ML20205F934

Safety Evaluation Authorizing Licensee 990108 Relief Request PV-38,from Requirements of ASME BPV Code Section XI IST Testing,Valve Program for Plant,Units 1,2 & 3

ML20205F934
Person / Time
Site:	Browns Ferry
Issue date:	04/01/1999
From:	NRC (Affiliation Not Assigned)
To:
Shared Package
ML20205F927	List: ML20205F923 ML20205F934
References
NUDOCS 9904070061
Download: ML20205F934 (4)

Text

o amou yo -% UNITED STATES o 3 is NUCLEAR REGULATORY COMMISSION WASHINGTON, D.c. 20555-0001

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO THE INSERVICE TESTING PROGRAM RELIEF REQUESTS BROWNS FERRY NUCLEAR PLANT UNITS 1. 2. AND 3 DOCKET NUMBERS 50-259. 50-260. AND 50-296 _

1.0 INTRODUCTION

The Code of Federal Requbtic.r.g,10 CFR 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 Section XI of the ASME Boiler and Pressure Vessel Code (the Code) and applicable addenda, except where alternatives have been authorized or i

relief has been requested by the licensee and granted by the Commission pursuant to '

Sections (a)(3)(i), (a)(3)(ii), or (t)(6)(i) of 10 CFR 50.553. 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. Section 50.55a authorizes the Commission to approve alternatives and to grant relief from ASME Code requirements upon making the necessary findings. U.S.

Nuclear Regulatory Commission (NRC) guidance contained in Generic Letter (GL) 89-04,

" Guidance on Developing Acceptable Inservice Testing Programs," provides alternatives to the Code requirements determined acceptable to the staff. Furthe. guidance was given in GL89-04, Supplement 1, NUREG-1482, " Guidelines for Inservice Testing at Nuclear Pow; r Planta" Browns Ferry Nuclear Plant (BFN) Units 1,2, and 3 are currently implementing the second ten-year IST interval. This interval began on Septembar 1,1992, and is scheduled to end on August 31. 2002. The Code of record for the IST Valve Program is the 1989 Edition of the {

ASME Section XI Code which incorporates Operations and Maintenance (OM) Stanaard Part 10 for IST of valves.

The NRC's findings with respect to authorizing an alternative or granting or denying the IST Valve Prog.am relief request are given below. i l

2.0 RELIEF REQUEST NUMBER PV-38 The licensee proposes an alternative to the disassembly requirements of OM-10 Paragraph  !

4.3.2.4(c) for 32 check valves in the system that supplies emergency equipment cooling water  ;

to the diesel generators. It is proposed that the interval between sample disassembly of each j check valve within a group of four be extended from 18 to 24 months. '

9904070061 990401 PDR P ADOCK 05000259 pog l Enclosure

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l 2.1 Licensee's Basis for Reauestina Relief The licensee states:

The valves are verified open quarterly by flow testing. System design prevents the valves from being verified closed by reverse flow. Each diesel generator cooler supply line utilizes two check valves in series without an intermediate tap Attempts' to detect  ;

closure of these check valves with non-intrusive test methods have been unsuccessful.

The isolation valves used to force the valves closed are not capable of quick opening or l

closing (the valves have handwheel operators). This means that most of the check l valve disks contact their seat lightly, generating insufficient noise to be detected by an acoustic monitor. The valves also utilize a soft-seat sealing material which dampens I seat contact noise. Noise in the system masks whatever closing signals are generated, and the close proximity of the series check valves would make it extremely difficult to identify closure of individual check valves in the best of conditions. The test equipment purchased to perform non-intrusive testing cannot obtain an ultrasonic trace of disk position for stainless steel check valves, which includes these check valves. Non-intrusive testing of these valves has, therefore, been determined to be impractical due to the inability to obtain repeatable acoustic and ultrasonic signals. Radiography has also been determined to be impractical after a review of test shots on similar check valves. The radiographs that were reviewed did not reveal sufficient detail to determine valve closure. Test equipment for detecting disk position using magnetics has not been purchased and was therefore unavailable for use in determining feasibility.

Since these check valves were changed from carbon steel to stainless steel material in the mid-1980's, none of the valves have failed a visualinspection during periodic disassembly. Through the Unit 2 Cycle 6 (May 1993) refueling outage the check valves were disassembled and inspected once per operating cycle. After the Unit 2 Cycle 6 refu3 ling outage, the check valves were grouped in fours in accordance with GL 89-04 Position 2. Since this was done, at least one check valve from each group has been c"sassembled and inspected on a rotating basis during each operating cycle.

Inspection results have detected only slignt discoloration of the valve disks. No corrosion, pitting, or other adverse conditions have been detected. The only maintenance required on the valves is an occasional replacement of the closure assist spring. Per the vendor, this spring is not required for the valve to close in liquid service under backflow conditions. Free motion of the valves has been unimpeded in all cases.

A review of the Equipment Performance and information Exchange System, previously the Nuclear Plant Reliability Data System (NPRDS), for this type of check valve has deterrnined that there have been no failures of valve function for this type and size of check valve in this particular application. The only failure listed for this particular application and type was for a broken closure assist spring.

