NL-11-0162, Submittal of Revision to Relief Request RR-V-4 Fourth 10-Year Interval Inservice Testing Program

Submittal of Revision to Relief Request RR-V-4 Fourth 10-Year Interval Inservice Testing Program

ML110280011
Person / Time
Site:	Hatch
Issue date:	01/26/2011
From:	Marino P Southern Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-11-0162
Download: ML110280011 (11)

Text

Paula M.Marina Southern Nuclear Vice President Operating Company, Inc Engineering 40 Inverness Center Parkway Birmingham, Alabama 35242 Tel 205.992.7707 Fax 205.992.6165 pmmarino@southernco.com January 26, 2011 COMPANY Docket Nos.: 50-321 NL-1 1-0162 50-366 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Edwin I. Hatch Nuclear Plant Revision to Relief Request RR-V-4 Ladies and Gentlemen:

Pursuant to 10 CFR 50.55a(a)(3)(ii), Southern Nuclear Operating Company hereby submits the enclosed proposed revision to RR-V-4 for the Edwin I. Hatch Nuclear Plant (HNP), Units 1 and 2, Fourth 10-Year Interval Inservice Testing Program (IST). This request submits Version 2.0 to RR-V-4 to the ASME OM Code, 2001 Edition with Addenda through OMb-2003. Appendix I of the OM Code, Paragraph 1-3410(d), requires that valves that have been maintained or refurbished in place, removed for maintenance and testing, or both, and reinstalled, shall be remotely actuated at reduced or normal system pressure to verify the open and close capability of the valve before resumption of electric power generation. The proposed revision to RR-V-4 identifies that, beginning with the refueling outage in Spring 2011 (2R21), HNP plans to replace the 2-stage Main Steam Safety Relief Valves (SRVs) with 3-stage SRVs. Therefore, RR-V-4 has been updated to include the alternative testing proposed for the 3-stage SRVs.

This letter contains no NRC commitments. If you have any questions, please contact Jack Stringfellow at (205) 992-7037.

Respectfully submitted, P. M. Marino Vice President - Engineering PMM/PAH/emm LLC)477

U.S. Nuclear Regulatory Commission Log: NL-11-0162 Page 2

Enclosure:

RR-V-4, Version 2.0

References:

1. RR-V-1 1 (Third 10 Year Inspection Interval), NRC Safety Evaluation dated September 5,1997 (TAC Nos. M99485 and M99486).

2. Revision to RR-V-1 1 approved via NRC safety Evaluation dated February 21, 2003 (TAC Nos. MB6655 and MB6656).

3. RR-V-4, Version 1.0 (Fourth 10 Year Inspection Interval), NRC safety Evaluation dated February 14, 2006 (TAC Nos. MC6837, MC6838, MC7626, and MC7627).

cc: Southern Nuclear Operating Company Mr. J. T. Gasser, Executive Vice President Mr. D. R. Madison, Vice President - Hatch Mr. M. J. Ajluni, Nuclear Licensing Director RTYPE: CHA02.004 U. S. Nuclear Requlatory Commission Mr. V. M. McCree, Regional Administrator Mr. P. G. Boyle, NRR Project Manager- Hatch Mr. E. D. Morris, Senior Resident Inspector - Hatch

Edwin I. Hatch Nuclear Plant Revision to Relief Request RR-V-4 Enclosure RR-V-4, Version 2.0

SOUTHERN NUCLEAR OPERATING COMPANY IST PROGRAM PROPOSED ALTERNARIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(ii)

RR-V-4, VERSION 2.0 VERSION 2.0 CHANGES ARE ENCLOSED BY BRACKETS "[ I" FOR EASY IDENTIFICATION PLANT/UNIT: Edwin I Hatch Nuclear Plant/Unit 1 and 2.

4 th Interval beginning January 1, 2006 and ending December 31, 2015.

INTERVAL:

COMPONENTS 1B21-F013A, B, C, D, E, F, G, H, J, K, & L AFFECTED: 2B21-F013A, B, C, D, E, F, G, H, K, L, & M CODE EDITION ASME OM Code-2001 Edition with Addenda through OMb-2003 AND ADDENDA:

REQUIREMENTS: Appendix I, paragraph 1-3410(d) of the OM Code requires that valves that have been maintained or refurbished in place, removed for maintenance and testing, or both, and reinstalled shall be remotely actuated at reduced or normal system pressure to verify open and close capability of the valve before resumption of electric power generation.

