, for the Surry Power Station (Surry), Unit 1, and RFA SPT-002, Revision 1, for Surry, Unit 2. The NRC staff has concluded based on the information provided by the licensee, that pursuant to Title 10 of the Code of Federal Regulations (1 0 CFR) 10 CFR 50.55a(a)(3)(ii), the Relief Request and alternative system leakage test of the RVHV lines is authorized on the basis that complying with the specified requirement would result in a hardship or unusual difficulty without a compensating increase in the level of quality and safety.

D. Heacock If you have any questions concerning this matter, please contact Dr. V. Sreenivas, at (301) 415-2597.

Docket Nos. 50-280 and 50-281

Enclosure:

Safety Evaluation cc w/encl: Distribution via Listserv / Sincerely, Robert J. Pascarelli, Branch Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REQUESTS FOR ALTERNATIVE SPT-003, REVISION 1, AND SPT-002, REVISION 1 REGARDING ALTERNATIVE TESTING REQUIREMENTS FOR SMALL DIAMETER REACTOR COOLANT SYSTEM PRESSURE BOUNDARY CONNECTIONS VIRGINIA ELECTRIC AND POWER COMPANY-DOMINION SURRY POWER STATION UNITS 1 AND 2 DOCKET NOS. 50-280 AND 50-281 INTRODUCTION By letter dated December 9, 2013 (Agencywide Documents Access and Management Systems (ADAMS) Accession No. ML 13350A 1 09), as supplemented by letter dated March 7, 2014 (ADAMS Accession No. ML 14072A009), Virginia Electric and Power Company-Dominion (the licensee) submitted for the U.S. Nuclear Regulatory Commission (NRC) approval requests for alternative (RFA) SPT-003, Revision 1, and RFA SPT-002, Revision 1. The licensee proposed an alternative to a certain requirement of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI. The requests relate to the inservice inspection (lSI) requirement of IWB-5222(b) when the licensee conducts system leakage tests of the reactor vessel head vent (RVHV) lines at or near the end of the inspection interval.

The licensee submitted RFA SPT-003, Revision 1, for the Surry Power Station (Surry), Unit 1, and RFA SPT-002, Revision 1, for Surry, Unit 2. In a letter dated December 12, 2002 (ADAMS Accession No. ML02351 0289), the licensee submitted for the NRC review and approval the original relief request SPT-003 for Surry, Unit 1. In a letter dated August 11, 2004 (ADAMS Accession No. ML042250379), the NRC approved the original relief request SPT-003 for Surry, Unit 1. In the December 9, 2013, letter, the licensee revised the original relief request SPT-003 and submitted it as SPT-003, Revision 1. The revised relief request contains all of the previously approved relief request items plus the new request for approval for the RVHV lines. Therefore, the licensee is only asking relief from IWB-5222(b) for the RVHV lines at this time. In a letter dated August 25, 2003 (ADAMS Accession No. ML032471647), the licensee submitted for the NRC review and approval, the original relief request SPT-002 for Surry, Unit 2. In a letter dated September 9, 2004 (ADAMS Accession No. ML042540167), the NRC approved the original relief request SPT -002 for Surry, Unit 2. Enclosure In the December 9, 2013 letter, the licensee revised the original relief request SPT-002 and submitted it as SPT-002, Revision 1. The revised relief request contains all of the previously approved relief request items plus the new request for approval for the RVHV lines. Therefore, the licensee is only asking relief from IWB-5222(b) for the RVHV lines at this time. *Specifically, pursuant to Title 10 of the Code of Federal Regulations (1 0 CFR) 50.55a(a)(3)(ii), the licensee proposed an alternative system leakage test of the RVHV lines on the basis that complying with the specified requirement would result in a hardship or unusual difficulty withouta compensating increase in the level of quality and safety . . REGULATORY EVALUATION Pursuant to 10 CFR 50.55a(g)(4), the ASME Code Class 1, 2, and 3 components (including supports) must meet the requirements, except the design and access provisions and the preservice examination requirements, set forth in the ASME Code,Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components," to the extent practical within the limitations of design, geometry, and materials of construction of the components.

