ML20211K916

Summary of Operating Reactor Events Meeting 87-04 on 870209. List of Attendees,Events Discussed,Significant Elements of Events & Summary of Events Encl

ML20211K916
Person / Time
Site:	Indian Point, Braidwood, Vermont Yankee, 05000000
Issue date:	02/19/1987
From:	Holahan G Office of Nuclear Reactor Regulation
To:	Harold Denton Office of Nuclear Reactor Regulation
References
OREM-87-004, OREM-87-4, NUDOCS 8702270020
Download: ML20211K916 (18)

Text

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< MEMORANDUM FOR: Harold R. Dint:n, Directov Office of Nuclear Reactor Regulation FROM: Gary M. Holahan, Director Operating Reactors Assessment Staff

SUBJECT:

SUMARY OF THE OPERATING REACTORS EVENTS MEETING ON FEBRUARY 9, 1987 - MEETING 87-04 On February 9,1987, an Operating Reactor Events meeting (87-04) was held to brief the Office Director, the Division Directors and their representatives on events which occurred since our last meeting on February 2,1987. The list of attendees is included as Enclosure 1.

The events discussed and the significant elements of these events are presented in Enclosure 2. Enclosure 3 provides a summary of those presented events that will be input to NRC's performance indicator program as significant events.

M/

Gary M. Holahan, Director Operating Reactors Assessment Staff

Enclosures:

As stated cc w/ Encl.:

See Next Page

_ DISTRIBUTION

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C. Helteres. S. Varoa R. Starostecki V. Rooney D. Ross J. Stevens T. Murley, Reg. I V. Noonan J. Nelson Grace, Reg. II J. Keppler, Reg. III R. D. Martin, Reg. IV J. B. Martin, Reg. V W. Kane Reg. I S. Ebneter, Reg. I R. Walker, Reg. II C. Norelius, Reg. III

• E. Johnson. Reg. IV D. Kirsch, Reg. V H. Thompson F. Miraglia R. Bernero T. Speis W. Russell T. Novak F. Schroeder W. Houston B. Sheron B. Boger -

D. Crutchfield E. Rossi G. Lainas V. Benaroya W. Regan D. Vassallo E. Jordan J. Rosenthal R. Baer E. Weiss R. Hernan S. Showe S. Rubin G. Arlotto

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EhCLOSURE 1 LIST OF ATTENDEES OPERATING REACTORS EVENTS BRIEFING (87-04)

FEBRUARY 9, 1987

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@$ DIVISION NAME DIVISION R. Vollmer NRR M. Johnson IE g

G. Holahan NRR M. Caruso NRR J. Carter NRR K. Naidu IE k

G. Dick NRR R. Emch NRR -

G. Cwalina NRR D. Vassallo NRR W. Regan NRR G. Murphy ORNL S. Varga NRR V. Benaroya NRR J. Stevens NRR M. Chiramal AE00 G. Klingler IE W. Haass IE A. Dromerick IE J. Henderson IE E. Weiss IE M. Slosson NRR D. Tondi NRR D. Tarnoff NRR

. ENCLOSURE 2 OPERATING REACTORS EVEllTS BRIEFING 87-04 FEBRUARY 9, 1987 INDIAN POINT 2 AUXILIARY FEEDWATER PUMPS FAILED SURVEILLANCE TEST INDIAN POINT 2 BATTERY BELOW MINIMUM TEMPERATURE FOR REQUIRED CAPACITY VERMONT YAflKEE RHR PUtlP MINIMUM FLOW PROBLEM BRAIDWOOD 1 LEAKAGE INTO CONTROL R00f1 i

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2/9/87 INDIAN POINT 2 - AUXILIARY FEEDWATER PUMPS FAILED SURVEILLANCE TEST FEBRUARY 3, 1987 - (M. SLOSSON, NRR)

PROBLEM:

TWO N0 TOR DRIVEN AUXILIARY FEEDWATER PUMPS FAILED 18 MONTH SURVEILLANCE TEST .

