ML20045G895
| ML20045G895 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 07/07/1993 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20045G894 | List: |
| References | |
| NUDOCS 9307160115 | |
| Download: ML20045G895 (5) | |
Text
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Enclosure
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E UNITED STATES
-l NUCLEAR REGULATORY COMMISSION W ASHINGTON. D.C. 20SWOOO1 SAFETY EVALVATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO THE INSERVICE TESTING PROGRAM RE0 VESTS FOR RELIEF NORTHEAST NUCLEAR ENERGY CORPORATION MILLSTONE NUCLEAR POWER STATION. UNIT 3 DOCKET N0. 50-423
1.0 INTRODUCTION
The Code of Federal Regulations,10 CFR 50.55a, requires that inservice testing (IST) of certain ASME Code Class 1, 2, and 3 pumps and valves be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable addenda, except where relief has been requested by the licensee and alternatives authorized or relief granted by the Commission pursuant to Sections (a)(3)(i), (a)(3)(ii), or (f)(6)(1) of 10 CFR 50.55a.
In-proposing alternatives or requesting relief, the licensee must demonstrate that: (1) the proposed alternatives provide an acceptable level of quality and safety; (2) compliance would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety; or (3) conformance is impractical for its facility.
NRC guidance contained in Generic Letter (GL) 89-04, " Guidance on Developing Acceptable Inservice Testing Programs,"
provided alternatives to the Code requirements detennined acceptable to the staff.
The Code of Federal Regulations,10 CFR 50.55a, authorizes the Commission to grant relief from ASME Code requirements upon making the necessary findings.
The NRC staff's findings with respect to granting or not granting the relief requested as part of the licensee's IST program are contained in this Safety Evaluation (SE).
2.0 BACKGROUND
The requested relief (CFR-32) is for a temporary extension of the biennial position indication testing for certain valves for which position indication can be verified only during cold shutdowns or refueling outages. As a result of an unusually long maintenance outage for service water system and erosion /
corrosion activities, the refueling outage was delayed from November 1992 to l
July 31, 1993.
Thus, the 2-year period for performing the position indication verification will be exceeded for certain valves.
The last refueling outage started in February 1991 with the position indication verification commencing on February 4, 1991.
Applying the allowed 25% extension to the test interval, the verification for specific valves will be due during the period between August 4, 1993, and September 24, 1993. The relief request proposes to perform all position indication verifications by September 30, 1993, and then return the valves to the required 2-year schedule.
The relief requests mentioned in the basis for relief for CSR-32 were approved in NRC's January 15, 1988, SE.
9307160115 930707 1
DR ADOCK 0500 3
_ _ 3.0 EVALVATION The licensee has requested relief from the requirements of IWV-3300 for verification of position indication every 2 years. The valves which are the subject of this relief request are listed in the following table, with associated relief request (R/R) numbers.
Valve Number Actuator Number R/R
. Function-3CHS-V393 3CHS-HV8109A 17 A RCP seal injection isolation 3CHS-V433 3CHS-HV81098 17 B RCP seal injection isolation 3CHS-V466 3CHS-MV8109C 17 C RCP seal injection isolation 3CHS-V500 3CHS-HV8109D 17 0 RCP seal injection icolation 3CHS-V532 3CHS-HV8112 17 RCP seal return containment isolation 3CHS-V533 3CHS-HV8100 17 RCP seal return containment isolation 3CHS-V700 3CHS-HCVl90A B train alternate charging control 3CHS-V705 3CHS-HCVl90B A train alternate charging control A train PORV block 3RCS-V167 3RCS-HV8000A 3RSC-V168 3RCS-PCV455A 3
A train PORV 3RCS-V169 3RCS-HV80008 B train PORV block 3RCS-V170 3RCS-PCV456 3
B train PORV 3RCS-V174 3RCS-AV8145 4
Alternate pressurizer spray isolation 1
3RHS-V004 3RHS-HV8716A 26 A train RHS pump discharge isolation 3RHS-V008 3RHS-MV8716B 26 B train RHS pump discharge isolation 3RHS-V994 3RHS-MV8702A 11 B RHR pump suction from RCS 3RHS-V995 3RHS-MV8701B 11 A RHR pump suction from RCS 3RHS-V996 3RHS-HV8702B 11 B RHR pump suction from RCS 3RHS-V997 3RHS-HV8701A 11 A RHR pump suction from RCS 3RHS-V998 3RHS-HV8702C 11 B RHR pump suction from RCS
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9 l Valve Number Actuator Number R/R Function 3RHS-V999 3RHS-HV8701C 11 A RHR pump suction from RCS 3SIH-V003 3SIH-HV8801A 12 A train SIH injection from CHS l
3SIH-V004 3SIH-HV8801B 12 B train SIH injection from CHS 3SIH-V010 3SIH-HV8806 13 SIH common suction from RWST f
3SIH-V020 3SIH-HV8835 27 SIH common injection to RCS CL CHS check valve test connection 3SIH-V065 3SIH-CV8843 A train SIH injection to RCS HL 3SIH-V093 3SIH-MV8802A B train SIH injection to RCS HL 3SIH-V100 3SIH-MV8802B SIH check valve test connection i
3SIH-V102 3SIH-CV8881 SIH check valve test connection 3SIH-V103 3SIH-CV8824 SIH check valve test connection 3SIH-V104 3SIH-CV8823 3SIH-V962 3SIH-HV8813 29 SIH pump common recirc.
