ML19354D552

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LER 89-038-00:on 891114,reactor Trip Occurred as Result of Inadvertent Closure of Main Feedwater Isolation Valve.Caused by Personnel Error.Disciplinary Action Taken Against Individual & Senior Mgt Personnel Put on shift.W/891214 Ltr
ML19354D552
Person / Time
Site: Arkansas Nuclear Entergy icon.png
Issue date: 12/14/1989
From: Ewing E, Taylor L
ARKANSAS POWER & LIGHT CO.
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
1CAN128908, LER-89-038, LER-89-38, NUDOCS 8912270264
Download: ML19354D552 (4)
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o na av w awn hp Te i y* ,4 f4 t p December 14, 1989 1CAN128908 U. S. Nuclear Regulatory Commission Document Control Desk Mail Station P1-137 Washington, D. C. 20555 SilBJECT: Arkansas Nuclear One - Unit 1 Docket No. 50-313 License No. DPR-51 Licensee Event Report No. 50-313/89-038-00 /

Gentlemen:

In accordance with 10CFR50.73(a)(2)(iv), attached is the subject report concerning a reactor trip on high Reactor Coolant System pressure which resulted from the inadvertent closure of a main feedwater isolation valve duc to personnel error.

[ Very truly yours,

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E. C. wing General Manager, Technical Support and Assessment ECE/RHS/sgw attachment cc: Regional Administrator Region IV

_ U. S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 1000 g Arlington, TX 76011 3!

INP0 Records Center i

1500 Circle 75 Parkway 0 Atlanta, GA 30339-3064

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8912270264 891214 PDR ADOCK 05000313

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  • ' ' Form 1062.01 A h NRC Form 366 U.S. Nuclear Regulatory Commission

-(9-83) Approved DMB No. 3150-0104 Expires: 8/31/85 LICEN5EE EVENT REPORT (L E R)

FACILITY NAME (1) Arkansas Nuclear One, Unit One IDOCKET NUMBER (2) (PAGE (3) 10151010101 31 11 311IDFl013 TITLE (4) Reactor Trip on High Reactor Coolant System Pressure Which Resulted f rom the inadvertent Closure of a Main Feedwater Isolation Valve Due to Personnel Error EVENT DATE (5) LER NUMBER (6) i REPORT DATE (7) l OTHER FACILITIES INVOLVE 0 (8) l l 1 ISequentiell IRevisiont l l l l Monthi Day lYear (Year I I Number 1 l Number IMonth! Day l Year i Facility Names 10ocket Number (s) l l l l l l i i i l l 10151010101 1 11 11 11 41 8I 91 81 91--! Of 31 81--I 01 01 11 fl 11 41 81 91 1015l010101 1 OPERATING l ITHIS REPORT IS SUBMIT 7ED PURSUANT TO THE REQUIREMENTS OF 10 CFR 5:

MODE (9) i NI (Check one or more cf the followino) (11)

POWERl l__t 20.402(b) l__l 20.405(c) l_xt 50.73(a)(2)(iv) l__t 73.71(b)

LEVELI l__l 20.405(a)(1)(1) l__l 50.36(c)(1) l__l 50.73(a)(2)(v) l__l 73.71(c)

'(10) 1017141 l 20.405(a)(1)(ii) l__l 50.36(c)(2) l__l 50.73(a)(2)(vii) 1 _l Other (Specify in l__l 20.405(a)(1)(tii) l__l 50.73(a)(2)(1) l__l 50.73(a)(2)(viii)(A)l Abstract below and I _I 20.405(a)(1)(iv) l__l 50.73(a)(2)(ii) i__l 50.73(a)(2)(viii)(B)l in Text, NRC Form I I 20.405(a)(1)(v) i 1 50.73(a)(2)(iii) 1 1 50.73(a)(2)(x) l 366A) tlCENSEE CONTACT FOP THIS LER (12)

