ML20149E893
| ML20149E893 | |
| Person / Time | |
|---|---|
| Issue date: | 07/02/1997 |
| From: | Les Cupidon NRC |
| To: | Siegel C NRC |
| Shared Package | |
| ML20148T199 | List: |
| References | |
| FOIA-97-177 NUDOCS 9707210119 | |
| Download: ML20149E893 (111) | |
Text
-
I i
TO:,
Cherie Siegel
. FROM:
Les Cupidon DATE:
July 2,1997 TOPIC:
Response to Request of information concerning check valves manufactured by_
' Crane Company and Chapman Company.
This is a response to the FOIA-97-177 request for information concerning LERs on Crane Company and Chapman Company check valves from 1985 until the present.
- The following two list are lists of LERs from NPRDS (Nuclear Plant Reliability Data System) and SCSS (Sequence Coding Search System). These lists are current up to July 1,1997.
Attached are the LER abstracts. To maintain consistency the abstracts for the four LERs found
.in NPRDS were downloaded from the SCSS data base as well.
LIST OF NPRDS LERs This list of four LERs was found in NPRDS under various key word searches. However, these LERs were not found in the SCSS search.-
The manufacturers of the valves in the NPRDS LERs are as follows: Crane Company, Crane Valve Products.
LER Number Date Discovered l
- 1) 213-86006 01/22/86
- 2) 220-90007 05/09/90
- 3) 247-86035 10/23/86
- 4) 302-88014 06/19/88 9707210119 970710'
'\\M PDR FOIA FOWLER 97-177 PDR
LIST OF SCSS LERs 1
The manufacturers of the valves in the SCSS LERs are as follows: Chapman Division of Crane Company, Chapman Valve & MFG., Crane Company, Crane Valve Company.
LER Number Date Discovered LER Number -
Date Discovered
- 1) 155-91013 11/30/91
- 25) 254-87016 09/12/87
- 2) 213-87010 07/30/87
- 26) 254-90010 05/22/90 -
- 3) 213-87011 07/21/87
- 27) 254-90029 11/15/90
- 4) 220-88005 02/20/88
- 28) 254-92020 09/21/92
- 5) 220-88011' 03/25/88_
- 29) 254-94005 03/14/94 6)_ 220-93003 03/07/93
- 30) 255-94006 02/17/94 l
7).237-86019 08/11/86
- 31) 265-85007 03/18/85
'8) 237-90009 09/23/90
- 32) 265-86014 10/12/86
- 9) 237-92031 09/28/92
- 33) 265-88007 04/20/88 -
i
- 10) 237-93002 01/21/93
- 34) 265-90003 02/05/90 1
- 11) 237-95018 06/10/95
- 35) 265-92002 01/03/92
- 12) 245-89021 10/19/89
- 36) 269-86008 05/10/86
- 13) 245-93009 07/06/93
- 37) 281-86020 12/09/86 l
- 14) 245-96012 12/03/95
- 38) 285-88010 04/15/88 1
- 15) 247-36036 10/23/86
- 39) 286-86003 05/19/86 j'
_16) 247-88006 06/17/88
- 40) 286-91005 03/22/91
- 17) 249-89009 12/07/89
- 41) 289-93007 10/08/95 4
- 18) 249-91007 09/10/91
- 42) 318-88003 03/17/83
- 19) 249-93006 03/13/93
- 43) 324-86017 06/18/86
- 20) 249-94009 03/12/94
- 44) 333-90012 04/04/90 2i) 249-95003 02/03/95
- 45) 373-85052 06/29/85
- 22) 250-96002 02/09/96
- 46) 409-86038 12/16/86
- 23) 254-86001 01/06/86
- 47) 483-85039 08/20/85
- 24) 254-87001 01/05/87
Attachment:
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http /scss.ornl. gov / scripts /scss/results/rLERDetil.cfm?lernmbr=21386006
/
LER ll:ader Listing for 21386006
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l LER Header Listingfor LER Number: 2H86006 Docket Number Year Report #
Rev #
DCS Number Event Date 213 1986 006 2
8610210124 01/15/86 Nuclear Plant: Haddam Neck Unit #1 NRC Region:
I f
Facility Operator: Conn. Yankee Atomic Power & Light Co.
Architect Engineer: Stone & Webster Engr., Corp.