REASON FOR This alternative is a re-submittal of NRC approved 3 rd Interval relief REQUEST: request RR-V-1 1 that was based on Appendix I of the ASME OM Code-1995 Edition, no addenda. This 4th Interval request for relief, RR-V-4, is based on Appendix I of the ASME OM Code-2001 Edition with Addenda through OMb-2003. There have been no substantive changes to this alternative, to the OM Code requirements or to the basis for use, which would alter the previous NRC Safety Evaluation conclusions. (See References for SER date and TAC numbers associated with RR-V-11)

[Plant Hatch plans to replace the existing Target Rock 2-Stage Pilot Operated Safety Relief Valves (SRVs) with Target Rock 3-Stage Pilot Operated SRVs and the specifics of the proposed alternative testing are slightly different. Therefore, this relief request has been revised to incorporate information specific to the replacement 3-Stage SRVs, as necessary.]

Exercising the main disc of the [2-Stage or 3-Stage] SRVs after reinstallation can only be performed during reactor startup when there is sufficient steam pressure to actuate the main disc. Past history [for the 2-Stage SRVs] indicates that the main and pilot discs routinely do not re-seat properly after being exercised during reactor startup resulting in steam leakage into the suppression pool. This leakage results in a decrease in plant performance and the potential for increased suppression pool temperatures which could force a plant shutdown to repair a leaking SRV. Past operating history [of the 2-Stage SRVs] indicates that the exercising performed during reactor startup is of no significant benefit in ensuring the proper operation of the individual SRV assemblies.

RR-V-4, Ver. 2.0 12-5a Version 2.0

SOUTHERN NUCLEAR OPERATING COMPANY IST PROGRAM PROPOSED ALTERNARIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(ii)

RR-V-4, VERSION 2.0 VERSION 2.0 CHANGES ARE ENCLOSED BY BRACKETS "[ I" FOR EASY IDENTIFICATION

[Existing] System Description The Unit 1 and Unit 2 SRVs are the Target Rock 2-Stage, Model 7567F design. The SRVs are dual-function valves capable of being independently opened in either the safety or the relief mode of operation.

A total of 11 SRVs are installed on each unit. In the safety mode of operation, each SRV opens when system pressure exceeds the valve's set-point pressure, which is controlled by pre-compression of the set-point spring acting down on the pilot disc.

Venting the volume on the reactor side of the pilot disc creates a differential pressure across the main piston, thereby providing a force to open the main disc and relieve system overpressure. Hence, reactor vessel steam is allowed to flow directly through the main disc to seat opening and to the suppression pool via the discharge piping. All 11 SRVs operate in the safety mode, which provides the safety function of over-pressurization protection. The requirements for this mode are listed in Technical Specification 3.4.3.

In the relief mode of operation, each SRV is opened by an electro-pneumatic actuator, which consists of a three-way solenoid valve, an attachment manifold, and a pneumatic operator. When the solenoid valve is energized, pneumatic pressure is routed into the operator to lift the pilot rod against the force of the compressed set-point spring. This allows system pressure to lift the pilot disc, venting the volume on the reactor side of the disc, and opening the valve as in the safety mode discussed above.

This mode of operation is used for Automatic Depressurization System (ADS), Low-Low-Set (LLS), and remote manual operation. Technical Specifications 3.5.1 and 3.6.1.6 provide requirements for the ADS and LLS System. Manual operation is not safety related and is not addressed by Technical Specifications. In each unit, seven SRVs are part of ADS, while the remaining four constitute LLS.

[Replacement System Description The Unit 1 and Unit 2 SRVs are being replaced with Target Rock 3-Stage, Model 0867F-001/09G-001 design. The new SRVs are also dual-function valves capable of being independently opened in either the safety or the relief mode of operation. A total of 11 SRVs will be installed on each unit. In the safety mode of operation, each SRV opens when system pressure exceeds the valve's set-point pressure, which is controlled by compression of the set-point spring acting on the pilot disc.