The inservice examination of components and system pressure tests conducted during the first 1 0-year interval and subsequent intervals must 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 conditions listed Pursuant to 10 CFR 50.55a(a)(3), alternatives to the requirements of paragraph (g) of 1 0 CFR 50.55a may be used, when authorized by the NRC, if the licensee demonstrates (i) the proposed alternatives would provide an acceptable level of quality and safety or (ii) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. Based on the above, and subject to the following technical evaluation, the NRC staff finds that regulatory authority exists for the licensee to request and the NRC to authorize the alternative requested by the licensee.

TECHNICAL EVALUATION The Licensee's Request for Alternative For Surry, Unit 1, the components affected are two parallel RVHV lines consisting of Line Number 1-RC-233-1502 between isolation valves 1-RC-SOV-1 OOA 1 and 1-RC-SOV-1 OOA2, and a parallel Line Number 1-RC-234-1502 between isolation valves 1-RC-SOV-1 0081 and 1-RC-SOV-1 0082. All four valves are normally closed. For Surry, Unit 2, the components affected are two parallel RVHV lines consisting of a 1 inch line between isolation valves 2-RC-SOV-200A-1 and 2-RC-SOV-200A-2, and a parallel1 inch line between valves 2-RC-SOV-2008-1 and 2-RC-SOV-2008-2.

All four valves are normally closed. The above RVHV lines are classified as ASME Code Class 1 pressure retaining boundary and Examination Category B-P, Item Numbers 815.50 and 815.70 in accordance with IWB-2500, Table IWB-2500-1.

The above RVHV lines were manufactured from ASME SA376 TP316/ASTM A-376-TP-316 austenitic stainless steel. The code of record for the fourth 1 0-year interval lSI at Surry, Units 1 and 2, is the 1998 Edition through 2000 Addenda of the ASME Code .. The ASME Code,Section XI, IWB-2500, Table IWB-2500-1, Examination Category 8-P, requires system leakage tests according to IWB-5220 and the VT-2 visual examination according to IWA-5240, during each refueling outage. As required by IWB-5221 (a), the system leakage test shall be conducted at a pressure *not less than the pressure corresponding to 100 percent rated reactor power. In accordance with IWB-5222(a), the pressure retaining boundary during the system leakage test shall correspond to the reactor coolant boundary, with all valves in the position required for normal reactor operation startup. The required visual examination shall, however, extend to and include the second closed valve at the boundary extremity.

In accordance with IWB-5222(b), the pressure retaining boundary during system leakage test conducted at or near the end of each inspection interval shall extend to all Class 1 pressure retaining components within the system boundary.

The licensee proposed an alternative to IWB-5222(b) when conducting a system leakage test of the above RVHV lines. Specifically, in lieu of the extended system leakage test boundaries of IWB-5222(b), the licensee proposed to use ASME Code Case N-798 "Alternative Pressure Testing Requirements for Class 1 Piping between the First and Second Vent, Drain, and Test Isolation Devices," when conducting the system leakage test of the RVHV lines. The licensee submitted AFA SPT-003, Revision 1, and RFA SPT-002, Revision 1, because the NRC staff has not approved ASME Code Case N-798 in the latest revision (Revision

16) of Regulatory Guide 1.147. ASME Code Case N-798 requires that for portions of Class 1 vent, drain, and test piping between the first and second isolation devices that normally remain closed during plant operation, only the boundaries of IWB-5222(a) shall apply. This means that under ASME Code Case N-798, the system leakage test of the subject piping can be performed without satisfying IWB-5222(b) as long as IWB-5222(a) is satisfied.

• The licensee provided its basis for hardship or unusual difficulty caused by compliance with the IWB-5222(b) requirement when conducting system leakage test of the RVHV lines. The licensee stated in letter dated March 7, 2014, that Technical Specification

3.1.7 requires

that at least two RVHV paths consisting of two isolation valves in a series, powered from emergency buses, shall be OPERABLE and closed whenever the reactor coolant system (RCS) temperature and pressure are greater than 350 degrees Fahrenheit

(°F) and 450 pound per square inch gauge (psig), respectively.