NSSS Vendor: Westinghouse Reactor Type: Pressurized Water Reactor Comm. Operations Date: 01/01/68 M M Mh!.M4 ABSTRACT POWER LEVEL - 000%. CONTAINMENT PENETRATION LOCAL LEAK RATE TESTING WAS CONDUCTED THROUGHOUT THE 1986 REFUELING OUTAGE IN ACCORDANCE WITH 10CFR50 APPENDIX J AND TECHNICAL SPECIFICATION 4.4.II.A. ON JANUARY 15,1986,IT WAS DETERMINED TIIAT THE
SUMMARY
OF PENETRATIONS TESTED EXCEEDED THE LEAK RATE ALLOWED BY TECHNICAL SPECIFICATION 4.4.II.B. AFTER RE-ESTABLISHING CONTAINMENT INTEGRITY PER TECHNICAL SPECIFICATION 3.11.B J
ON APRIL 29,1986, PENETRATION P-60 (CC-CV-885) WAS TESTED. THE PENETRATION j
TESTED EXCEEDED BY ITSELF THE LEAK RATE ALLOWED BY TECHNICAL SPECIFICATION 4.4.II.B. THE EXCESSIVE LOCAL LEAK RATE DURING BOTH EVENTS WAS CAUSED BY LEAKAGE THROUGH CONTAINMENT ISOLATION VALVES.
PREVENTATIVE AND CORRECTIVE ACTION HAS BEEN TAKEN TO REDUCE THE LEAKAGE TO WITIllN ACCEPTABLE LIMITS. THE ISOLATION VALVES THAT EITHER INDIVIDUALLY EXCEEDED THE TECHNICAL SPECIFICATION LEAKAGE CRITERIA OR WERE SIGN!FICANT CONTRIBUTORS ARE LISTED UNDER ' FAILURE ANALYSIS SECTION'. REVISION 2 PROVIDES ADDITIONAL INFORMATION CONCERNING PENETRATION FAILURE, SAFETY CONSEQUENCES, CORRECTIVE ACTION, AND INCLUDES INFORMATION ON THE APRIL 29,1986 PENETRATION FAILURE.
Unit Conditions at Time of Reportable Event:
Power Level: 0%
Operating Mode: Unknown Reportability Reason (s) for This LER:
10 CFR 50.73(a)(2)(i) - Shutdowns or Technical Specification Violations. [10]
Primary Cause(s) for This LER:
There is no Primary Cause for This LER Iof2 07/02/97 10:55:54
('LERIIcader Listing for 21386006 http 1/scss.ornl. gov / scripts /scss/results/rL E RDetl l.c fm71ernmbr=21386006 Emergency Classification (s) for This LER:
There is no Emergency Classification for This LER Referenced LERs:
LER Event j
Number Reactor UnitName Date LER Title j
i 2.1381018-Iladdam Neck Unit #1 11/14/81-The title for this LER is not currently available.
21383004:
lladdam Neck Unit #1 i 01/26/83 lThe title for this LER is not currently available.
21384012 fladdam Neck Unit #1 l 08/19/84 The title for this LER is not currently available.
Contact Mike Poore at ORNL with uestions or comments concerning SCSS Web site content.
Contact Dale Yellding at NRC with quest ns concerning SCSS Web site access.
Copyright O 1996 Oak Ridge National Laboratory (ORNL)
Last modified: January 31,1997 l
i
- 2 of 3 07/0 197 10:55:54
LER lisader Listing for 22090007 htt p://scss.orn l.go v/scr ipts/scss/results/rLE RDet i l.c fm71ernmbr=22090007 bggig;gj[gggghm
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LER Header Lis?ingfor LER Number: 22090007 Docket Number Year Report #
Rev#
DCS Number Event Date 220 1990 007 0
9000000000 05/15/90 Nuclear Plant: Nine Mile Point Unit #1 NRC Region:
1 Facility Operator: Niagara Mohawk Power Corporation j
Architect Engineer: Niagara Mohawk Power Corporation NSSS Vendor: General Electric Reactor Type: Boiling Water Reactor
]
Comm. Operations Date: 12/01/69
- nZ 3 HMRMi#iiM IR BIBiiGMl ABSTRACT POWER LEVEL - 000%. ON 5/15/90, AT 2215 HOURS, WITH NINE MILE POINT UNIT 1 IN A MAJOR REFUELING OUTAGE, THE CONTAINMENT SPRAY INLET CHECK VALVE #80-38 WAS DETERMINED TO LEAK IN THE REVERSE FLOW DIRECTION RESULTING IN THE UNSATISFACTORY PERFORMANCE OF SURVEILLANCE TEST PROCEDURE N1-ST-R10,
'DRYWELL TO TORUS LEAK RATE TEST'. AN INSPECTION WAS PERFORMED OF THE VALVE INTERNALS WIIICH IDENTIFIED THAT A PLASTIC SAMPLE BOTTLE WAS WEDGED BETWEEN THE DISC AND SEAT OF THE VALVE, HOLDING THE VALVE OPEN.