CAUSE:

LEAKAGE OF TWO VALVES ON CONDENSATE STORAGE TANK (CST)

RECIRCULATION LINE ATTRIBUTED TO EROSION DUE TO HIGH AP -

DISCUSSION:

PLANT IN HOT S/D TEST INDICATED LOW FLOW PUMP # 23 (420 GPM VS 450 GPM REQUIRED)

TEST INDICATED LOW HEAD PUMP # 21 (2643 FEET VS 2688 FEET REQUIRED)

LEAKAGE PAST 2 NEEDLE VALVES ON THE RECIRCULATION LINES FROM CST VALVES NORMALLY LOCKED CLOSED VALVES OPENED DURING SHUTDOWN FOR RECIRCULATION TO CONTROL TEMPERATURE OF CST VALVES REPLACED WITH GLOBE VALVES (STOP VALVE PURPOSE)

PUMP SURVEILLANCE TEST SUCCESSFUL FOLLOW-UP:

LICENSEE REVIEWING CAUSE OF EROSION  !

LICENSEE REVIEWING 400 GPM FLOW REQUIREMENT TECH SPECIFICATION REVISION APPLICATION TO REDUCE FLOW REQUIREMENT TO 300 GPM EXPECTED - NRR TO REVIEW REGION FOLLOWING OVERCURRENT PROTECTION LIMITS

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2/9/87 INDIAN POINT 2 - BATTERY BELOW MINIMUn TEMPERATURE FOR REQUIRED CAPACITY JANUARY 28, 1987, - (E. WEISS, IE)

PROBLEtj: STATION CLASS IE BATTERY #21 BELOW TEMPERATURE REQUIRED TO DELIVER REQUIRED CAPACITY CAUSE:

COLD WEATHER INADE00 ATE ROOM HEATING -

LOUVERS OPEN TO OUTSIDE BATTERY ROOM NOT ON LICENSEE'S FREEZE PROTECTION PROGRAM BATTERY LOADING ALLOWS LITTLE MARGIN FOR LOW TEMP SIGNIFICANCE:

SURVEILLANCE REQUIREMENTS AND PROCEDURES DO NOT ADDRESS THIS PROELEM BATTERY CAN BECOME IN0PERABLE WITHOUT OPERATOR'S KNOWLEDGE DISCUSSION:

PLANT AT 96% POWER REGION I INSPECTOR NOTES COLD CABLE SPREADING ROOM AND QUESTIONS OPERABILITY OF BATTERIES ELECTROLYTE TEMP 54*F ON 1 0F 4 CLASS 1E BATTERIES LICENSEE CONSULTED WITH BATTERY MANUFACTURER WHO RAN COMPUTER PROGRAM INDICATING THAT MINIMUM TEMPERATURE FOR BATTERY f21 SHOULD BE 65'F 1000 HRS - LICENSEE DETERMINES BATTERY INOPERABLE IG20 HRS - PLANT BEGINS ORDERLY REDUCTION OF POWER BATTERY CAPACITY DEGRADES AS TEMP DECREASES BATTERY LIFE DEGRADES AS TEMP INCREASES VERY LOW TEMPERATURES CAN PERMANENTLY DAMAGE BATTERY LICENSEE CLOSED LOUVERS WITH 2X4 AND ADDED PORTABLE HEATERS TO ROOM t

2/9/87 INDIAN POINT 2,(CON'T,)

LICENSEE FIRST ATTEMPTED TO RAISE TEMPERATURE WITH DISCHARGE AND THEN SUCCESSFULLY USED E0VALIZING CHARGE TO RAISE ELECTROLYTE TEMPERATURE 2145 HRS - PLANT TERMINATES SHUTDOWN AT 49% POWER WHEN ELECTROLYTE AT 65'F FOLLOW-UP; IE PREPARING INFORMATION NOTICE

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2/9/87 VERM0hT YANKEE - RHR PUMP MINIMUM FLOW PROBLEM JANUARY 29 1987, - (E. WEISS, IE)

PROBLEM: PUMP MANUFACTURER HAS SPECIFIED A MINIMUM FLOW FOR RHR PUMPS THAT WOULD REQUIRE SUBSTANTIAL MODIFICATIONS CAUSE:

