ML19327B351

LER 89-018-00:on 890921,while Testing Main Steam Safety Valve,Lift Pressure Found to Be Below Tech Spec Allowable Value.Caused by Error in Judgement by Plant Supervisors. Event Will Be Discussed by superintendent.W/891023 Ltr

ML19327B351
Person / Time
Site:	Waterford
Issue date:	10/23/1989
From:	Mcgaha J, Packer D LOUISIANA POWER & LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-89-018, LER-89-18, W3A89-0197, W3A89-197, NUDOCS 8910300230
Download: ML19327B351 (7)

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Ref: 10CFR50.73(a)(2)(1)

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October. 23, 1989 U.S., Nuclear Regulatory Commission ATTENTION: Docu'nent Control Desk Washington, D.C, 20555

Subject:

Waterford 3 SES Docket No. 50-382

' License No. NPF-38 Reporting of Licensee Event Report Centlement Attached is Licensee Event Report Number LER-89-018-00 for Waterford Steam Electric Station Unit 3. This Licensee Event Report is submitted pursuant to 10CFR50.73(a)(2)(i).

Very truly yours,

\ g J.R. McGaha 4 Plant Manager - Nuclear JRM/PTG/rk (w/ Attachment) cci Messrs. R.D. Martin J.T. Wheelock - INPO Records Center E.L. Blake W.M. Stevenson D.L. Wigginton NRC Resident Inspectors Office G910300230 891023 ADOCK0500gj2

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On September 21, 1989 Waterford Steam Electric Stntion Unit 3 was operating at 100% reactor power. At 1241 hours0.0144 days <br />0.345 hours <br />0.00205 weeks <br />4.722005e-4 months <br />, while testing a Main Steam Safety Valve, the lift pressure was found to be below the Technical Specification (TS) allowable value. Because of aberrations in indicated main steam pressure when repositioning a gauge that was improperly installed, the surveillance was invalidated. Test equipment (pressure gauge and chart recorder) were checked for proper calibration and operation by 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br />. The surveillance test was performed again at 2037 hours0.0236 days <br />0.566 hours <br />0.00337 weeks <br />7.750785e-4 months <br />. The valve was found to lift at a higher pressure but still below the TS limit. The valve setpoint was out of tolerance low in l excess of the four hour TS action requirement. This event is therefore reportable as a condition prohibited by TS.

Although event has identified areas that require improvements, there was a good

' faith effort by all involved personnel to meet the TS requirement by completing the test promptly while insuring accurate test results. The design basis for l

main steam safety valves (overpressure protection) was maintained throughout this event. Therefore, this event did not result in an increased risk to the health and safety of the public or plant personnel.

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0l1l8 - 0j0 0l2 or 0 j6 vom . = mic w.mmim On September 21, 1989. Waterford Steam Electric Station Unit 3 was operating ,

't at 100% power when Main Steam Safety Valve (MSSV) (Ells identifier - RV) testing was being conducted in accordance with mechanical maintenance (MM) proc'edure MN-07-015. "Trevitest of Main Steam Safoty Valves." This procedure involves a minimum of 3 lift tests per safety valve Test #1 - To verify the valve vill open and close properly; Test #2 - To measure valve operating  ;

pressure; and Test #3 - To verify test #2 results.

At approximately 1239, maintenance personnel completed Test #1 on valve MS-106A with the valve opening and reseating satisfactorily. Test #2 was then performed I

at 1241. Both of these tests indicated a lift pressure of 1030 psig which is below minimum acceptable (setpoint: 1070 psig i 1%). Several aberrations in l the performance of the in-use Trevitest Heise gauge caused the Mechanical l- Maintenance Supervisor (MMS) and the Operations Shift Supervisor (SS) to question.the test results and invalidate the surveillance test. These [

aberrations included: The Heise gauge indication (Main Steam pressure) changed  ;

10 psi vhen the gage was re-oriented from a horizontal to a vertical configuration (90 degree rotation); when depressurized and vented. the gauge indicated less than zero gauge pressure (-10 psig); the valve lif t pressure for the horizontal gauge configuration appeared inordinately low based on the ,

1 previous experience of the MMS. The Heise gauge was delivered to the site L. Metrology Lab for a calibration check. The safety valve was not declared inoperable by the SS at this time because of the questionable p u formance of the gauge.

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operability for valve MS-106A with the Trevitest unit still connected to the ,

valve. The SS decided to remove the test unit from MS-106A until he could provide the inspector with a response. At 1535, the inspector questioned the SS on his intentions for dealing with the lov lif t pressure ou test #2 for j MS-106A. The SS informed the inspector that the calibration of the gauge was being verified. The inspector responded by informing the SS that he had just i left the Main Steam Valve area and was told by the HMS that the gauge calibration checked satisfactorily. The SS and Operations Superintendent -

l Nuclear conferred on the issue and deelded to proceed with testing as soon as possible. At approximately 1540, the MMS was contacted to have the Trevitest unit (including the Heise gauge) reinstalled and proceed with testing on MS-106A. j At 1605, the SS was informed : hat testing was ready to commence however,  !

approximately 30 minutes later, the SS was notified that the chart recorder [

being used on the Trevitest unit was not operating correctly (the printing arm was skipping on the paper) and would have to be checked.