Tile SOURCE OF THE BOTTLE BEING INTRODUCED IN TO THE SYSTEM IS UNKNOWN. IT APPEARS TilAT THE BOTTLE MAY IIAVE BEEN INTRODUCED INTO THE SYSTEM DURING ORIGINAL CONSTRUCTION, DURING THE INSTALLATION OF THE CONTAINMENT SPRAY RAW WATER HEAT EXCHANGERS OR DURING THE INSTALLATION OF TIIE CONTAINMENT SPRAY CROSS TIE PIPING, DURING THE 1984-1986 TIME FRAME CORRECTIVE ACTIONS INCLUDED REMOVAL OF TILE PLASTIC SAMPLE BOTTLE, A VISUAL INSPECTION OF THE OTHER CONTAINMENT SPRAY INLET CHECK VALVE AND SUCCESSFUL COMPLETION OF REVERSE FLOW TESTING AND DRYWELL TO TORUS LEAK RATE TESTING PER PROCEDURE N1-ST-R10.
Unit Conditions at Time of Reportable Event:
Power Level: 0%
Operating Mode: Unknown Reportability Reason (s) for This LER:
No CFR - Special Report, Part 21 Report, etc. [21]
Primary Cause(s) for This LER:
Human Action [30 )
Emergency Classification (s) for This LER:
Iof2 07/02/97 10:59:30
LER licader Listing for 22090007 http1/scss.ornl.g v/ scripts /scss/results/rLERDet!!.cfm?lemmbr=22090007 There is no Emergency Classification for This LER j
Referenced LERs:
There are no Referenced LERs for This LER Contact Mike Poore at ORNL with questions or comments concerning SCSS Web site content.
Contact Dale Yellding at NRC with questions concerning SCSS Web site access.
Copyright C 1996 Oak Ridge National Laboratory (ORNL)
Last modified: January 31,1997
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07/02/97 10:59:30
^ ' LER llender Listing for 24786035 http://scss.ornl. gov / scripts /scss/results/rLERDetl l.cfm?lernmbr=24786035
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LER Header Listingfor LER Number: 24786035 Docket Number Year Report #
Rev#
DCS Number Event Date 247 1986 035 0
8612010147 10/20/86 Nuclear Plant: Indian Point Unit #2 NRC Region:
1 Facility Operator: Consolidated Edison Co.
Architect Engineer: United Engineers & Construction NSSS Vendor: Westinghouse Reactor Type: Pressurized Water Reactor Comm. Operations Date: 08/01/74 TGulMiliiWil DWfEsiefl E*llidfiP51 ABSTRACT POWER LEVEL - 100%. UNIT 2 REACTOR TRIPPED FROM 100% POWER WHEN REACTOR TRIP BREAKER (BKR) 'B' OPENED. OME OF THE REACTOR PROTECTION (AA) RELAYS (RLY) DEENERGlZED WHILE A MONTHLY SAFETY INJECTION SURVEILLANCE TEST WAS BEING PERI'ORMED IN A NEARBY EQUIPMENT RACK. SEVERAL LOOSE CONNECTIONS IN THE ASSOCIATED CIRCUITRY WEIG DISCOVERED DURING THE SUBSEQUENT TROUBLESilOOTING. SAFEGUARDS EQUIPMENT FUNCTIONED NORMALLY EXCEPT FOR PORTIONS OF THE AUXILIARY FEEDWATER SYSTEM.
FEEDWATER WAS MAINTAINED BY ONE OF THE MOTOR DRIVEN AUXILIARY FEED PUMPS (P). Tile OTHER MOTOR DRIVEN AUXILIARY FEEh PUMP (P) TRIPPED AFTER STARTING AND WAS SUBSEQUENTLY RESTARTED SUCCESSFULLY BY THE CONTROL ROOM OPERATOR. IN ADDITION, THE RELIEF VALVE (RV) ON THE TURBINE DRIVEN AUXILIARY FEED PUMP (P) LIFTED WHEN ITS STEAM CONTROL VALVE (PCV) OPENED.
Tile RELAY RACK TERMINALS WERE CHECKED FOR LOOSE CONNECTIONS, THE AUXILIARY FEED PUMP CIRCUlT BREAKER TRIP POINT WAS READJUSTED, AND THE AUX 11,IARY FEEDWATER SYSTEM (BA) WAS RETESTED AND RESTORED TO OPERABLE CONDITION THE HEALTil AND SAFETY OF THE PUBLIC WERE NOT AFFECTED.
Unit Coaditions at Time of Reportable Event:
Power Level: 100%
Operating Mode: Unknown Reportability Reason (s) for This LER:
10 CFR 50.73(a)(2)(iv)- ESF Actuations. [13]
Primary Cause(s) for This LER:
There is tio Primary Cause for This LER Emergency Classification (s) for This LER:
Iof2 07/02/97 11:00:30
http /scss.ornl. gov / scripts /scss/rtsults/rLERDetll.cfm71ernmbr=24786035 1
. LER Header Listing for 24786035 There is no Emergency Classification for This LER Referenced LERs:
There are no Referenced LERs for This LER Contact Mike Poorg at ORNL with uestions or comments concerning SCSS Web site content. ~
Contact Dale Yellding at NRC with quest as concerning SCSS Web site access.