Extending the pressure retaining boundary during the ASME Code required system leakage test to the RVHV lines beyond the first normally closed valve would require a number of temporary system configuration changes. The licensee stated that temporary installation of testing equipment would be required to achieve test pressures at system segments beyond the first isolation valve. While the RCS is being brought to the normal operating temperature of 54rF and pressure of 2235 psig in accordance with the Technical Specification, the RVHV lines are isolated from the RCS. Consequently, the RVHV lines would be at pressures and temperatures that are less than the full RCS pressure and temperature.

The plant design configuration complies with Technical Specification 3.1. 7 and the RCS boundary requirements for double isolation, but cannot satisfy the ASME Code system leakage test requirement for nominal operating pressure associated with 1 00 percent rated reactor power between the piping segments.

The licensee stated that use of temporary hoses, a test rig, and a hydrostatic pump to raise the RVHV lines to the RCS pressure and temperature to perform the ASME Code system leakage test constitutes a personnel safety hazard and could adversely affect plant safety. Temporary hoses are not qualified to meet all aspects of the plant design. The failure of temporary hoses during pressure testing could result in personnel injury, as well as the loss of the reactor coolant pressure boundary and reactor coolant inventory.

Establishment and restoration from such temporary configurations could take a considerable amount of time to complete, result in an unwarranted increase in worker radiation exposure, ar.1d contaminate test equipment.

The dose associated with use of a test rig to conduct the ASME Code system leakage testing of similar piping during the 1998 North Anna, Unit 1, refueling outage, was estimated to be 1 .5 roentgen equivalent man (rem). It is expected that conditions at Surry, Units 1 and 2, would yield comparable exposure results if the ASME Code system leakage testing were performed.

The licensee stated that the connections in the RVHV lines are typically welds that received a surface examination after installation.

The affected portion of the RVHV lines is isolated during normal operation and does not experience pressure loading unless there is a leak at the first isolation valve. These RVHV lines and its associated components are near the free end of a cantilever configuration (i.e., stub end isolated by either a valve or a flange). There is no brace or support for this portion of the pipe. Consequently, this portion does not experience any thermal loading. The valves do not have an extension operator, so the rotational accelerations at the valve do not produce additional stresses.

The stresses toward the free end of the cantilever, due to other types of loading, are only a small fraction of the applicable ASME Code. The licensee stated that the RCS leakage and radiation levels are monitored in accordance with the requirements of the applicable Technical Specifications.

The licensee stated that operating experience reviews of its fleet and Surry, and survey of other similar plants, did not identify potential degradation such as stress corrosion cracking or fatigue in socket welds of the RVHV lines. The licensee submitted these requests for the remainder of the fourth 1 0-year lSI interval of Surry, Unit 1, which commenced on October 14, 2003, and Surry, Unit 2, which commenced on May 10, 2004. The licensee stated that the fourth 1 0-year lSI interval will be extended by one year in accordance with IWA-2430(d) to end on October 13, 2014, for Surry, Unit 1, and on May 9, 2015, for Surry, Unit 2. The NRC staff notes that the extension of the lSI interval within one calendar year is permitted under IWA-2430.

NRC Staff Evaluation The NRC staff has evaluated RFA SPT-003, Revision 1, and RFA SPT-002, Revision 1, pursuant to 10 CFR 50.55a(a)(3)(ii).

The NRC staff focuses on whether compliance with the specified requirements of 10 CFR 50.55a(g), or portions thereof, would result in a hardship or unusual difficulty, without a compensating increase in the level of quality and safety. The NRC staff determined that compliance with the ASME Code,Section XI, IWB-5222(b), during system leakage testing of the RVHV lines* would result in a hardship.

The basis for the hardship is as follows. In order to perform the system leakage testing of the RVHV lines in accordance with IWB-5222(b), the licensee would either have to (1) open the first isolation valve which would pressurize the pipe segment between the two isolation valves to the RCS pressure, or (2) leave both isolation valves closed and pressurize the pipe segment between the two closed valves using a temporary connection.

The first option would defeat the double isolation criteria of the design basis and violate the plant Technical Specification requirement.

This would reduce safety of the *plant operation.