RR-V-4, Ver. 2.0 12-5b Version 2.0

SOUTHERN NUCLEAR OPERATING COMPANY IST PROGRAM PROPOSED ALTERNARIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(ii)

RR-V-4, VERSION 2.0 VERSION 2.0 CHANGES ARE ENCLOSED BY BRACKETS "[ I" FOR EASY IDENTIFICATION Venting the volume on the reactor side of the pilot disc provides pressure to lift the 2 nd Stage disc. Lifting the 2 nd Stage disc vents the volume on the reactor side of the 2nd Stage disc which creates a differential pressure across the main piston resulting in opening the main disc to relieve system overpressure. Hence, reactor vessel steam is allowed to flow directly through the main disc to seat opening and to the suppression pool via the discharge piping. All 11 SRVs operate in the safety mode, which provides the safety function of over-pressurization protection. The requirements for this mode are listed in Technical Specification 3.4.3.

In the relief mode of operation, each SRV is opened by an electro-pneumatic actuator, which consists of a three-way solenoid valve, an attachment manifold, and a pneumatic operator. When the solenoid valve is energized, pneumatic pressure is routed into the operator to lift the 2nd Stage disc which vents the volume on the reactor side of the 2 nd Stage disc.

This creates a differential pressure across the main piston resulting in opening the main disc to relieve system pressure. This mode of operation is used for Automatic Depressurization System (ADS), Low-Low-Set (LLS), and remote manual operation. Technical Specifications 3.5.1 and 3.6.1.6 provide requirements for the ADS and LLS System. Manual operation is not safety related and is not addressed by Technical Specifications. In each unit, seven SRVs are part of ADS, while the remaining four constitute LLS.]

[Historical] Testing [of 2-Stage SRVs] at Plant Hatch Testing of Plant Hatch SRVs is performed to satisfy Technical Specifications Surveillance Requirements (SRs) and the ASME OM Code-2001, "Code for Operation and Maintenance of Nuclear Power Plants." Certain tests are performed with the SRVs installed (in situ),

while others are performed as "bench tests" after the valve is removed and transported to a maintenance and testing facility. Current requirements are as follows:

1. SRs 3.5.1.12 and 3.6.1.6.1 provide SRV manual actuation testing requirements for the ADS and LLS Functions, respectively, to demonstrate operability of the SRV relief mode.

2. Remote manual actuation is also required by the ASME OM Code, Appendix I, paragraph 1-3410(d), to verify open and close capability of the valve before resumption of electric power generation. This applies to valves that have been either maintained or refurbished in place, or removed for maintenance and testing and reinstalled. This remote manual actuation is performed at zero system pressure.

RR-V-4, Ver. 2.0 12-5c Version 2.0

SOUTHERN NUCLEAR OPERATING COMPANY IST PROGRAM PROPOSED ALTERNARIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(ii)

RR-V-4, VERSION 2.0 VERSION 2.0 CHANGES ARE ENCLOSED BY BRACKETS "[ I" FOR EASY IDENTIFICATION Plant Hatch currently meets the two above testing requirements by opening and closing each SRV by defeating the control room switches and leakrate testing the pilot air operators and associated accumulator piping. Valve opening and closing capability is then confirmed by measuring the change in depth of the pilot rod.

3. Plant Hatch Units 1 and 2 Technical Specifications SRs 3.5.1.11 and 3.6.1.6.2 require that the SRVs be opened on an actual or simulated automatic initiation signal to demonstrate that the solenoids operate when initiated by a signal. Actual valve actuation is excluded from these tests which are performed on a once per operating cycle frequency.

Plant Hatch currently meets the above testing requirement by performing the test in conjunction with Logic System Functional Tests (LSFT) for the initiating instrument logic, which are also required by Technical Specifications.

4. ASME OM Code-2001, 1-3310 (d) and (e) require that SRV auxiliary components be tested in place as follows: solenoid valve and pneumatic actuator integrity is verified by performance of leak rate tests, and solenoid valve electrical function is verified.

Plant Hatch satisfies the above requirement by tests performed following maintenance on the valves which demonstrate operability of the valve pneumatic actuation system.

[Historical] Testing [of 2-Stage SRVs] at Outside Facilities During each refueling outage, all 11 SRV pilot assemblies and approximately one-third of the main stages are removed and shipped to Wyle Laboratories for "as-found" testing, which includes visual inspection, leakage testing, pilot disc-to-seat sticking testing, and set pressure testing. The tests are performed on a valve prior to maintenance on the valve. The leakage and set pressure tests are performed at a steam pressure of approximately 1035 psig. Both tests meet the requirements of ASME OM Code-2001, 1-3310 (a), (b), and (c).

