ML20046D594

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LER 93-006-00:on 930721,determined That Valves Could Not Be Closed by Use of Normal Motive Power Due to Inadequate Procedural Controls Resulting in Valve Actuators Being Set Up Improperly.Isolated Affected penetration.W/930818 Ltr
ML20046D594
Person / Time
Site: Hatch Southern Nuclear icon.png
Issue date: 08/18/1993
From: Beckham J, Tipps S
GEORGIA POWER CO.
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
HL-3433, LER-93-006-02, LER-93-6-2, NUDOCS 9308250035
Download: ML20046D594 (7)
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Text

,

e Georgia Power company -

40 invorrw.zs Center Parkway

. Post O' face Bcx 1295 Birmirgham, Alanama 3s201 Telephone 205 877-7279 J. T. Dockham, Jr. Georgia Power Vice Proudent - Nucuar '

Hatch Project I'v s .st' nen,m ,  ; n August 18, 1993 Docket No. 50-366 HL-3433 006072 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555 Edwin I. Hatch Nuclear Plant - Unit 2 -

Licensee Event Report less Than Adequate Procedure Results in Inoperable Reactor Coolant Pressure Boundary Isolation Valves Gentlemen: ,

In accordance with the provisions of 10 CFR 50.73(a)(2)(v), Georgia Power Company is submitting the enclosed Licensee Event Report (LER) concerning a less than adequate procedure which resulted in inoperable reactor coolant pressure boundary isolation valves. This event occurred at Plant Hatch -

Unit 2.

Sincerely, J. T. Beckham, Jr.

JKB/cr

Enclosure:

LER 50-366/1993-006 cc: Georaia Power Company Mr. H. L. Sumner, General Manager - Nuclear Plant '

NORMS U.S. Nuclear Reaulatory Commission. Washinaton. D.C.

Mr. K. Jabbour, Licensing Project Manager - Hatch U.S. Nuclear Reaulatory Commission. Reaion II ,

Mr. S. D. Ebneter, Regional Administrator l Mr. L. D. Wert, Senior Resident Inspector - Hatch DO k khhd$66 '

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LICENSEE EVENT REPORT (LER)

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PIANT E.1. IIATCil, UNIT 2 05000366 i lgp l 6 liiti (4) 1ESS TilAN ADEQUATE PROCEDURE RESULTS IN INOPERABLE REACTOR COOLANT PRESSURE BOUNDARY iso 1ATION VAlNES I EVEN1 DATE (5) LER huMbEk (6) REFORT DATE (7) OinEk FACILITIE5 INv0LviD (6)

MONIr LAY YEAR YEAR SEQ NUM REW *LNIn DA1 tiAR FACILITY NAME5 DDCAEi NUMBEk(5) l 05000 -i i

07 21 93 93 006 00 08 18 '93 05000  :

GPERATING I '

MODE (9) 1 pg,402(b) 20 4Ci(c) 50.73(a)(2)(tv) 73.71(b) 20,405(a)(1)(t) 50.3E(c)(1) ^ 50.73(a)(2)(v) 73.71(c)

POWER M 83 (

20.405(a)(1)(ii) 50.36(c)(2) 50.73(a)(2)(vii) OTHER (Specify in 20.405(6)(1)(iii) 50.73(a)(2)(i) 53.73(a)(2)(viii)(A) Abstract below) i 20.405(a)(1)(iv) 50.73(a)(2)(ii) 50.73(a)(2)(viii)(B) 20.405(a)(1)fv) 50.73(a)(2)(iii) 50.73fa)(2)(x) 1.ICENSEt CCNTACI F04 inib LER (12)

NAME TELEPh0NE NUMBER .

JE A CDOE STEVEN B. TIPPS, MANAGER NUCIEAR SAFETY AND COFiPLIANCE. HATCH 912 367-7851 CLMPLETE ONE LINi FGR (ACH FAILukE DESCRibfD lh IHi5 REP 0ki (13) ,

AU5E ;Y5 TEM COMPONENT MANUIAC- ,Rl P CE I CAUSE >YSTEM COMPONENT M.ANUFAC- REPORT TURER iO NPROS iURfr TO NFRDS SUPi L E ME NT AL REPlai EMECIED (14) MONin DAY TEAR t EXPECTED t SU6 MISSION l