The second option would require the licensee to make temporary changes to the existing piping configuration in order to pressurize the piping segments between the two closed valves. Personnel involved in installing and removal of the temporary configuration changes and conducting the ASME Code required system leakage test of the RVHV lines would incur additional dose which would be of a concern from an as low as is reasonably achievable criteria.

The licensee's use of temporary hoses to facilitate manual pressurization of the RVHV lines to the RCS pressure could fail which would create hazards for personnel and contaminate test equipment.

Therefore, the NRC staff determines that performing system leakage testing in accordance with IWB-5222(b) constitutes a hardship.

The NRC staff finds that the licensee will conduct the system leakage test of the RVHV lines in accordance with ASME Code Case N-798 accompanied with the VT-2 visual examination according to IWA-5240.

Specifically, the licensee will VT-2 examine the non-isolated portion of the RVHV lines while the piping is under the RCS pressure and temperature.

The licensee will also VT -2 examine the isolated portion of the RVHV liries with all valves in a closed position, extended to include the second closed valve at the boundary extremity in accordance with ASME Code Case N-798. The NRC staff determined that by performing the VT-2 examination of isolable and non-isolable segments of the RVHV lines according to IWA-5240, during normal reactor startup, the licensee will be able to detect any leakage originating from a flaw in the RVHV lines without any major design modifications to the existing piping. The licensee stated that a review of operating experience of its fleet, including Surry, and a survey of other similar plants, did not identify any documented degradation of the RVHV lines. Based on the review of operating experience, the NRC staff has not identified any documented degradation due to stress corrosion cracking and fatigue in the RVHV lines. Furthermore, the NRC staff determined that the existing reactor coolant leakage detection systems are sufficient to provide warning to the control room operator in an unlikely event of a through wall leak in the RVHV lines. The NRC staff finds that if the subject piping developed a through wall flaw, the reactor coolant leakage detection systems will be able to identify the leakage during normal operation, and the licensee will take appropriate corrective actions in accordance with the plant Technical Specifications.

Use of ASME Code Case N-798 The NRC has not yet accepted ASME Code Case N-798 in Regulatory Guide 1.147 by rulemaking (1 0 CFR 50.55a). The NRC staff authorizes use of ASME Code Case N-798 for the system leakage testing of RVHV lines at Surry, Units 1 arid 2; however, its use is limited to the end of the fourth 1 0-year lSI interval or until such time as this code case is approved by the NRC in Regulatory Guide 1.147, whichever occurs earlier. Should the NRC approve the code case and if the licensee intends to continue using this code case, it must follow all provisions of ASME Code Case N-798 with conditions as specified in Regulatory Guide 1.147 and 10 CFR 50.55a(b)(4), (b)(5), and (b)(6), if any. In summary, the NRC staff finds that the proposed system leakage testing of the RVHV lines is adequate to provide reasonable assurance of the structural integrity and leak tightness of the subject piping segments. CONCLUSION As set forth above, the NRC staff determines that the proposed alternative provides reasonable assurance of the structural integrity and leak tightness of the RVHV lines. The NRC staff finds that complying with the specified ASME Code requirement would result in a hardship or unusual difficulty without a compensating increase in the level of quality and safety. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(a)(3)(ii).

Therefore, the NRC staff authorizes use of RFA SPT-003, Revision 1, for the RVHV lines at Surry, Unit 1, for the fourth 1 0-year lSI interval which commenced on October 14,2003, and will end on October 13,2014. Furthermore, the NRC staff authorizes use of RFA SPT-002, Revision 1, for the RVHV lines at Surry, Unit 2, for the fourth 1 0-year lSI interval which commenced on May 1 0, 2004, and will end on May 9, 2015. All other ASME Code,Section XI, requirements for which relief was not specifically requested and authorized herein by the staff remain applicable, including the third party review by the Authorized Nuclear In-service Inspector.

Principal Contributors:

Ali Rezai, NRR May 28, 2014 D. Heacock If you have any questions concerning this matter, please contact Dr. V. Sreenivas, at (301) 415-2597.

Docket Nos. 50-280 and 50-281

Enclosure:

Safety Evaluation cc w/encl: Distribution via Listserv DISTRIBUTION:

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