Following the "as-found" testing, the SRVs are given a dimensional inspection followed by refurbishment, if required. This work is performed by the valve supplier, Target Rock Corporation.

RR-V-4, Ver. 2.0 12-5d Version 2.0

SOUTHERN NUCLEAR OPERATING COMPANY IST PROGRAM PROPOSED ALTERNARIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(ii)

RR-V-4, VERSION 2.0 VERSION 2.0 CHANGES ARE ENCLOSED BY BRACKETS "[ I" FOR EASY IDENTIFICATION Valve warming for post maintenance testing is performed at a steam pressure of approximately 1010 psig. Post maintenance testing includes initial valve leakage testing, safety mode valve actuation to satisfy requirements for set pressure, reseat pressure, main disc stroke time, and final leakage testing. Final seat leakage tests are performed at approximately 1070 psig. Upon successful test completion, each valve receives written certification from the lab and is returned to Plant Hatch for reinstallation. To receive certification, the valve must have zero seat leakage and meet the acceptance criteria for set pressure. These tests meet the requirements of ASME OM Code-2001, 1-3310 and Technical Specifications SR 3.4.3.1 (for lift set-point pressure verification).

General Change Justification Leaking SRVs result in the following challenges to Plant Hatch components and operation:

1. Leakage during operation may cause the valve to inadvertently actuate, possibly resulting in an unplanned plant shutdown, with its attendant challenges to plant safety systems and components. This has occurred previously on at least one domestic BWR plant.

2. Leaking SRVs create operational problems associated with the suppression pool. SRV leakage increases both pool temperature and level, requiring more frequent use of the suppression pool cooling mode of the Residual Heat Removal (RHR) system.

3. Plant efficiency is impacted because the transfer of heat to the suppression pool is a source of thermal heat loss from the power generation steam cycle, thereby reducing electrical generating capacity. SRV leakage results in radiological challenges since radioactive nuclides contained in the steam can become a potential source for personnel contamination.

RR-V-4, Ver. 2.0 12-5e Version 2.0

SOUTHERN NUCLEAR OPERATING COMPANY IST PROGRAM PROPOSED ALTERNARIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(ii)

RR-V-4, VERSION 2.0 VERSION 2.0 CHANGES ARE ENCLOSED BY BRACKETS "[ I" FOR EASY IDENTIFICATION As described previously, each [2-Stage] SRV pilot assembly and approximately one-third of the main stages are bench tested at Wyle Laboratories during each refueling outage. The valves are refurbished as necessary to meet the acceptance criteria of zero leakage, and are certified in writing as being leak free. The valves are then reinstalled in the plant and proper pilot operation is confirmed through leakrate testing of the pilot air operators and associated accumulator piping and in situ measurements of the pilot rod movement. Following this surveillance test, Plant Hatch has typically experienced one or more leaking valves from what was originally a leak-free population supplied by the vendor (Wyle Laboratories). For example, Plant Hatch Unit 1 was shutdown in February 2002 due to Main Condenser Off-Gas System problems.

During this forced outage, three SRVs were replaced with leak-tight valves, which were actuated as described above. One of the replacement SRVs then began leaking following startup.

Several aspects of [2-Stage or 3-Stage] SRV design and operation can contribute to valve leakage. As mentioned earlier, these include test pressure, pilot valve disc and rod configuration, and system and valve cleanliness. Actuation of the SRVs after laboratory testing by any means allows these contributors to impact the ability of the valve to re-close completely. Plant Hatch has made significant efforts to minimize the effects of these contributors. However, elimination of in situ valve testing under any condition that disturbs the pilot disc/seat interface is expected to have the most positive impact in reducing SRV leakage.

Additionally, reducing challenges to the SRVs is a recommendation of NUREG-0737, "TMI Action Plan Requirements" item II.K.3.(16). This recommendation is based on a stuck open SRV being a possible cause of a Loss of Coolant Accident (LOCA). This submittal is consistent with that NRC recommendation.

