ML20236H007
| ML20236H007 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 10/26/1987 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20236G993 | List: |
| References | |
| NUDOCS 8711030388 | |
| Download: ML20236H007 (10) | |
Text
. _ _ ._
- ,dua [0, UNITED STATES
[.
5 g* 1 ih NUCLEAR REGULATORY COMMISSION WASHING TON, D. C. 20555 g
- a....
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO D E INSERVICE TESTING PROGRAM AND REQUESTS F_0R RELIEF UNION ELECTRIC COMPANY CALLAWAY NUCLD R PLANT , UNIT 1 DOCKET NO. 50-483 INTRODUCTION Technical Specification 4.0.5 for the Callaway Nuclear Plant, Unit 1, (Callaway) states that inservice inspection of ASME Code Class 1, 2, and 3 pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and Applicable Addenda as required by 10 CFR Part 50, Section 50.55a(g),
except where specific written relief has been granted by the Commission pursuant I to 10 CFR Part 50, Section 50.55a(g)(6)(i). Certain requirements of the applicable Code edition and addenda of Section XI are impractical to perform because of j certain plant system and components designs.
Regulation 10 CFR 50.55a(g)(6)(1) authorizes the Commission to grant relief from these requirements upon making the necessary findings. This Safety Evalua-tion (SE) contains the NRC staff's findings with respect to granting or not granting reliefs submitted as part of the licensee's Inservice Testing Program (IST).
l The Callaway first 10-Year IST program was found acceptable by the staff (with certain exceptions) for implementation by letter dated August 19, 1987.
That review encompassed the initial submittal of January 31, 1984 and subsequent revisions up to and including the April 14, 1987 revision. On July 28, 1987, Union Electric Company (UE) provided a further update of the first 10-year IST program and additional information related to requests for relief from certain code requirements which they determined to be impractical to perform on Callaway.
The update involved one deleted and three revised relief requests, the deletion of three valves from the program, and a justification for cold shutdown testing of one valve. The program for the first 10-year interval is based on the requirements of the 1980 Edition through the Winter of 1981 Addendt of Section XI of the ASME Code and these requirements remain in effect until December 19, 1994 EVALUATION The IST program update and the requests for relief from the requirements of Section XI that have been determined to be impractical to perform have been reviewed by the staff's contractor, EG&G, Idaho, Inc. (EG&G). The Letter Report provided in Attachment 1 is EG&G's evaluation of the licensee's inservice testing program and relief requests.
8711030388 DR 071026 P ADOCK 05000403 pon C____________. ,
a
> I I
-?- j l
The staff has reviewed the Letter Report and concurs with the evaluations and conclusions in the Letter Report. A sunmary of the relief request determinations
- and program changes is presented in Table' 1. The granting of relief is based upon the fulfillment of any commitments made by the licensee in its basis for each relief request and the alternate proposed testing.
CONCLUSION l
Based on the review of the. licensee's IST program revision and relief requests, the staff concludes that the IST program will provide reasonable assurance of the operational readiness of the pumps and valves covered by the IST program to perform their safety-related functions. The staff has determined that t pursuant to 10 CFR 50.55a(g)(6)(i) granting relief where the Code requirements are impractical is authorized by law and will not endanger life or property, or the common defense and security. The staff has also concluded that granting relief is otherwise in the public interest considering the burden that could result if the requirements were imposed on.the facility. During the review of the licensee's revised inservice testing program, the staff has not identified any significant misinterpretations or omissions of Code require-ments. Thus, the IST program update transmitted by letter dated July 28, 1987, is acceptable for implementation. Relief requests contained in any subsequent revisions may not be implemented unless approved by the staff.
Attachment:
Letter Report (EG&G)
Principal Contributor: T. McLellan Date: October 26, 1987
, l 1
l TABLE 1
SUMMARY
OF IST PROGRAM CHANGES AND RELIEF REQUESTS Description of Changes NRC Evaluation Remarks and/or Relief Requests and Conclusions Relief Request AL-2 was Acceptable Licensee will be deleted from the program testing per IWV-3522 i because the licensee !
determined that the check valves listed can ,
be manually exercised while monitoring the opening torque.
j Relief Request AE-1 was Granted The alternative changed to include an methods of testing optional alternate testing the valves consist method to verify the reverse of performing dis-flow closure of valves AE-V120, assembly, inspection l V121, V122, and V123. and manually exer-cising the valve disk or doing a seat leakage test during refueling outages.
Designation EJ-5 was added Acceptable Relief previously because the original granted.
Relief Request designation was inadvertently omitted during the last program l
revision.