PUMP MANUFACTURER HAS SPECIFIED MINIMUM FLOW VALUES BASED ON CONSERVATIVE ESTIMATE OF THE ONSET OF VIBRATION SIGNIFICANCE:

IF REVISED MINIMUM FLOW VALUE IS CORRECT, THEN -

PLANT SAFETY ANALYSIS MAY BE INCORRECT MODIFICATION OF MINI-FLOW LOGIC AND/0R MINI-FLOW LINES NECESSARY PROBLEM LIKELY TO BE GENERIC AND VALID FOR OTHER PUMPS DISCUSSION:

ON 11/13/86 BINGHAM-WILLAMETTE PUMP C0, (BWC) WROTE VT, YANKEE THAT RHR PUMP MINI FLOW SHOULD BE AT LEAST 2075 GPM FOR OPERATION LESS THAN 2 HOURS IN 24 HOUR PERIOD RHR PUMP IS LARGE SINGLE STAGE TYPE CVIC MANUFACTURED FOR GE VT, YANKEE NOW HAS MINI-FLOW ORIFICE DESIGNED FOR 350 GPM RHR PUMP SIZE IS 8800 GPM AT RUN00T (PER BWC)

OTHER LICENSEES RECEIVING SIMILAR LETTERS INCLUDE:

COOPER PILGRIM BROWNS FERRY PEACil BOTTOM VT. YANKEE CONTRACTED FOR G.E. TO PROVIDE ANSWER G.E 's POSITION (BASED ON ANALYSIS BUT NOT TEST) WILL BE THAT 350 GPM IS ADEQUATE BASED ON ESTIMATE OF 20 HOURS OVER 40 YEAR LIFE OF PLANT EMERGENCY GUIDELINES THAT CALL FOR SHUTTING OFF RHR/LPCI WHEN DEAD HEADED NOTING THAT BWC VALUE IS BASED ON 29,200 HOURS OVER 40 YEAR LIFE ABSENCE OF DISABLING DAMAGE

e 2/9/87 VERMONT YANKEE,(CON'T.)

BINGHAM-WILLAMETTE DEVELOPED REVISED MIN 1-FLOW VALUES AFTER INSPECTING PEACH BOTTOM RHR PUMPS UNRELATED PROBLEM WITH WEAR RING CAUSED INSPECTION NO PUMP BEARING DAMAGE, NO PITTING 0F IMPELLER BLUE COLOR ON IMPELLER INCREASED MINI-FLOW SHOULD PREVENT VIBRATION FOLLOW-UP: .

IE (EAB) CONTACTED BINGHAM WILLAMETTE IE (EAB), IE (VPB), AND REGION I CONTACTED G.E.

REGION I WILL REQUEST ASSISTANCE OF NRR

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• Post ottice Don 10247 + Potand. Oregon G7210 USA * (SG3) 226 5200

• To ea 360 $

November 13, 1986 Vermont Yankee Nuclear Power Corporation P.O. Box 157 Governor Hunt Road 7 Vernon, Vermont 053'54 Attention: James E. Edolhauser i

Subject:

RHR Pumps

, ' Bingham Purp Serial Nos. 270839/42 16 x 1B x 26 CVIC Gentlement The recommended continuous minimum flow for these pumps is 2700 GPM.

This minimum flow value is given for continuous operation.

Continuous hours of operation. operation is considered as more than 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s-in any 24 For intermittent operation the minimum flow can be reduced to 2075.

If you should.have any questions, please let us know.

VerytrIlyyours, DINGHAM-WILLAMETTE COMPANY

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G<rdon J. Parks Supervisor, Field Service G39/11 i) /G. 6 g.cr Wa- Inu: ,

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2M1/87 BRAIDWOOD 1 - LEAKAGE INTO CONTROL ROOM JANUARY 20, 1987 - (J. STEVENS, NRR)

PROBLEM:

MILD PHOSPHORIC ACID SOLUTION DRAINED INTO CONTROL ROOM CAUSE:

OBSTRUCTION IN FLOOR DRAINAGE SYSTEM SIGNIFICANCE: .