Copyright C 1996 Oak Ridge National Laboratory (ORNL)
Last modified: January 31,1997
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j 2 of 2 07/02/97 11:00:30
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http /scss.ornl. gov / scripts /scss/results/rLERDetll.cfm?lernmbr=30288014 1
LER lisader Listing for 3028801/
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LER Header Listingfor LER Number: 30288014 Docket Number Year Report #
Rev#
DCS Number Event Date 302 1988 014 1
8901270305 06/21/88 Nuclear Plant: Crystal River Unit #3 NRC Region:
2 Facility Operator: Florida Power Corp.
Architect Engineer: Gilbert Associates,Inc.
NSSS Vendor: Babcock & Wilcox Reactor Type: Pressurized Water Reactor Comm. Operations Date: 03/13/77 M IMEMEK4 MEMBRE ABSTRACT POWER LEVEL - 100%. ON 6/21/88, CRYSTAL RIVER UNIT 3 WAS OPERATING IN MODE 1 AT 100% RATED TIIERMAL POWER, GENERATING 880 MWE. ELEVATED TEMPERATURES EXCEEDING THE SYSTEM DESIGN WERE PRESENT IN PART OF Tile EMERGENCY FEEDWATER SYSTEM. IN ADDITION, THIS EVENT CAUSED A MECHANICAL CONTAINMENT PENETRATION TO EXCEED ITS DESIGN TEMPERATURE SYSTEM WALKDOWNS FOLLOWING THE EVENT ALSO REVEALED EXPANSION ANCHORS OF ONE PIPE RESTRAINT WERE PARTIALLY PULLED I.OOSE FROM THEIR STRUCTURAL ATTACHMENT. THE ELEVATED TEMPERATURE RESULTED FROM THE COMBINATION OF A SMALL LEAK PAST THE SEAT OF CHECK VALVE FWV-43 AND SMALL BODY TO BONNET LEAKS IN VALVES EFV-18 AND EFV-33. A SYSTEM TO REINJECT THE LEAKOFF WATER INTO Tile EMERGENCY FEEDWATER PIPING WAS INSTALLED ON JUNE 22,1988.
THIS SYSTEM WAS SUCCESSFUL IN REDUCING THE PIPE TEMPERATURES TO WITHIN DESIGN VALUES. TESTS. INSPECTIONS, AND ANALYSES WERE PERFORMED TO DETERMINE Tile EFFECT OF THE ELEVATED TEMPERATURES ON THE EFW SYSTEM. NO DAMAGE WAS FOUND OR ANALYZED TO HAVE OCCURRED WHICH WOULD HAVE LEFT THE EMERGENCY FEEDWATER SYSTEM UNABLE TO PERFORM ITS FUNCTION. THE LEAKING VALVES WERE REPAIRED DURING A COLD SHUTDOWN OUTAGE IN OCTOBER,1988.
Unit Conditions at Time of Reportable Event:
Power Level: 100%
Operating Mode: Unknown Reportability Reason (s) for This LER:
10 CFR 50.73(a)(2)(ii)- Unanalyzed Conditions. [11]
Primary Cause(s) for This LER:
Equipment Failure (20 ]
I of 2 07/02/97 11:01:22
http://scss.ornl. gov / scripts /scss/resultvrLERDetil.c fm71ernmbra3
' LER Header Listing for 30288014 Emergency Classification (s) for This LER:
There is no Emergency Classification for This LER
' Referenced LERs:
There are no Referenced LERs for This LER Contact Mike Poore at ORNL with uestions or comments concerning SCSS Web site content.
as concerning SCSS Web site access.
Contact Dale Yellding at NRC with quest Copyright C 1996 Oak Ridge National Laboratory (OHNL)
Last modified: January 31,1997 i
07/02/97 11:01:22
- 2 of 2 -
w http /scss.ornl. gov / scripts /scss/results/rLERDetll.cfra71ernmbr=155910D 1
LER ileader Listing for 15591013 g,O'\\
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mm mm LER Header Listinefor LER Number: 15591013 Docket Number Year Report #
Rev#
DCS Number Event Date 155 1991 013 0
9205290268 11/30/91 Nuclear Plant: Big Rock Point Unit #1 NRC Region:
3 Facility Operator: Consumers Power Co.
Architect Engineer: Bechtel Power Corp.