3SIH-V988 3SIH-CV8871 SIH check valve test CIV 3SIL-Vll 3SIL-HV8809B 24 B train normal LPSI injection isolation 35WP-V25 35WP-HV054A 14 A RSS HX isolation i
35WP-V27 35WP-HV054C 14 C RSS HX isolation 3SWP-V33 35WP-HV050A 28 A train SWP supply to CCP HX 35WP-V58 3SWP-HV0548 14 B RSS HX isolation 3SWP-V60 3SWP-HV054D 14 D RSS HX isolation 3SWP-V65 3SWP-HV050B 28 B train SWP supply to CCP HX CDTT vent to radioactive gaseous 3VRS-V2 3VRS-CTV20 i
waste The Northeast Nuclear Energy Company (licensee) indicates in the letter of April 13, 1993, that the Millstone Unit 3 Final Safety Analysis Report (FSAR) states that the mispositioning of a motor-0)erated valve, due to a malfunction in the control circuitry in conjunction wit 1 an accident, has been previously analyzed and found to be a very-low probability event.
In addition, power l
lockouts are provided in the control room for each valve to limit operator j
actions which could result in mispositioning a valve.
The licensee has l
determined that the lack of remote position indication does not impact the ability of the safety systems to perform their intended functions for the associated valves. Valves with relief requests listed in the table above
. cannot be practically stroked during power operations.
Valves 3CHS-V700 and 3CHS-V705 require observation of flow or other positive means to verify actual valve position; however, the systems that contain these valves cannot pass flow, or flow cannot be isolated during normal operation. The remaining valves in the list are located inside containment and the position indication verification requires local observation of the stem movement; therefore, verification is impractical during power operation due to as low as is reasonably achievable (ALARA) and personnel safety concerns.
3.1.1 Licensee's Basis for Relief The licensee states:
Stroke testing of the valves with associated existing relief requests (i.e., CSRs 3, 4, 11, 12, 13, 24, 26, 27, 28, and 29, and R-14, and R-17) during power operation is impossible.
Position indication test (PIT) of valves (identified by #) requires flow observation because the valve stems are not visible.
Flow observation requires that the reactor coolant pump seal flow be stopped, which is not practical. Justification is provided in Relief Request 17. A PIT of valves (identified by *) requires local stem observation inside containment which should not be performed during power operation due to ALARA and safety concerns."
3.1.2 Alternative Testina The licensee proposes:
"Ihese valves will be position indication tested during cold shutdown or refueling (when plant conditions permit), whichever is earlier, but no later than September 30, 1993.
This alternative test applies to Cycle 4 only."
3.1.3 Evaluation ASME Section XI, Subsection IWV-3300, requires that valves with remote position indicators be observed at least once every 2 years to verify operation is accurately indicated.
If the plant Technical Specifications allow, it is acceptable to apply a 25% extension to the 2-year test frequency.
For the listed valves, even with the 25% extension, the allowable time period for the surveillance could be exceeded by as much as 2 months.
The original requirement for the position indication verification is stated in Section XI Interpretation XI-1-89-10 which indicates that the intent was to ensure valve position was accurately indicated for valve stroke time i
measurement rather than for reliance of plant operators in post-accident situations.
For the listed valves that are tested only during cold shutdowns or refueling outages (as indicated by Relief Request Number), there will be no stroke time tests performed unless a plant shutdown occurs.
For the remaining valves, the extension may result in one additional quarterly stroke time test before verification of the accurate indication of position indicating lights.
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. Imposition of the Code requirements would result in a required plant shutdown to complete the verifications by August 4, 1993. This would present a considerable hardship on the licensee in that it would possibly-require the refueling outage to begin prior to the scheduled date, reducing the useful life of the core.
Alternatively, without a plant shutdown, the testing would require systems to be operated in an off-normal configuration and could result in exposing personnel to unsafe conditions in order to observe the valve stem movement.
Delaying the position verification does not significantly impact the safety of the plant in that (1) a valve's capability of operation is not compromised, (2) there are operating parameters that an operator could observe in an accident condition that would indicate the erroneous position indication if both an accident occurred before the verification can be performed and the indicating lights are incorrect, and (3) the extension is for a one-time, short duration.
4.0 CONCLUSION
The staff finds that the proposed extension of the test frequency for verifying the position indication for the listed valves will not endanger the public health and safety.
No significant impact on plant safety occurs from the delay.
Pursuant to 10 CFR 50.55a(a)(3)(ii), the staff finds that the hardship imposed upon the licensee if the required ASME Code test frequency were imposed is disproportionate and would not result in a compensating i
increase in the level of quality and safety.
The licensee may delay the position indication testing of these valves required during Cycle 4 to test them "during cold shutdown or refueling (when plant conditions permit),
whichever is earlier, but no later than September 30, 1993."
Principal Contributor:
P. L. Campbell i
Date:
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