Relief Request P01 was revised Granted The change was made to to increase the RHR heat removal bring the gauge tolerance pump discharge pressure gauge in the relief request in tolerance from *5 psig to t10 psig. line with the tolerance specified by the gauge !
manufacturers.
Delete valves EM-HV-8870A,8 Acceptable The valves were inadver-and EM-HV-8883 from IST tently added to the program program, in the last revision, these valves are not required to l be tested.
Cold shutdown justification Acceptable None !
added for valve EM-HV-8835 to allow testing during cold shutdown.
_______ a
7 ATTACHMENT 1 LETTER REPORT, TECHNICAL EVALUATION OF REVISION 9 CHANGES FOR PUMP AND VALVE INSERVICE TESTING PROGRAM, CALLAWAY NUCLEAR PLANT, UNIT 1 Relief Request PO-1 The licensee has revised Relief Request P0-1 to increase the residual heat removal pump discharge pressure gauge tolerance from 5 psig to 10 psig. This change was made to bring the gauge tolerance in the relief request in line with the tolerance specified by the gauge manufacturer. The revised relief request is evaluated below. This evaluation supersedes Section 3.2.1 of the Technical Evaluation Report, Pump and Valve Inservice Testing Program, Callaway Nuclear Plant, Unit 1.
Relief Request. The licensee has requested relief from the instrument accuracy requirements of Section XI, Paragraph IWP-4110, for the residual heat removal pump discharge pressure gauges, and proposed to use the permanently installed system gauges.
Licensee's Basis for Requesting Relief--Reference values for discharge pressures for these pumps are between 200 psig and 300 psig. This would require a discharge pressure gauge of 0-600 psig maximum. The accuracy required for this gauge would be 2% of 600 psig, which is 12 psig.
The permanent discharge pressure gauges we have installed are 0-700 psig 110 psig. Although the permanent instruments are above the maximum range limits they are within the accuracy requirements and are therefore suitable for the test. We propose to use the permanently installed discharge pressure gauges.
Evaluation--The installed residual heat removal pump discharge pressure gauges do not meet the range requirements of Section XI, Paragraph IWP-4120, however, their accuracy is better than is required by Paragraph IWP-4110. The installed instruments with ranges of 0 to 700 psig would read 28.6% of full-scale for the most restrictive reference pressure of 200 psig instead of the required 33.3% of full-scale. Therefore, meeting the Code 1
- b
\
I l
requirement is impractical with the installed instruments. However, since the installed instruments are more accurate than the Code requirements, they should provide adequate information to allow the determination of pump hydraulic condition. Because'the installed discharge pressure instruments -
will not diminish the licensee's ability to monitor pump condition and detect- !
hydraulic degradation, requiring plant modifications to meet this Code .
requirement would impose an undue burden on the licensee.
l Based on the impracticality of meeting the Code requirements, and the fact that the instruments.that the licensee has proposed to use are more accurate than the Code requires, relief may be granted from the Code requirements as requested.
Relief Request AE-1 The licensee has revised Relief Request AE-1 to include an optional alternate testing method to verify the reverse flow closure of valves AE-V120, V121, V122, and V123. The following is an evaluation of this revised relief request. This' evaluation supersedes Section 4.2.1 of the Technical Evaluation Report, Pump and Valve Inservice Testing Program, i Callaway Nuclear Plant, Unit 1.
i Relief Request. The licensee has requested relief from the exercising requirements of Section XI, Paragraph IWV-3520, for AE-V120, V121, V122, and V123, the main feedwater header check valves, and proposed to test these valves by either using a sample disassembly and inspection of these valves on
) a refueling outage frequency or by performing a seat leakage test of all these valves each refueling outage.
Licensee's Basis for Requesting Relief--Exercising these valves during power operation would require isolation of feedwater to the steam l
generator which would result in a severe transient in the steam generator, possibly causing a unit trip. Valve disassembly requires draining the steam generator below the feed header which can only be performed during longer term shutdowns.
i
l One of the following test methods shall be adhered to:
i
- a. A different valve from this group will be disassembled, inspected, and manually full-stroked at each refueling, until the entire group has been tested. If the full-stroke capability of the disassembled valve is in question, the remainder of the valves in that group will also be disassembled,' inspected, and manually full-stroked at the same outage.
- b. Each valve of this group will be seat leak tested at refueling to prove valve closure capabilities.