NSSS Vendor: General Electric Reactor Type: Boiling Water Reactor Comm. Operations Date: 03/29/63 rniliiiiiiiMip1 Fuiiswism Fmere4 ABSTRACT POWER LEVEL - 000%. DURING A PLANNED REACTOR SHUTDOWN FOR THE 1991/92 REFUELING OUTAGE Tile #1 SHUTDOWN COOLING LOOP WAS INSERVICE PROVIDING DECAY llEAT REMOVAL REACTOR SYSTEM TEMPERATURE WAS APPROXIMATELY 134F AND Tile PLANT WAS IN A COLD SHUTDOWN CONDITION WITH ALL CONTROL RODS INSERTED. ON NOVEMBER 30,1991 (1735 IIOURS) THE #2 SilUTDOWN COOLING LOOP WAS PLACED INSERVICE TO FLUSH THE PIPING TO REDUCE RADIATION LEVELS IN Tile AREA. ABOUT AN IIOUR LATER, REACTOR TEMPERATURE INCREASED ABOUT 6F INDICATING INADEQUATE COOLING. AT THAT TIME THE #1 LOOP WAS RETURNED TO SERVICE AND REACTOR COOLDOWN WAS RE-ESTABLISHED. FOLLOWING INSPECTION OF THE #2 SilUTDOWN COOLING LOOP, THE DISCHARGE CHECK VALVE WAS FOUND RESTRICTED AND DIFFICULT TO OPEN BEYOND THE 30 DEGREE OPEN POSITION. UPON DISASSEMBLY, THE PlVOT PINS WERE FOUND BENT CAUSING DISC BINDING. THE PROBABLE CAUSE OF Tile BENT PINS WAS IMPROPER VALVE ASSEMBLY FOLLOWING AN ISI EXAMINATION ON JULY 7,1989. ON JANUARY 22,1992, Tile CHECK VALVE WAS REPAIRED AND RETURNED TO SERVICE.
Unit Conditions at Time of Reportable Event:
Power Level: 0%
Operating Mode: Unknown Reportability Reason (s) for This LER:
No CFR - Special Report, Part 21 Report, etc. [21]
Primary Cause(s) for This LER:
lluman Error [35 ]
Poor Ergonomics or lluman Environment [38 ]
Iof2 07/02/97 09A9:31
=
LER Header Listing for 15591013 http1/scss.ornl. gov / scripts /scas/results/rLERDetll.c fm?lernmbr= 15591013 Emergency Clas lfication(s) for This LER:
There is no Emergency Classification for This LER Referenced LERs:
There are no Referenced LERs for This LER Contact Mike Poore at ORNL with uestions or comments concerning SCSS Web site conter t.
Contact Dale Yellding at NRC with quest ns concerning SCSS Web site access.
Copyright C 1996 Oak Ridge National Lahorntory (ORNL)
Last modified: January 31,1997 2cf2I 07/02/97 09:49:3I
LER Header Listing for 21387010 http://scss.oml. gov / scripts /scss/results/rLE RDetl l.c fm?le mm br=21387010 O%
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LER Header Listingfor LER Number: 21387010 Docket Number Year Report #
Rev #
DCS Number Event Date 213 1987 010 1
8901030196 07/30/87 Nuclear Plant: Haddam Neck Unit #1 NRC Region:
1 Facility Operator: Conn. Yankee Atomic Power & Light Co.
Architect Engineer: Stone & Webster Engr., Corp.
NSSS Vendor: Westinghouse Reactor Type: Pressurized Water Reactor Comm. Operations Date: 01/01/68 EMBMd EMU MidiDEEl ABSTRACT POWER LEVEL - 000%. HIGH PRESSURE SAFETY INJECTION SYSTEM CHECK VALVES SI-CV-862A, B, C, AND D ARE PERIODICALLY LEAK TESTED AS PRESSURE ISOLATION VALVES TO VERIFY COMPLIANCE WITH TECHNICAL SPECIFICATION 3.14. ON JULY 30, 1987, WITH Tile PLANT SHUTDOWN IN MODE 6, SAFETY INJECTION CHECK VALVE SI-CV-862B FAILED THE LEAKAGE TEST BY EXCEEDING THE 1 GPM TECHNICAL SPECIFICATION LIMIT. IMPROPER SEATING DUE TO EXCESSIVE DISC HINGE WEAR.
DISC HINGE, AND DISC SHAFT. A RETEST WAS PERFORMED SEPTEMBER 19,1981 WITH SATISFACTORY RESULTS. THIS EVENT IS REPORTABLE PER 10CFR50.73(A)(2)(I) SINCE IT INVOLVES A COFDITION PROHIBITED BY TECHNICAL SPECIFICATIONS. THIS REVISION HAS BEEN ISSUED TO PROVIDE ADDITIONAL INFORMATION ON THE EVENT DESCRIPTION, THE CAUSE AND THE CORREC flVE ACTIONS TAKEN AND PLANNED.
Unit Conditions at Time of Reportable Event:
Power L.evel: 0%
Operating Mode: Unknown Reportability Reason (s) for This LER:
10 CFR 50.73(a)(2)(i) - Shutdowns or Technical Specification Violations. [10]
Primary Cause(s) for This LER:
Equipment Failure [20 ]
Emergency Classification (s) for This LER:
There is no Emergency Classification for This LER Referenced LERs:
There are no Referenced LERs for This LER I of't 07/02/97 09:50:47
LER Header Listing for 21387010 http://scss.orntg:v/ scripts /scss/results/rLERDetll.cfm?lernmbr=21387010 Contact Mike Poore at ORNL with guestions or comments concerning SCSS Web site content.
