ML20216E967

Safety Evaluation Accepting ,As Suppl by 970902 Request for Relief to Request RR-V-11 Re IST & S/Rv

ML20216E967
Person / Time
Site:	Hatch
Issue date:	09/05/1997
From:	NRC (Affiliation Not Assigned)
To:
Shared Package
ML20216E960	List: ML20216E956 ML20216E967
References
NUDOCS 9709110118
Download: ML20216E967 (3)

Text

men 1 UNITED STATES g NUCLEAR REGULATORY COMMISSION p WASHINGTON, D.C. 30006 4001 SAFETY EVALUATION BY 7tlE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO REllEF REQUEST RR-V 11 REGARDING INSERVICE TESTING OF SAFETY / RELIEF VALVES SOUTHERN NUCLEAR OPERATING COMPANY. INC.. ET AL.

EDWIN 1. HATCH NUCLEAR PLANT. UNITS 1 AND 2 DOCKET NOS, 50-321 AND 50-366

1.0 INTRODUCTION

The Code of Federal Reaulations (10 CFR), Section 50.55a, requires that inservice testing (IST) of certain American Society of Mechanical Engineers (ASME) Code Class 1,2, and 3 pumps and valves be performed in accordance with the ASME Boiler and Pressure Vessel Code (ASME Code) and applicable addenda, except where relief has been requested and granted or proposed attematives have been authorized by the Commission pursuant to 10 CFR 50.55a(f)(6)(l), (a)(3)(i), or (a)(3)(ii). In proposing attematives or requesting relief, the applicant must demonstrate that: . (1) conformance is impractical for its facility; (2) the proposed attemative provides an acceptable level of quality _ and safety; or (3) compliance would result in a hardship or unusual difficulty without a compensating increase in the level _of quality and safety.

2.0 BACKGROUND

By letter dated January 7,1997, as supplemented by letter dated July 2,1997, Georgia

, Power Company and Southem Nuclear Operating Company, Inc, et al. (the licensee) submitted a request for relief RR-V-11 from certain ASME Code IST requirements pertaining to testing of the plant safety / relief valves (S/RVs). The July 2 letter provided additional information regarding the testing proposed for the plant S/RVs. The applicable ASME Code IST requirements for the pint S/RVs are provided in ASME OM Code-1995, Appendix 1, which was approved by the staff in a letter dated August 29,1995, for use at Hatch, Units 1 and 2. Specifically, this request seeks relief from paragraph I 3.4.1(d) of Appendix 1 that requires stroke testing of the S/RV main stages after being maintained or refurbished in place, removed for maintenance or testing, or both, and reinstalled.

Each plant S/RV is a Target Rock 2-Stage pilot-operated S/RV with an attached pneumatic -

~

actuator. - There are a total of 11 S/RVs in each of the Hatch units' main steam systems and are identified below:

Unit 1: 1821-FO13A, B, C, D, E, F, G, H,-J, K, and L Unit 2: 2821-FO13A, B, C, D, E, F, G, H, K, L, and M

....ai4 9709110118 970905 PDR ADOCK 05000321 P- PDR +

.f r  ; These valves have both a safety mode and a relief mode of operation. In the safety mode, each S/RV opens when system pressure exceeds the self-actuating setpoint pressure, wnich is controlled by the setpoint spring acting down on the pilot disk. When the pilot disk opens, the resulting differential pressure across the main piston opens the main disk to relieve system overpressure. The relief mode Automatic Depressurization System (ADS) and Low.

Low Set (LLS) functions are accomplished by an automatic control circuit that applies electric power to solenoidsi which provide control air to the pneumatic diaphragm assembly (i.e.,

auxiliary actuating device) that removes the pilot spring force allowing the pilot disk to open.

Once the pilot disk is open, steam pressure provides the necessary force to open the main S/RV disk. In both Units 1 and 2, seven S/RVs are for the ADS function and the remaining four are for the LLS function.

3.0 BASIS FOR RELIEF Currently, in order to meet plant Technical Specifications and the above OM Code requirements, the Hatch units' S/RV main stages are exercised open and closed with reactor steam pressure at least once every 18 months during startup from a refueling outage. This testing is performed with system pressure of at least g20 psig. The licensee has linked this testing to leakage of the valves, and has provided several examples of instances where these S/RVs began to leak after the in situ stroke testing was performed. The licensee states that, if the pilot stage leakage becomes severe enough, the S/RV setpoint could drift ard : sad to spurious actuation and/or failure of the valve to reseat. The licensee also states that S/RV leakage can cause increases in suppression pool temperature and level and increased use of the Residual Heat Removal (RHR) system for suppression pool cooling. Further, the licensee states that S/RV leakage reduces electrical generating capacity and could increase radiation hazard for personnel.

4.0 PROPOSED ALTERNATIVE TESTING As an attemative to the testing required by the ASME OM Code-1995, Appendix 1, paragraph I 3.4.1(d), the licensee proposes to exercise the S/RVs before reactor steam is generated. The licensee states that the ability of the pilot and main disks to open wo'uld be 4 shown by the required safety mode actuation performed by a bench test. The licensee states that all 11 S/RV pilot stages and three or four main disk assemblies are sent to Wyle Labs and tested with steam pressure each refueling outage. The licensee states that as a result, the main disks are fully stroked and stroke timed at approximately a five-year frequency.

This testing also verifies the resent pressure and closure of the S/RVs. The licensee added that due to the test facility limitations, flow through the main stage is limited such that full flow is not discharged. However, the discharge restriction is not accomplished by impeding the movement of the main disk, but by restricting the size of the discharge path downstream of

- the main disk discharge. The licensee states that after reinstallation of the tested valves, the solenoid valves would be energized, the actuator would stroke, and the pilot disk rod lift -

would be measured,- but that, because there would be no steam pressure, the pilot and main disks would not be lifted. The licensee states that the combination of both the testing at Wyle

. Labs and that performed after the valves are reinstalled, completely demonstrates operability of the S/RVs,

p.rt-

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- 5.0 EVALUATION The staff has reviewed the licensee's relief request and agrees that the ASME Code requirement to perform in situ stroke testing of the S/RVs may contribute to undesirable S/RV

. leakage and could result in spurious actuation of the valves during power operation and/or failure to ressat, increased use of RHR for suppression pool cooling, decreased generating-capacity, and increased radiation hazard. - The altemative testing method proposed by the licensee provides periodic verificati;a of all of the individual S/RV components which are currently being tested except that some tests are to be performed at a test facility instead of in situ with reactor steam, The staff agrees that the proposed surveillance and testing of the S/RVs and associated components provide reasonable assurance of adequate valve operation and readiness, Therefore, the staff finds that the proposed attemative testing--

method to that required by ASME OM Code-1995, Appendix 1, paragraph I 3.4,1(d) is

- acceptable.

6.0 CONCLUSION

1 Based on the above evaluation, the staff concludes that,' pursuant to 10 CFR 50.55a(a)(3)(ii),

the licensee's attemative to ASME OM Code requirements is authorized.- As described above, the staff has determined that the licensee has demonstrated that compliance with the applicable ASME Code testing requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety .

Principal Contributo.: ' G Hammer Date: September 5, 1997 e

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