A review of the Electric Power Research Institute (EPRI) Application Guide for Check Valves in Nuclear Power Plants for location-related problems has determined that there is no potential for problems resulting from the location of the valves. There is more b

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l i l than enough straight pipe upstream of the valves to preclude any turbulent flow concerns. The flow velocities in the system are appropriate for this type of check valve.

Extension of the inspection frequency for any particular check valve in a group is consistent with GL 89-04 Position 2, currently centained in NUREG-1482, provided (1) the NPRDS database is searched for similar valves, (2) the location of the valves is addressed with regard to the EPRI Guide for Check Valves in Nuclear Power Plants and (3) the disassembly of the valves is documented in detail regard:ng condition and ability to full-stroke. Previously, BFN has disassembled these check valves on a once per operating cycle frequency in accordance with Relief Request PV-14. With a yearly diesel generator outage, disassembly of the check valves once per operating cycle was easily incorporated into the diesel generator maintenance outages. BFN has recently changed the schedule for diesel generator maintenance outages from an annual to a biennial frequency. This means that in order to perform check valve disassembly, the affected diesel generator must be made inoperable for approximately 1Z hours (minimum) every other operating cycle. This reduces the diesel generator availability and does not provide a compensating increase in the safety of the plant (based on the inspection results of the check valves to date).

2.2 Alternate Testina The licensee proposes:

These check valves will be grouped (4 valves maximum) according to design and service conditions and disassembled on a rotating basis in accordance with Position 2 of NRC Generic Letter 89-04 with the exception that the interval between inspections will be scheduled with the diesel generator maintenance outages (biennially). This will extend the current inspection interval from once every 18 months to once every 24 months. Therefore, each valve will be inspected at least once every eight years instead of every six years. Based on the inspection results of the check valves to date, as described above, this interval extension provides an acceptable level of quality and safety.

If a check valve selected for disassembly fails the criteria of GL 89-04, then an evaluation will be performed promptly (generally by the end of the shift during v nich the failure occurs) to determine the failure mechanism and the likelihood that the romaining valves in the test group are affected significantly by this mechanism. The remaining check valves in the affected test group will be disassembled and inspected per GL 89-04 Position 2 criteria using the guidance of GL 91-18. The remaining valves in the test group will be disassembled within a time frame determined by the failure mechann.m evaluation.

3.0 EVALUATION The valves for which the licensee requests relief are in the emergency equipment cooling water (EECW) system. They include the EECW to the Unit 1 diesel generator cooler check valve s 0-C KV-67-507, 508, 514, 515, 521, 522, 528, 529, 624, 625, 627, 628, 630, 631, 634, and 635, and also the EECW to the Unit 3 diesel generator cooler check valves 3-CKV o

693,694,695,696,703,704,705,706,713,714,715,716,723,724,725, and 726. In the open position, these valves pass rated cooling water flow for the Units 1 and 3 diesel generator engines. In the closed position, the valves prevent backflow from the opposite EECW header. OM-10 Paragraph 4.3.2 requires that check valves be exercised to the position required to perform theirintended function at least every refueling outage. As an j alternative, Paragraph 4.3.2.4(c) allows check valves to be disassembled each refueling outage to verify operability.

The licensee's Relief Request PV-14, authorized by the staff's Safety Evaluation dated May 6, 1996, allows a sample disassembly program to be used for these valves. In the authorized alternative testing, valves are placed into groups of four and the interval between disassembly of each valve in the group is 18 months. This interval easily coincides with a yearly diesel generator maintenance outage. The interval between diesel generator maintenance outages has now increased from 12 to 24 months and so the licensee has now proposed extending the interval between valve disassembly to 24 months to once again coincide with diesel generator maintenance outages.

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In accordance with the guidance in Appendix A of NUREG-1482, the licensee developed information to support extension of the valve disassembly and inspection interval. The licensee's basis for relief contains sufficient information conteming:

(a) the condition of the valves and their capability to be full-stroked, (b) a review of industry experience regarding the same type of valve used in similar service, 4 and (c) a review of the installation of each valve addressing the "EPRI Applications Guidelines for Check Valves in Nuclear Power Plants" for problematic locations.

Disassembly of these check valves is not dependent on refueling outages, but on biennial diesel generator outages. Previously, with a yearly diesel generator outage, disassernbly of the check valves once per operating cycle was easily incorporated into the diesel maintenance outages. It wculd be a hardship for the licensee to perform the valve disassembly at a time other than when the dieselis out for maintenance because the affected diesel generator would have to be made inoperable for approximately 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. This would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

The licensee's proposed Mtemative is consistent with the guidance provided in Appendix A of NUREG-1482 and provi es an acceptable level of quality and safety.

4.0 CONCLUSION

The proposed nitemative to the requirements of OM-10 Paragraph 4.3.2.4(c) is authorized pursuant to 10 t/R 50.55a(a)(3)(ii). Compliance with the specified requirements of this I section would result in hardship or unusual difficulty without a compensating increase in the l level of quality and safety.

f Principal Reviewer: M. Kotzalas l

Date:

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