POTENTIAL FOR CONTROL ROOM FLOODING DISCUSSION:

PLANT IN COLD SHUTDOWN (MODE 5). NO APPRECIABLE FISSION PRODUCTS SINCE PLANT HAS NOT YET REACHED CRITICALITY, AT APPR0XIMATELY 2205, SOLUTION STARTED ENTERING CONTROL ROOM

- ACTIONS TAKEN TO PROTECT MAIN CONTROL BOARD AT APPR0XIMATELY 2245, SOLUTION SOURCE IDENTIFIED AND STOPPED AT APPROXIMATELY 2245, NUCLEAR INSTRUMENTATION CABINETS WERE DEACTIVATED FOR PROTECTION OF EQUIPMENT

- BOTH REQUIRED SOURCE RANGE MONITORS PREVIOUSLY IN0PERABLE AND LCO IN EFFECT

- DEACTIVATION AFFECTED ALL SOURCE RANGE, INTERMEDIATE RANGE, AND POWER RANGE MONITORS

- REACTOR TRIP SIGNAL INITIATED FROM MONITORS

- NO R0D MOVEMENT SINCE REACTOR TRIP SIGNAL BREAKERS ALREADY OPEN AT 0140, CONTROL ROOM OPERATIONS WITH THE EXCEPTION OF THE NUCLEAR INSTRUMENTATION WERE RETURNED TO NORMAL

- NUCLEAR INSTRUMENTATION BACK IN SERVICE WITHIN 24 HOURS FOLLOW-UP:

! LICENSEE WILL DETERMINE POSSIBLE CORRECTIVE ACTIONS

LICENSEE WILL ENSURE OBSTRUCTION TOTALLY CLEARED FROM l DRAIN LINE REGION III WILL FOLLOW-UP ON CORRECTIVE MEASURES NRR WILL FOLLOW-UP ON FIRE PROTECTION REQUIREMENTS l

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REACTOR SCRAM

SUMMARY

WEEK ENDING O2/08/87 6

I. PLANT SPECIFIC DATA UNIT POWER RPS CAUSE COMPLI- YTD YTD YTD DATE SITE TOTAL CATIONS ABOVE BELOW 15% 15%-

2 100 M PERSONNEL NO 1 0 1 02/05/G7 SAN ONOFRE O 2 2 02/05/G7 BVRCf4 2 6A PERSONNEL NO .

5A EQUIPMENT NO O 1 1 02/07/87 SAINT LUCIE 1 NO 2 0 2 02/07/87 BEAVER VALLEY 1 100 A EOUIP/ MECH.

5 4

II. COMPARISON OF WEEKLY STATISTICS WITH INDUSTRY AVERAGES SCRAMS FOR WEEK ENDING O2/08/87 SCRAM CAUSE POWER NUMBER 1987 1986 1985 OF WEEKLY WEEKLY WEEKLY SCRAMS (5) AVERAGE AVERAGE AVERAGE YTD (3) ( ) (8) (9)

• + FOWER >15%

EQUIP. RELATED >15% 2 3.9 4.3 5.4 PERS. RELATED(6) >15% 0 1.4 1.8 2.0 OTHER(7) >15% 0 0.9 0.4 0.6

• • Subtotal +*

2 6.2 6.5 8.0

• • POWER <15%

• EQUIP. RELATED 015% 1 0.9 1.4 1.3 PERS. RELATED <15% 1 0.9 0.8 0.9 OTHER <15% 0 0.0 0.2 0.2 o* Subtotal **

2 1.8 2.4 2.4

• • • Total ***

4 S.0 8.9 10.4 MANUAL VS AUTO SCRAMS TYPE NUMBER 1987 1986, 1965 OF WEEKLY WEEKLY WEEKLY U SCRAMS AVERAGE AVERAGE AVERAGE YTD MANUAL SCRAMS 1 0.9 1.0 1.0 AUTOMATIC SCRAMS 3 7.1 7.9 9.4

NOTES

1. PLANT SPECIFIC DATA BASED ON INITIAL REVIEW 0F 50.72 REPORTS FOR THE WEEK 0F INTEREST. PERIOD IS MIDNIGHT SUNDAY THROUGH MIDNIGHT SUNDAY SCRAMS ARE DEFINED AS REACTOR PROTECTIVE ACTUATIONS WHICH RESULT IN R00 MOTION, AND EXCLUDE PLANNED TESTS OR SCRAMS AS PART OF PLANNED SHUTDOWN IN ACCORDANCE WITH A PLANT PROCEDURE.