Contact Dale Yellding at NRC with questens concerning SCSS Web site access.
Copyright C 1996 Oak Ridge National Laboratory (ORNL)
L.ast modified: January 31,1997 l
1 i
il
^ 2cf2 07/02/97 09:50:47
http /scss.omt. gov / scripts /scas/results/rLERDetll.cfm?lernmbr=2138701!
LER lleader Listing for 21387011 J
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LER Header Listingfor LER Number: 21387011 Docket Number Year Report #
Rev#
DCS Number Event Date 213 1987 011 1
8712090070 07/21/87 Nuclear Plant: IIaddam Neck Unit #1 NRC Region:
1 Facility Operator: Conn. Yankee Atomic Power & Light Co.
Architect Engineer: Stone & Webster Engr., Corp.
NSSS Vendor: Westinghouse Reactor Type: Pressurized Water Reactor Comm. Operations Date: 01/01/68 M Ff6iiiiiHiRSI L*RMilHZl ABSTRACT POWER LEVEL 000%. CONTAINMENT PENETRATION LOCAL LEAK RATE TESTING (TYPE B AND C) WAS CONDUCTED DURING THE REFUELING OUTAGE IN ACCORDANCE WITH IOCFR50 APPENDIX J AND TECHNICAL SPECIFICATION 4.4.11. ON JULY 21,1987, WITH THE PLANT SHUTDOWN IN MODE 5, CONTAINMENT PENETRATION P-30 (SPACE IIEATING STEAM SUPPI.Y, HS-CV-295 AND llS-CV-295A) FAILED ITS TYPE C LOCAL LEAK RATE TEST. THE TEST METIIOD WAS QUESTIONED BY SITE ENGINEERING AND A RETEST WAS PERFORMED. THE PENETRATION FAILED THE RETEST, ALTHOUGH ONE OF ITS TWO SERIES INSTALLED CHECK VALVES (HS-CV-295) DID PASS. THE CAUSE OF THE FAILURE WAS ATTRIBUTED TO IMPROPER SEATING AND VALVE MISAPPLICATION. THE PENETRATION WAS SUBSEQUENTLY MODIFIED WHICH ELIMINATED llS-CV-295 AND HS-CV-295A AS CONTAINMENT ISOLATION VALVES. THIS EVENT IS REPORTABLE PER 10CFR50.73 (A)(2)(I) SINCE IT INVOLVES A CONDITION PROlllBITED BY TECHNICAL SPECIFICATIONS. THIS REVISION HAS BEEN ISSUED TO
. PROVIDE ADDITIONAL INFORMATION ON LONG TERM CORRECTIVE ACTIONS TAKEN.
Unit Conditions at Time of Reportable Event:
Power Level: 0%
Operating Mode: Unknown Reportability Reason (s) for This LER:
10 CFR 50.73(a)(2)(i) - Shutdowns or Technical Specification Violations. [10]
Primary Cause(s) for This LER:
Equipment Failure [20 ]
Procedural Deficiency [40 ]
Emergency Classification (s) for This LER:
Iof2 07/02/97 09:52:17
LER lleader Listing for 21387011 http//scss.ornl. gov / scripts /scss/results/rLERDeti l.cfm?tirnmbr=2138701 1 There is no Emergency Classification for This LER '
Referenced LERs:
LER Event Number
. Reactor Unit Name Date LER Title 21381018:
11addam Neck Unit #1
. I1/14/81 The title for this LER is not currently available.
21383004-11addam Neck Unit #1 Tl/26/83 The title for this LER is not currently available.
21384012 lladdam Neck Unit #1 j 08/19/84 The title for this LER is not currently available.
21386006l 11addam Neck Unit #1 01/15/86 Local Leak Rate Testing Contact Mike Poort at ORNL with uestions or comments concerning SCSS Web site content.
Contact Dale Yellding at NRC with quest ons concerning SCSS Web site access.
Copyright O 1996 Dak Ridge National Laboratory (ORNL)
Last modified: January 31,1997 i
'2cf2' 07/02/97 09:52:17
LER ileader Listing for 22088005 htt p://scss.ornl. gov / scripts /scss/results/r L E RDetl l.cfm71ernmbr=22088005 s......