] YES(If ye s, complete EXPICIED SLBMI5510N DATE) DATE (15) i

% NO At< 5Is Ac i (it>)

t On 7/21/93, at 1800 CDT, Unit 2 was in the Run mode at a power level of 2046 CMWT (85 percent rated thermal power). At that time, licensed personnel were notified that valves 2B31-F014 and F020 could not be fully closed against '

reactor pressure by use of normal motive power. These valves are air operated, i normally open, spring to close valves that provide isolation capability for a three-quarter inch diameter reactor coolant sample line. The valves function as Primary Containment Isolation System (PCIS) valves and Reactor Coolant Pressure i Boundary (RCPB) isolation valves. As RCPB isolation valves, pursuant to ,

10 CFR 50, Appendix A. General Design Criterion 55, they should be capable of closing against normal operating pressure. Based on previous local leak rate testing results, the valves would isolate the Primary Containment in the event ,

of a design basis accident. By 2035 CDT, outboard valve 2B31-F020 had been  !

de-activated and mechanically secured in the closed position.

The cause of the condition was less than adequate procedural controls resulting l in the valve actuators being set up improperly. Corrective actions include -

isolating the affected penetration, developing a procedure, reviewing similar valve applications, and properly setting up the valves during the next outage of  !

sufficient duration.

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  • LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FA'CILITY NAME (1) DCCKET NUMBER (2) LER NUMBER (5) PAGE (3) tiAk 5EQ hum REW PIR7f E. I. HATCH, UNIT 2 05000366 93 006 00 2 0F 6 IE x1 PIANT AND SYSTEM IDENTIFICATION General Electric - Boiling Water Reactor Energy Industry Identification System Codes are identified in the text as (EIIS Code XX).

DESCRIPTION OF EVENT on 7/21/93, at 1800 CDT, Unit 2 was in the Ran mode at a power level of 2046 '

CMWT (85 percent rated thermal power). At that time, licensed personnel were notified that valves 2B31-F019 and F020 could not be fully closed against reactor pressure by use of normal motive power. These valves provide isolation capability for the three-quarter inch diameter reactor coolant sample line at the point where the line penetrates Primary Containment. As such, the valves ,

function as Primary Containment Isolation System (PCIS, EIIS Code JM) valves and Reactor Coolant Pressure Boundary (RCPB) isolation valves. These valves are equipped with air operators and are designed such that the valves open with air pressure and close with spring force.

A review of the local leak rate testing results performed during the previous refueling outage showed that the valves would perform their required Primary Containment isolation function in the event of an accident. However, it was determined that the valves would not fully close against a normal reactor pressure of approximately 1005 psig. As RCPB isolation valves, pursuant to 10 CFR 50, Appendix A, General Design Criterion 55, the valves should be capable of closing against normal operating pressure.

The condition was identified while preparing to backflush the sample line as ,

part of a semiannual periodic maintenance activity. The backflush requires isolating and draining the sample line downstream of outboard PCIS valve 2B31-F020. When it was apparent that the line could not be drained, an I investigation into the problem showed that even though valves 2B31-F019 and F020 indicated closed they were not fully seated. When valve 2B31-F020 was later manually secured, the valve was found to move approximately one-eighth of an inch before it fully seated against reactor pressure.

Subsequently, the licensed Shift Supervisor was notified of the condition. The  ;

Technical Specifications do not contain any explicit requirements for the RCPB '

isolation function of these valves. Consequently, as a conservative action, the Technical Specification requirement regarding inoperable PCIS valves was implemented, even though local leak rate and stroke time test data indicated  ;

that the valver. would function satisfactorily as PCIS valves. This 3 specification required that at least one valve in the affected penetration be l de-activated and secured in the closed position within four hours or be in Hot  ;

Shutdown in the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and Cold Shutdown in the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. '

Therefore, at 203S CDT, valve 2B31-F020 was de-activated and secured in the closed position by use of a gagging device, achieving full compliance with the

, aforementioned Technical Specifications requirement. Inboard valve 2B31-F019 was not gagged because it is located inside Primary Containment and thus was not accessible.

