ML20217B119
| ML20217B119 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 09/05/1997 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20217B106 | List: |
| References | |
| NUDOCS 9709230236 | |
| Download: ML20217B119 (4) | |
Text
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UNITED STATES g
j NUCLEAR REGULATORY COMMISSION e
WASHINGTON, D.C. 30006 4001
....e SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.208TO FACILITY OPERATING LICENSF. DPR-57 AND AMENDMENT NO.150ro FACILITY OPERATING LICENSE NPF-5 SOUTHERN NUCLEAR OPERATING COMPANY. INC.. ET AL.
EDWIN 1. HATCH NUCLEAR PLANT. UNITS 1 AND 2 DOCKET NOS. 50-321 AND SQ30$
1.0 INTRODUCTION
By letter dated January 7,1997, as supplemented July 2,1997, Georgia Power Company and Southem Nuclear Operating Company, Inc, et al. (the licensee) proposed license amendments to change the Technical Specifications (TS) for the Edwin 1. Hatch Nuclear Plant, Units 1 and 2. The proposed changes are associated with surveillance testing that would require manually actuating every safety / relief valve (S/RV) during each unit startup from a refueling outage. The proposed changes provide an attomate method of testing of the S/RV during shutdown conditions rather than during unit startup as is currently done. The July 2,1997, letter provided clarifying information that did not change the initial proposed no significant hazards consideration determination.
2.0 EVALUATION in its letter dated January 7,1997, the licensee proposed a change to modify the S/RV TS surveillance requirement to perform manual actuations once every 18 months as part of startup testing activities. The licensee supplemented this request with a letter dated July 2, 1997, which provided additional information regarding the testing proposed for the plant-S/RVs. The specific TS change evaluated herein is for TS Surveillance Requirements (SRs) 3A.3.2,3.5.1.12, and 3.6.1.6.1. Current relief mode SR 3.5.1.12, for the Automatic Depressurization System (ADS), and SR 3.6.1.6.1, for the Low-Low Set (LLS) system, require that each S/RV be actuated at pressure conditions. The licensee proposes to revise these SRs to require the S/RVs to be manually actuated in the relief mode during an outage before steam is generated. The licensee also proposes to delete safety mode SR 3A.3.2, which requires that each S/RV be manually actuated.
3.0 BACKGROUND
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 installed on each of the Hatch units' main steam systems, all of which operate in the safety mode for overpressure protection. In the safety mode, each S/RV opens when system pressure exceeds the self-actuating setpoint pressure, which is controlled by the setpoint spring acting down on the pilot disk. When the pilot disk g923023697"O905
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2-opens, the resulting differential pressure across the main piston opens the main disk to relieve syMem overpressure. The relief mode ADS and LLS functions are accomplished by an automatic control circuit that applies electric power to volenoids 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 Hatch, Units 1 and 2, seven S/RVs are for the ADS, and the remainirig four are for the LLS.
Currently, both Hatch Units 1 and 2 TS SRs 3.4.3.2,3.5.1.12, and 3.6.1.6.1 require that, at least once every 18 months, the S/RVs be functionally exercised with reactor steam pressure.
This testing is performed with system pressure of at least g20 psig during reactor startup following an outage. The licensee has linked this testing to leakage of the valves, providing several examples of instances where these S/RVs began to leak after the in situ htroke l
testing was performed. The licensee states that, if the pilot stage leakage is severe enough, the S/RV setpoint could drift and lead to spurious actuation and/or failure of the valve to ressat. 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.
3.1 Proposed Technical Soecihtion Chances The licensee proposes to revise TS SRs 3.5.1.12 and 3.6.1.6.1 and to eliminate TS SR 3.4.3.2. This would allow S/RV functional exercising to be performed before reactor steam is generated..The licensee states that with the proposed changes, 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 disk would not be lifted. The licensee states that the ability of the plot disk to open would be shown by the required safety mode actuation performed by a bench test Further, the licensee states that all 11 S/RVs and three or fear 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 5-year frequency. This testing also verifes 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 the combination of both the testing at Wyle Labs and that performed after the valves are reinstalled, completely demonstrates operability of the S/RVs.
3.2 Staff Evaluation The staff has reviewed the licensee's proposed TS changes and agrees that the current TS requirement to perform the 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 reseat, increased use of RHR for suppression pool cooling, decreased generating capacity, and increased radiation hazard. The testing proposed by the licensee
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3-1 provides periodic verification 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 ADS S/RVs and associated components provide aswrence of adequate valve operation.
One difference between the current TS-required stroking of the S/RV main stages during plant startup and the licensee's proposal is that, with the proposed testing, there would be less frequent stroking of the S/RV main stages. Instead of stroke testing the S/RV main stages after each refueling outage, only three or four main stages would be stroked each-refueling outage, which would result in an approximate 5-year frequency. However, because the main stage disks of these valves have a proven history of reliable performance at bolling water reactor plants, the staff agrees that the proposed stroking of three or four main stage -
disks each refueling is adequate.
Amner difference between the current TS-required stroking and the licensee's proposal is that, when performing the testing in situ as required by the current TS, the testing verifies that the S/RV discharge line is not blocked. However, the licensee stated that thero are foreign material exclusion controls in place at the plant which, together with the horizontal orientation of the discharge line mating connections, provide reasonable assurance that no obstruction exists in the lines. The staff agrees that the likelihood of blockage of an S/RV discharge line is remote as demonstrated by operational history and that the licensee has acceptably addressed this concem.
Based on the above evaluation, the staff concludes that the licensee has demonstrated the adequacy of the proposed changes to the Hatch Nuclear Plant, Units 1 and 2, TS. The proposed changes provide for testing of the S/RVs to demonstrate proper operation without the need for in situ stroking of the S/RV main stages with reactor steam. Therefore, the proposed changes to TS SRs 3.4.3.2,3.5.1.12, and 3.6.1.6.1 for Hatch Nuclear Plant, Units 1 and 2, are acceptable.
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3.0 STATE CONSULTATION
in accordance with the Commission's regulations, the Georgia State official was notifed of the proposed issuance of the amendments. The State official had no comments.
4.0 ENVIRONMENTAL CONSIDERATION
The amendments change surveillance requirements. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding (62 FR 4350 dated January 29, 1997). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.
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5.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to I
the common defense and security or to the health and safety of the public, Principal Contributor, G. Hammer Date:
September 5, 1997
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