LER Header Listingfor LER Number: 2208800S i
Docket Number Year Report #
Rev#
DCS Number Event Date 220 1988 005 1
8903020638 02/20/88 l
Nuclear Piant: Nine Mile Point Unit #1 NRC Region:
1 i
Facility Operator: Niagara Mohawk Power Corporation Architect Engineer: Niagara Mohawk Power Corporation NSSS Vendor: General Electric Reactor Type: Boiling Water Reactor
. Comm. Operations Date: 12/01/69
[MWiiiiiiid RiiiiiEMM HIF616MiF48 ABSTRACT POWER LEVEL - 000%. THIS REPORT IS BEING SUBMITTED AS SUPPLEMENT 1 TO LER 88-05. ON FEBRUARY 19 AND 20,1988, WITil NINE MILE POINT UNIT I IN A REFUELING OUTAGE, BOTH CONTAINMENT ISOLATION VALVES OF THE LIQUID POISON SYSTEM FAILED TIIElR LOCAL LEAK RATE TEST. THE TESTS WERE BEING PERFORMED IN ACCORDANCE WITH 10 CFR 50 APPENDIX J REQUIREMENTS. THE RESULTS OF THESE TESTS DETERMINED TIIE MEASURED LEAKAGE TO BE GREATER THAN THE TECHNICAL SPECIFICATION ALLOWABLE LIMIT. THE ROOT CAUSE FOR FAILURE OF Tile CONTAINMENT ISOLATION VALVES WAS DETERMINED TO BE DIRT ACCUMULATION ON THE TEFLON SEATS, AND MINOR BINDING OF THE PIVOT PIN.
INITIAL CORRECTIVE ACTION INVOLVED DECLARING THE LIQUID POISON SYSTEM INOPERABLE AND GENERATING STATION WORK REQUESTS TO INSPECT THE VALVES AND REPAIR AS NECESSARY. THE LIQUID POISON SYSTEM VALVES WERE DISASSEMBLED AND INSPECTED. THE VALVES WERE REINSTALLED, AND THE SYSTEM WAS REASSEMBLED. THE VALVES SATISFACTORILY PASSED THE LLRT.
Unit Conditions at Time of Reportable Event:
Power Level: 0%
Operating Mode: Unknown Reportability Reason (s) for This LER:
10 CFR 50.73(a)(2)(i)- Shutdowns or Technical Specification Violations. [10]
10 CFR 50.73(a)(2)(ii)- Unanalyzed Conditions. [11]
Primary Cause(s) for This LER:
Equipment Failure [20 ]
Emergency Classification (s) for This LER:
Iof2 07/02/97 09:54:29
=
http /scss.ornl. gov / scripts /scss/results/rLERDetll.cfm71ernmbr=22088005
/
' LER Header Listing for 22088005 There is no Emergency Classification for This LER Referenced LERs:
LER Event Number Reactor UnitName Date LER Title 22081016 Nine Mile Point Unit #1 05/15/81; The title for this LER is not currently available.
- 220810274 Nine Mile Point Unit #1 06/23/81 The title for this LER is not currently available.
22082017 Nine Mile Point Unit #1 12/06/82 The title for this LER is not currently available.
22082018 Nine Mile Point Unit #1 12/07/82 The title for this LER is not currently available.
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Copyright C 1996 Gak Ridge National Laboratory (ORNL)
Last modified: January 31,1997 l
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LER Ileader Listing for 2208801I http://scss.ornl. gov / scripts /scss/rrsults/rL E RDetl l.c fm71ernmbr=2208801 1 M
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LER Header Listinefor LER Number: 22088011 Docket Number Year Report #
Rev #
DCS Number Event Date 220 1988 011 0
8804280557 03/25/88 Nuclear Plant: Nine Mile Point Unit #1 NRC Region:
1 Facility Operator: Niagara Mohawk Power Corporation Architect Engineer: Niagara Mohawk Power Corporation NSSS Vendor: General Electric Reactor Type: Boiling Water Reactor Comm. Operations Date: 12/01/69 j
ff M~ DiiiiiEiEl 52BiREREM ABSTRACT POWER LEVEL - 000%. ON MARCH 25,1988, WITH NINE MILE POINT UNIT 1 (NMPI)IN A REFUELING OUTAGE, IT WAS DISCOVERED THAT 'l HE SUM OF THE TYPE B AND C TEST RESULTS EXCEEDED THE ALLOWABLE LIMIT (0.60 LA) AS DEFINED IN 10 CFR 50 APPENDIX J. UNDER THIS CONDITION, TIIE PRIMARY CONTAINMENT IS CONSIDERED INOPERABLE WITH RESPECT TO PROVIDING A LEAKAGE BOUNDARY AND REPRESENTS A DEGRADATION OF A PRINCIPAL SAFETY BARRIER. THE SUM OF THE TEST RESULTS WAS APPROXIMATELY 16,000 SCFD AT 35 PSIG, WHEREAS,0.60 LA FOR NMP1 IS 9274.84 SCFD AT 35 PSIG. THE ROOT CAUSE OF THIS EVENT IS THE AMOUNT OF LEAKAGE THROUGH THOSE VALVES THAT FAILED THEIR LOCAL LEAK RATE TESTS.