Evaluation--Isolation of feedwater flow during power operations to exercise these valves closed would result in loss of steam generator level control which could cause a plant trip. Since this testing would cause a plant transient which could lead to a plant trip, it is not considered to be practical during power operations. The only methods available to verify valve closure during cold shutdowns and refueling outages are leak testing and disassembly and inspection of the valves. The system design requires the steam generators to be isolated and pressurized to leak test these valves which is impractical during cold shutdowns. To disassemble these valves, the licensee must first drain the steam generators below the feed nozzle and subsequently drain the main feedwater header which is an evolution that is impractical during cold shutdowns since it could delay plant start-up.
Either a seat leakage test or disassembly, inspection, and manually exercising the valve disk during reactor refueling outages would provide an indication of valve mechanical condition and its ability to close to perform its safety related function.
Compliance with the Code required testing frequency would be burdensome since this would require quarterly shutdown and valve disassembly or leak testing. Based on the burden to the licensee of complying with the Code I
required testing frequency, and the licensee's proposed alternate testing of verifying valve closure by disassembly, inspection, and manually exercising the valve disks or performing a seat leakage test during reactor refueling outages, relief may be granted from the Code requirements as requested. 1 3 -- ------------------U
j
. \
Both of these alternate testing methods will provide confirmation that the valve disk is on its seat as required by IWV-3522(a), and as such are acceptable. The licensee requested this option in order to allow them to use the leak test method, if it proves to be feasible for this system configuration, with the sample disassembly as the back-up method if the leak test cannot be performed.
\
4
1 1
VALVES TESTED DURING COLD SHUTDOWNS The following is a Category B valve that meets the exercising i requirements of the ASME Code,Section XI, and is not full-stroke exercised every three months during plant operation. This valve is specifically j identified by the owner in accordance with Paragraphs IWV-3412 and -3522 and is full-stroke exercised during cold shutdowns and refueling outages. This valve has been reviewed and the reviewer agrees with the licensee that testing this valve during power operations is not practical due to the valve type, location, or system design. This valve should not be full-stroke exercised during power operations.
HIGH PRESSURE COOLANT INJECTION SYSTEM Category B Valve ;
EM-HV-8835, the isolation valve in the discharge line from the safety injection pumps to the reactor coolant system cold legs, cannot be exercised during power operation because this valve is required to remain open during power operations by the plant Technical Specifications. Failure of this valve in the closed position during testing could render an entire safety-system inoperable. This valve will be full-stroke exercised during cold shutdowns and refueling outages.
i 5
r OTHER PROGRAM REVISIONS NOT ADDRESSED IN THE TER Relief request AL-2 was withdrawn because the licensee determined that AL-V006, V009, V012, and V015, the check valves in the essential service water supply lines to the suctions of the auxiliary feedwater pumps, can be manually exercised while monitoring the opening torque. The ASME Code,Section XI,. subsection IWV-3522 permits exercising check valves by using a mechanical exerciser to move the valve disk, the operating force.or torque applied to the exerciser must be measured and must meet the requirements specified in IWV-3522. Since the licensee will be complying with the Code requirements for these valves, relief is not necessary and relief request AL-2 may be deleted. Section 4.3 of the Technical Evaluation Report, Pump and Valve Inservice Testing Program, Callaway Nuclear Plant, Unit 1, is no longer aoplicable.
Relief Request EJ-5 is added to the valve listing table relief request :
column for valve EJ-HV-8840. The omission of the relief request designator for this valve was a typographical error. EJ-HV-8840 is included in relief request EJ-5 and was evaluated in Appendix A, Section 8.3, of the Technical Evaluation Report, Pump and Valve Inservice Testing Program, Callaway Nuclear Plant, Unit 1. This change corrects a typographical error that was recognized during the preparation of the TER and it has no programmatic effect, therefore, no change is necessary in the TER. 1 Valves EM-HV-8870A, -88708, and -8883 are deleted from the IST program because they were inadvertently added to the program and the licensee stated that they are not required to be tested. Valve EM-HV-8883 is the isolation valve between the boron injection recirculation pumps and the boron injection tank (BIT), and valves EM-HV-8870A and -8870B are isolation valves in the recirculation flow path between the BIT and the boron injection surge tank.
The BIT at Callaway has been analyzed to contain water with as low as O ppm j boric acid concentration. Since the BIT does not have high concentrations of l l
boric acid, it is not necessary to recirculate the water in the tank to prevent the boric acid from coming out of solution, therefore, these valves are not required to change position to perform a safety function. There are l
L______---_.---_-- 6 ,
P no testing requirements for these category B-passive valves and deleting them from the IST program would have no programmatic effect, therefore, no change is necessary in the TER.
s
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ._ _ 7