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. LXCENSEE EVENT REPORT (LER)

TEXT CONTINUATION rAtill1Y 4AME (1) DOCKET NUMBER (2) LER NUMBER (5) PAGE (3) l TEAR SEQ hum KEV PiAh7 E. I. HATCH, UNIT 2 05000366 93 006 00 3 CF 6 I TEu CAUSE OF EVENT The cause of this event was less than adequate procedural controls resulting in insufficient preload being applied to the actuator spring. As previously discussed, these valves are designed such that the spring functions to close the valve. The ability of the valve to fully close against design differential pressure is in part dependent upon the preload applied to the actuator spring during setup of the actuator. Therefore, it is important that spring preload either be set to a predetermined value or be set at its maximum value. No procedure exists for setting up the actuator involved in this event. Also, the intent of the maintenance activity was to repair the valve so that it would pass the local leak rate test (a test performed to prove Primary Containment >

1 solation capability). Consequently, in October of 1992 following maintenance on the valves, the stroke and, thus, the preload of the actuator spring, apparently was set only to ensure that the valves would seat against the 57.5 psid applicable to local leak rate testing. As such, without the valve actuator adjusted to obtain the design preload on the actuator spring, the actuator could not provide sufficient force to close the valves against normal reactor pressure.

i REPORTABILITY ANALYSIS AND SAFETY ASSESSMENT This report is required pursuant to 10 CFR 50.73(a)(2)(v) because an event occurred which alone could have prevented the fulfillment of the safety function of a system designed to control the release of radioactive material.

Specifically, less than adequate procedural controls resulted in a maintenance activity rendering two redundant Reactor Coolant Pressure Boundary isolation valves incapable of fully closing against reactor pressure. The capability of the valves to close or be closed under normal operating conditions in the event of a line rupture is required by General Design Criterion 55 of 10 CFR 50, Appendix A.

The affected reactor water sample line is a three-quarter inch line providing a continuous supply of water to the backup in-line reactor coolant conductivity monitor and to the Crack Arrest Verification system. Valves 2B31-F019 and F020 are independent and redundant isolation valves located on the line inboard and outboard of the Primary Containment. As stated previously, the valves perform two functions: Primary Containment isolation and RCPB isolation. .

. PCIS provides automatic isolation capal ity of Primary Containment pe: cations  ;

to preclude the release of radioactis 'aterial in the event of an acci The ability of the valves to isolate ti,e associated Primary Containment penetration in the event of an accident was not affected by this event. Prior to startup up from the last refueling outage, local leak rate testing was satisfactorily performed on these valves de- snstrating their ability to provide 3

isolation of the sample line penetration against the peak Primary Containment

! pressures postulated during an accident. This testing was performed following the maintenance activity that resulted in the improper setup of the valve.

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TEXT CONTINUATION TACILITY NAME [1) DOCKET NUMEER (2) LER NUMBER (5) PAGE (3) !f; sto NuM REv PIMT E. I. HATCH, WIT 2 05000366 93 006 00 4 CF 6 IEAT t Reactor Coolant Pressure Boundary isolation valves, as required by General - Design Criterion 55 of 10 CFR 50, Appendix A, are designed such thi.t the RCPB can be isolated in the event of a break in a line that forms part of the boundary in order to limit the loss of reactor coolant inventory and to control the release of radioactive material. Implicit in this requirement is that the line should be capable of being isolated under normal operating conditions, which in this case would be a differential pressure across the valve seat of approximately 1005 psid. The reactor coolant sample line is a three-quarter inch diameter Class 1 pipe that taps off of a Recirculation System pump discharge line. In this event, the valves were found to be incapable of fully isolating the sample line against normal reactor pressure. Had a rupture of the sample line downstream of the outboard isolation valve occurred, the valves would have been incapable of fully closing resulting in an non-isolable leak of reactor coolant into Secondary Containment. This condition would be identified by either personal observation, an Area Radiation Monitor alarming, or the Reactor Building Ventilation radiation monitoring system alarming and isolating i Secondary Containment. Reactor coolant inventory loss through this three-quarter inch line would be made up by the Feedwater system and therefore would be undetectable from a reactor water level standpoint. After attempts to isolate the leak using the valve remote control switches proved unsuccessful, the reactor would be shut down and depressurized. At some point during depressurization, the differential pressure across the valves would decrease to the point that they would close and isolate the penetration. The postulated small line rupture analyzed in the Final Safety Analysis Report (FSAR) is that of an instrument line. The analysis assumes tha* the instrument line ruptures outside of Primary Containment and upstream of the excess flow isolation valve. This failure results in an non-isolable leak and a release of reactor coolant to secondary Containment until the reactor is depressurized. A one-quarter inch diameter orifice is located in the instrument line which limits flow out of the ruptured line. In this scenario, the FSAR assumes that within 10 minutes of the failure occurring, it would be identified. Within 12 minutes, action would be initiated to mitigate the condition. If the line could not be isolated, the reactor would be shut down and depressurized within the following four hours per existing emergency procedures. The results of the analysis show that the offsite doses expected in such an event would be significantly less than the 10 CFR 100 limits. The amount of coolant discharged from the failed line would have no appreciable effect on the reactor coolant inventory and, thus, no effect on reactor core cooling. Therefore, no fuel failures would occur. The fission products released to Secondary Containment are based on those the analysis assumes to exist prior to the event due to leaking fuel assemblies. Some of this radioactive material is expected to plate out in Secondary Containment. The remaining radioactive material will be processed by the Standby Cas Treatment System (SGTS, EIIS Code BH) and discharged at an elevated release point. These releases would result in offsite doses being a fraction of the 10 CFR 100 limits. 