PROPOSED [2-Stage SRVs1 ALTERNATIVE As an alternative to the testing required by ASME OM Code-2001, AND BASIS: Appendix I, paragraph 1-3410(d), SNC proposes to actuate the [2-Stage]

SRVs in the relief mode at the test facility (i.e., Wyle Laboratory). The solenoid valve will be energized, the actuator will stroke, and the pilot rod lift will be measured. This test will verify that, given a signal to energize the solenoid, the pilot disc rod will lift. The rod movement measurement will be performed using calibrated equipment and will be recorded in the test documentation package for future reference, as needed.

RR-V-4, Ver. 2.0 12-5f Version 2.0

SOUTHERN NUCLEAR OPERATING COMPANY IST PROGRAM PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(ii)

RR-V-4, VERSION 2.0 Alternate testing is justified since the remaining segments of the SRV mode of operation are proven by other tests. The ability of the pilot disc to open is shown in the safety mode actuation bench test. The integrity of the pneumatic and solenoid system for the SRVs are verified by performance of post maintenance leakrate testing and the "click" test, respectively. Automatic valve actuation is proven operable by logic system functional tests which include verification that the solenoid actuates from the automatic signal.

Each refueling outage, all 11 pilot assemblies and approximately one-third of the main disc assemblies are sent to Wyle Laboratories and tested with steam pressure. As a result, even though actual valve movement is not performed after the SRV is re-installed in the plant, all pilot assemblies are tested with steam pressure once per cycle and all the main discs are tested with steam pressure approximately once every three cycles. This testing adequately demonstrates SRV operational readiness.

[3-Stage SRVs As an alternative to the testing required by ASME OM Code-2001, Appendix I, paragraph 1-3410(d), SNC proposes to actuate the 3-Stage SRVs in the relief mode at the vendor test facility (i.e., Wyle Laboratory) each operating cycle. In addition to vendor testing, after installation the solenoid valve will be energized, resulting in actuator stroke and the air actuator rod movement will be visually confirmed for each SRV. This test will verify that, given a signal to energize the solenoid, the 2 "dStage disc will lift.

Alternate testing is justified since the remaining segments of the SRV mode of operation are proven by other tests. The ability of the pilot disc to open is shown in the safety mode actuation (bench) test. The integrity of the pneumatic and solenoid systems for the SRVs are verified by performance of post maintenance leakrate testing and verification of electrical and pneumatic connections through mechanical/electrical testing after installation but prior to electrical power generation.

Automatic valve actuation is proven operable by logic system functional tests which includes verification that the solenoid actuates from the automatic signal.

Each refueling outage, all 11 SRV assemblies will be sent to Wyle Laboratories and bench tested with steam pressure. As a result, even though actual valve movement is not performed after the SRV is re-installed in the system; all SRVs are tested with steam pressure once per cycle. This testing adequately demonstrates SRV operational readiness.]

RR-V-4, Ver. 2.0 12-5g Version 2.0

SOUTHERN NUCLEAR OPERATING COMPANY IST PROGRAM PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(ii)

RR-V-4, VERSION 2.0 The above proposed surveillance and testing of the [2-Stage and 3-Stage]

SRVs and associated components provide reasonable assurance of adequate valve operation and readiness. On the basis that compliance with OM Code testing requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety, this proposed alternative should be granted pursuant to 10 CFR 50.55a(a)(3)(ii).

DURATION: [Remainder of] 4 th IST Interval (January 1, 2006 through December 31, 2015).

PRECEDENTS: This Relief Request was approved [for the 2-Stage SRVs] as RR-V-11 for the Third 10 Year IST Interval.

[RR-V-4 for the 2-Stage SRVs was approved for the Fourth 10 Year IST Interval.]

REFERENCES:

[Third 10 Year Interval] NRC Safety Evaluation dated September 5, 1997 - TAC Nos. M99485 and M99486. Subsequent revision to RR-V-11 was approved via NRC Safety Evaluation dated February 21, 2003 -

TAC Nos. MB6655 and MB6656.

[Fourth 10 Year Interval NRC Safety Evaluation dated February 14, 2006

- TAC Nos. MC6837, MC6838, MC7626 and MC7627.]

STATUS: [Version 2.0 for 3-Stage SRVs submitted to NRC for approval.]

RR-V-4, Ver. 2.0 12-5h Version 2.0