Name l Telephone Number l Area l Larry A. Taylor, Nuclear Safety and Licensing Specialist Itode l 1510111916l41-13111010 COMPLETE ONE LINE FOR EACH COMPONENT F AILURE DESCRIBE 0 IN THIS REPORT (L3) l 1 i IReportablel l I I I IReportabiel Causelsysteel Component IManufacturert to NPROS l ICauselSysteel Component IManufacturert to NPROS I l 1 I l 1 1 1 1 1 i l l I I I I I I I I I I I I I i i I i I i ! i 1 I I i l i l i i l I i i i 11 I I i I 1 i i i I I I I I I I I I I I I I I I SUPPLEMENT REPORT EXPECTED (14) 1 EXPECTED 1 Monthi Day - Year

~ i SUBMISSION l l l l Yes (if yes, complete Expected Submission Date) til No 1 DATE (15) l l t l l 1 ABSTRACT (Limit to 1400 spaces, i.e. , approximately fif teen single-space typewritten lines) (16)

On November 14, 1989 at approximately 2323, a reacter trip occurred as a result of the inadvertent closure of a main feedwater isolation valve. While attempting to close the suction isolation valve from the condensate storage tank to the steam driven emergency feed pump during MOVATS testing, a licensed plant operator inadvertently closed the main feedwater isolation valve for the "B" steam

. generator. This action caused the "B" main feedwater pump to trip on high discharge pressure and resulted in a reactor trip due to a high Reactor coolant System (RCS) pressure of 2355 psig. The initial plant response following the trip was normal. However, due to various steam leakage paths in the secondary system, steam Denerator pressures gradually decayed to 935 psig and 891 psig for the "A" and "B" generators respectively. RCS temperature decreased to 540.2 degrees. These values are slightly below those normally anticipated during post trip conditions. The root cause of this event was personnel error in that the operator f ailed to ensure that be was menipulating the correct valve. Disciplinary action was taken against the operator responsible for manipulating the wrong valve. In addition, a

- Secondary Systems Evaluation Team was formed to identify and recommend corrective actions for secondary system deficiencies. Significant deficiencies were corrected during mid cycle outage IM89, which is presently in progress.

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Form 1062.01B NRC Form 366A U.S. Nuclear Regulatory Commission 4 (9 83)= Approved OMB No. 3150-0104 -

Expirest 8/31/85 LICENSEE EVENT REPORT (LER) TEX 1 CONTINUATION FACILITY NAME (1) IDOCKET NUMBER (2) l LER NUMBER (6) l PAGE (3) l l l lSequentiali IRevision!

Arkansas Nuclear Ono, Unit One l l_Yearl 1 Number l l Humber l 10151010101 31 11 31 81 91- I 01 31 81--l 01 Ol01210F1013 TEXT (If more space is required, use additional NRC Form 366A's) (17)

A. Plant Status-At the time of this event Arkansas Nuclear One, Unit 1 (ANO-1) was operating at approximately 74 percent of rated power. Reactor Coolant System (RCS) [AB) pressure was 2150 psig and RCS average I- temperature was 579 degrees. The "D" reactor coolant pump was out of service due to an oil l

1eak.

B. Event Description On November 14, 1989, at approximately 2323, a reactor trip occurred as a result of the inadvertent closure of the main feeWater isolation valve (CV-2630) for the "B" Once Through Steam Generator (OTSG).

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l While attempting to close the suction isolation valve from the condensate storage tank to the steam driven emergency feedwater pump (CV-2802) during MOVATS testing, a licensed plant operator inadvertently closed CV-2630, the main feedwater isolation valve for the "B" OTSG. Tnis action resulted in main feedwater pump (MFWP) P-1B tripping on high discharge pressure. The reduced feedwater flow caused a decrease in heat transfer capability, which resulted in an increase in RCS temperature and pressure. Although the operators attempted to control RCS pressure by manually iritiating full pressurizer spray and opening the pressurizer electromagnetic relief valve (ERV),

a reactor trip was initiated by the Reactor Protection System (RPS)(JC) at an RCS pressure of 2355 psig. An operator had also been immediately dispatched to try to stop CV-2630 from fully closing by opening its circuit breaker. Although there were no security barriers which significantly impeded the operator's action, there was not sufficient time for him to prevent the valve from closing. The initial plant response following the trip was as expected with all post trip parameters being normal. However, due to various steam leakage paths in the secondary system, the OTSG pressures gradually decayed (approximately 37 minutes) to 935 psig and 891 psig for the "A" and "B" OTSGs respectively. The decrease in OTSG pressures in turn caused RCS average temperature to decrease to 540.2 degrees. These values are slightly below those normally anticipated during post trip conditions (pressure - 2005 psig, temperature - 545 degrees).