2. REC 0VERY COMPLICATED BY EQUIPMENT FAILURES OR PERSONNEL ERRORS UNRELATED TO CAUSE OF SCRAM.

3. 1986 INFORMATION DERIVED FROM ORAS STUDY OF UNPLANNED REACTOR TRIPS IN 1986. WEEKLY DATA DETERMINED BY TAKING TOTAL TRIPS IN A GIVEN CATEGORY AND DIVIDING BY 52 WEEKS / YEAR.

4. IN 1986, THERE WERE AN ESTIMATED TOTAL OF 461 AUTOMATIC AND MANUAL UNPLANNED REACTOR TRIPS AT 104 REACTORS (HOLDING OPERATINGLICENSES). THIS YIELDS AN AVERAGE RATE OF 4.4 TRIPS PER REACTOR PER YEAR AND AN AVERAGE RATE OF 8.8 TRIPS PER WEEK FOR ALL REACTORS.

5. BASED ON 104 REACTORS HOLDING AN OPERATING LICENSE.

6. PERSONNEL RELATED PROBLEMS INCLUDE HUMAN ERROR, PROCEDURAL DEFICIENCIES, AND MANUAL STEAM GENERATOR LEVEL CONTROL PROBLEMS.

7. "0THER" INCLUDES AUTOMATIC SCRAMS ATTRIBUTED TO ENVIRONMENTAL CAUSES (LIGHTNING), SYSTEM DESIGN, OR UNKNOWN CAUSE.

8. 1985 INFORMATION DERIVED FROM AN ORAS STUDY OF UNPLANNED REACTOR TRIPS IN 1985. WEEKLY DATA DETERMINED BY TAKING TOTAL TRIPS IN A GIVEN CATEGORY AND DIVIDING BY 52 WEEKS / YEAR.

9. IN 1985, THERE WERE AN ESTIMATED TOTAL OF 541 AUTOMATIC AND MANUAL UNPLANNED REACTOR TRIPS AT 93 REACTORS (HOLDING FULL POWER LICENSES). THIS YIELDS AN AVERAGE RATE OF 5.8 TRIPS PER REACTOR YEAR AND AN AVERAGE RATE OF 10.4 TRIPS PER WEEK FOR ALL REACTORS.

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4 ENCLOSURE 3

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'( SISNIFICANT EVENTS FREQUENCY PERFCRMANCE INDICATCR No. 3 EVENT. EVENT DESCRIPTION CAUSE QTR PLANT NAME I CATE I ERAIDWCOD 1 01/29/87 DILUTED FHOSPHCRic ACID SOLUTION DRAINED INTO CCNTROL MAINTEhANCE I RDOM DUE TO OBSTRUCTION IN FLOOR DRAINA6E SYSTEM INDIAN FDINT 2 01/28/97 STATION CLASS IE BATTERY FOUND BELCW TEMP. REQUIRED TO PERSONNEL ERRCR I b DELIVER REQUIRED CAPACITY. COLD WEATHER INDUCED PROBLEM INDIAN POINT 2 02/03/S7 TW3 MOTOR DRIVEN AFW PUMPS FAIL 18 MONTH SURVEILLANC TEST EQUlFMENT FAILURE 2 DUE TO LEAKAGE CF VALVE ON RECIRCULATION LINE OF CST C VER;CNT Y N EE 01/29/97 REVISED MINIMUM FLOW VALUES OF RHR PUMPS SPECIFIED BY DESIGN / CONSTRUCTION /INSTALLATICNERRER I MANUFACTURER MAY REQUIRE SUBSTANTIAL MODIFICATIONS OF MINIFLCW LOGIC AND/CR MINIFLOW LINES C

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