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         .         LICENSEE EVENT REPORT (IER)                                                                  ,

TEXT CONTINUATION FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (5) PAGE (3) YEAR SEQ hum REW PiANT E. I. HATQi,1 NIT 2 05000366 93 006 00 5 or 6 IEAT The consequences of a failure of the sample line, even though not bounded by the instrument line rupture analysis results, would be very similar. The sample line is not equipped with a flow restricting orifice. However, the flow characteristics associated with the two partially closed isolation valves would approxivate that of the flow limiting orifice in the instrument line. As stated i previously, the globe valve disk of valve 2B31-F020 was found to be approximately one-eighth of an inch from its close seat when subjected to i reactor pressure. Calculations show that this configuration would result in less discharge out of the sample line than that out of an instrument line equipped with a one-quarter inch diameter flow limiting orifice. It is understood that a rupture of the line at the downstream side of the valve vould result in a higher differential pressure across the valve and most likely the globe valve disk being positioned higher in the valve bc,dy cage. However, based

  • on engineering judgement, it was concluded that the increased flow in this situation would be minimal but could exceed that analyzed in an instrument line.

An increase in flow would result in a proportional increase in fission products released to Secondary Containment. The resulting offsite doses therefore could exceed that for the instrument line break by a small amount. However, it would still amount to offsite doses being a fraction of the 10 CFR 100 limits.

  • Based on the above information, it is concluded that this event had no adverse impact on nuclear safety. This assessment applies to all operating conditions.

CORRECTIVE ACTIONS Valve 2B31-F020 was mechanically gagged closed, isolating the Primary r Containment penetration. A procedure will be developed for setting up the actuators of the model and make involved in this event. The procedure will be issued by 11/5/93. A review of other Reactor Coolant Pressure Boundary isolation valves showed that a similar condition may exist with the counterpart Unit i valves 1B31-F019/F020. As a result, both of these valves have been closed and deactivated, and the outboard valve (IB31-F023) has been manually gagged in the closed position as a conservative measure. No leakage was detected when the valves were first closed using the control switch. Further actions will be taken to ensure operability of the valves prior to opening the.n. During the next outage of sufficient duration, the valve stroke and spring preload for each of the valves will be set properly and the valves will then be returned to service. ADDITIONAL INFORMATION No systems other than those previously identified in this report were affected by this event. J 

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  • LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION I rACILITY NAME (1) DOCKET NUMEER (2) LER NUMBER (5) PAGE (3) ' YEAR SEQ NuM REv PIRiT E. I HATCH, UNIT 2 05000366 93 006 00 6 0F 6 IEAT Two previous similar events occurred in the past two years in which a maintenance activity or an inadequate procedure resulted in a loss of safety function of a system. These events were' addressed in LERs 50-321/92-003, dated f 3/31/92, and 50-366/92-006, dated 3/25/92. In each of these events an inadequate procedure resulted in the High Pressure Coolant Injection System (HPCI, EIIS Code BJ), a single train safety system, being rendered inoperable. In both cases, the procedures were revised. These corrective actions could not have prevented this event because neither the involved procedures nor their revision had any bearing on the set up of the 2B31-F019 and F020 valve actuators. No failed components contributed to or resulted from this event. t ( 2 3 3 ____ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ . _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ . _ _ _ . _ _ _ _ _ _ _}}

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