In conjunction with the continuing plant maintenance activities to resolve the

. problems with MS-106A, plant management became involved via telephone f conference calls with NRC Region IV personnel and the NRC resident inspector.

The Plant Manager and the Chief of Project Section A of Region IV conversed at j approximately 1700. During this conversation, it was discussed whether or not Waterford 3 should commence reducing power to comply with TS 3.7.1.1 (four hour action requirement). The response was not however, Waterford 3 should pursue an enforcement discretion evaluation and inform the NRC of the results within several hours. With this guidance, plant personnel developed an evaluation for presentation to Region IV.

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0 l0 0l4 0F 0 l6 I rart -. = w a =ac wamm nn In a follovup telephone conference at approximately 2000. Waterford 3 management recommended enforcement discretion until the surveillance could be performed. ,

This recommendation was based on an engineering evaluation of current valve status. If the valve did not meet the acceptance criteria when retested TS

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3.7.1.1.a action requirement vould be entered. The NRC initially discussed granting enforcement discretion until the next scheduled plant shutdown  :

(approximately 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> away). The NRC requested a chort recess so that Region IV personnel could review the engineering evaluation as well as the scope of the request.

During this period of deliberation, af ter repairing, calibrating, and installing ,

the chart recorder, testing of MS-106A recommenced at 2037. Test #2 (record [

test) was performed with a lift pressure of 1038 peig. MS-106A was declared ,

inoperable and TS 3.7.1.1.a entered.

In a final telephone conversation, Region IV was prepared to grant enforcement I discretion as requested by plant management. However, the valve had already failed the surveillance retest and TS 3.7.1.1.a action requirement had been entered. After the valvo lift setpoint was adjusted, the valve was tested i satisfactorily at 1080 psig (Final Test for Record) and 1072 psig (Final Test for Verification), and was declared operable. TS 3.7.1.1.a was exited at 2100.

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The root cause of this event was an error in judgement b

plant staff supervisors. When the questionable test results were first identified, the indeterminate status should have been discussed with higher levels of plant management. As a minimum, the safety valve should have been declared inoperable when the calibration check indicated that the gauge was in calibration; i.e.,

enough information was available to support valve inoperability as a good conservative decision. Enforcement discretion should have been pursued in parallel with these actions within the TS time constraints (four hour action requirement). The design basis for the valve (overpressure protection, which the valve was still capable of providing) and the uncertaiuty of the test equipment are believed to have significantly influenced the decision by the SS for the valve to remain operable. Contributing to this event was some procedure inadequacies, mostly in the areas of human factors, and-specific guidance on how to install the test gauge. The procedure was also nine months past its biennial review date and was written in a format that deviated from that of other Waterford 3 surveillance tests.

To prevent recurrence of this event, this report will be provided to all senior reactor operator license holders at Waterford 3 and will be specifically discussed by the Operations Superintendent - Nuclear with each SS. The above actions should be completed by November 7, 1989. This information has been promulgated in meetings held with maintenance personnel to discuss this event.

Additionally, procedure MM-07-015 is being revised to include: test equipment set-up criteria which will maintain valve operability while testing; notifying the SS on indeterminate conditions found during testing; recording as-f ound test data; and specifying complete calibration requirements for test equipment.

This revision should be completed by March 30, 1990. Additional guidance will be provided by plant management on how to handle test instrumentation malfunctions or indcterminance. A review of other surveillance procedures will be conducted to identify and resolve similar procedure problems. This review and the associated revision of any identified procedures will be completed by March 30, 1990, or before the procedure is used.

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0l0 0l6 0F 0 l6 wc w an wim j Preliminary reviews of the Final Safety Analysis. ASME Section 111 article NC7000, and the TS bases indicate that the -l% lift setpoint limit is a standard ASME tolerance value and does not correlate to valve operability as defined in the TS bases. Following further evaluation, a TS change may be initiated to request relief from the lov side limit setpoint or establish a larger allowable variance consistent with industry experience and good engineering practices.

Although this event has identified areas which require improvements, there was a good faith effort by all involved personnel to satisfy TS requirements by completing the testing promptly while inouring accurate test resultr. The design basis for main steam safety valves (overpressure protection) was maintained throughout this event. Therefore, this event did not result in an increased risk to the health and safety of the public or plant personnel.

gMILAR EVDITS None PLANT CONTACT l D.F. Packer, Assistant Plant Manager - Operations & Maintenance. 504/464-3134.

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