The major contributor to the OTSG pressure degradation and associated RCS cooldown was steam leakage through the moisture separator reheater (MSR) isolation valves, then through manually positioned MSR distiller level control valves to high pressure heaters E-1A and B. The relief valves on the shell side of the high pressure heaters lifted, relieving steam to the atmosphere.

Other sources of steam leakage included the feedwater pump turbine stop and governor valves, and the main turbine bypass valves.

The operators recognized the abnormal plant response and took timely corrective actions to isolate the various steam leakage paths and stabilize the plant in the hot shutdown condition.

At 0042 on Noves.ber 15, 1989, reactor startup was commenced. The main turbine was tied to the grid on November 15 at 1254.

C. Safety Significance During this event, a reactor trip was initiated at an RCS pressure of 2355 psig, as required, and all control rods inserted. All plant parameters remained within normal bands with the exception of OTSG pressures and RCS average temperature, which were slightly lower than normally anticipated during post trip conditions. The operators took timely and appropriate corrective actions to stabilize the plant in the hot shutdown condition. Although the malfunctioning of secondary system components complicated the operator's post trip responses, they did not create any significant safety problems. Therefore, the safety significance of this event is considered minimal.

O. Root Cause The root cause of this event was determined to be personnel error. Although the valve operating hand switches for CV-2802 and CV-2630 are in close proximity to each other, the operator f ailed to ensure that he was operating the correct valve.

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Form 1062.01B NRC Forn 366A U.S. Nuclear Regulatory Commission (9-83) Approved OMB No. 3150 0104 Expires: 8/31/85 LICENSEE EVENT REPORT (LER) TEXT CONTINUATION q FACILITY NAME (1) IDOCKET NUMBER (2) l LER NUMBER (6) l PAGE (3) l l l 15equentiall IRevision)

Arkansas Nuclear Onei Unit One l l Yearl l Number l l Number l 10151010101 31 11 31 81 91--I 01 31 81--I 01 Ol01310Fl0f3 TEXT (if more space is required, use additional NRC f ore 366A's) (17) l E. Basis for Reportability '

This event is reportable pursuant to 10CFR50.73(a)(2)(iv) as an automatic actuation of the RPS.

L This event was also reported in accordance with 10CFR50.72 on November 15, 1989 at 0026.

F. Corrective Actions l

Disciplinary action was taken against the operator responsible for manipulating the wrong valve. l Additionally, senior management personnel were placed on shift from November 15 intil Monday, November 20 to observe plant operation and to collect feedba:k from Operations psrsonnel with respect to problems associated with the plant and/or work environment. During this time, a reduction in maintenance activity was in effect in which only Technical Specifications or safety related maintenance was performed.

As a result of the post trip complications caused by secondary system equipment prcblems during this and a previous trip, which occurred on November 10, 1989 (LER 50-313/89-037-00), a secondary Systems Evaluation Team was formed at the direction of the Vice President, Nuclear CPS, to address problems associated with the operation of various ANO-1 secondary systems. This team consisted of senior personnel from various disciplines as well as two representatives of other utilities. Its objectives were to identify the long standing material problems on the ANO-1 secondary plant that created the need for operator compensatory actions during transients. The problems were addressed l- individually and on an integrated system operation basis considering plant impact in terms of l performance, post trip activity, personnel safety and other considerations.

As a result of these evaluations - six of the deficiencies were identified as significant enough to require correction prior to restart. These six deficiencies were:

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  • Feedwater pump turbine high pressure stop and governor valves leak.

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  • Moisture Separator reheater distiller level controllers are inoperable.
  • Heater drain tank T40 high level dump valves leak excessively.
  • Main feedwater pump recirculation valves leak excessively.

Appropriate interim or permanent corrective actions were implemented with respect to each of these deficiencies and for many of the less significant items. The remaining items will be addressed during future outages.

G. Additional Information Previous siellar events in which reactor trips were caused by personnel error were reported in LERs 50-313/87-005-00 and 50-313/88-018-00.

Energy Industry Identification System (EIIS) codes are indicated in the text as [XX).

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