THERE WERE 18 VALVES AND ONE PENETRATION THAT FAILED AND SIGNIFICANTLY CONTRIBUTED TO THE OVERALL LEAKAGE RATE VALUE. ALSO DISCOVERED WAS A PROCEDURAL DEFICIENCY EXISTING IN THE CONTROLLING PROCEDURE AS IT DOES NOT REQUIRE A RUNNING TOTAL OF TEST RESULTS DE KEPT. THIS RESULTED IN A DELAY IN IDENTIFYING WHEN THE REGULATORY LIMIT WAS EXCEEDED.
CORRECTIVE ACTIONS FOR THE PENETRATION AND VALVE FAILURES CONSISTED OF DECLARING THE COMPONENTS INOPERABLE AND PERFORMING THE ASSOCIATED ADMINISTRATIVE FUNCTIONS. THESE COMPONENTS HAVE BEEN, OR WILL BE, REPAIRED OR REPLACED AND RETESTED DURING THE CURRENT REFUELING OUTAGE.
Unit Conditions at Time of Reportable Event:
Power Level: 0%
Operating Mode: Unknown Reportability Reason (s) for This LER:
10 CFR 50.73(a)(2)(ii) - Unanalyzed Conditions. [l 1]
Primary Cause(s) for This LER:
Iof2 07/02/97 09:55:13
http /scss.ornt. gov / scripts /scss/results/rLERDettl.cfm?lernmbr-22088011
. LER licader Listing for 2208801I 1
Equipment Failure [20 J Procedural Deficiency [40 ]
j i
Emergency Classification (s) for This LER*
i There is no Emergency Classification for This LER l
Referenced LERs:
j There are no Referenced LERs for This LER
/% '
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Contact Dale Yellding at NRC with questions concerning SCSS Web site access.
Copyright C 1996 Oak Ridge National Laboratory (ORNL)
- Last modified: January 31,1997 i
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LER Header Listing for 22093003 http /scss.ornl. gov / scripts /scss/results/rLERDetll.cfm?lernmbr=22093003 1
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LER Header Listingfor LER Number: 22093003 Docket Number Year Report #
Rev #
DCS Number Event Date 220 1993 003 1
9307130372 03/07/93 Nuclear Plant: Nine Mile Point Unit #1 NRC Region:
1 Facility Operator: Niagara Mohawk Power Corporation Architect Engineer: Niagara Mohawk Power Cornoration NSSS Vendor: General Electric Reactor Type: Boiling Water Reactor Comm. Operations Date: 12/01/69
[iiiiEtisiin9y Ff#iiiiffisiT1 FB0iiiiM1 ABSTRACT POWER LEVEL - 000%. On March 6,1993 and March 7,1993, with the Nine Mile Point Nuclear Station Unit 1 (NMP1) in a refueling outage and primar" containment not required, valves 33-01 R (Reactor Water Clean Up Inside Isolation Valve) and 33-03 (Reactor Water Clean Up Outside Isolation Valve), respectively, failed their Technical Specification required Local Leak Rate Test (LLRT)'imit of 5 percent L sub to (i.e.,13.07 Standard Cubic Feet per Hour [SCFli]). These valves are the isolation valves for primary containment penetration X-154. As a result of the failed LLRTs, the primary containment Technical Specification required leak rate limit of L sub t (348.85 SCFH at 22 psig) and the 10CFR50 Appendix J limit of 0.6 L sub a (386.45 SCFH at 35 psig) were exceeded. Under this condition, the primary containment is considered inoperable with respect to providing a leakage boundary and represents a degradation of a principal safety barrier. Root cause analyses were performed for both valve 33-01R and for valve 33-03. The cause of valve 33-01R failing its LLRT was improper lit-up of the valve's internal components and incorrect seat tightness when the valve was installed in 1991. The cause of valve 33-03 failing its LLRT was deterioration of the soft seat due to a design deficiency. The immediate corrective action was to declare the valves inoperable. Valves33-01R and 33-03 were repaired and they subsequently passed their LLRTs.
Unit Conditions at Time of Reportable Event:
Power Level: 0%
Operating Mode: Unknown Reportability Reason (s) for This LER:
10 CFR 50.73(a)(2)(ii) - Unanalyzed Conditions. [l 1]
Primary Cause(s) for This LER:
Design Error or inadequacy [34 }
Procedural Deficiency [40 ]
Iof2 07/02/97 09:55:46
LER IIcader Listing for 22093003 http://scss.ornl. gov / scripts /scss/res ults/rLE RDcti l.c fm71ernmbr=22093003 Emergency Classification (s) for This LER:
There is no Emergency Classification for This LER Referenced LERs:
LER Event Number Reactor UnitName Date LER Title 22088011 Nine Mile Point Unit #1 03/25/88 Summation o,f L,ocal Leak Rate Tests Exceed j
Regulatory Limit 22091005 Nine Mile Point Unit #1 02/24/91: Local Leak Rate Tests Exceed Regulatory Limit 1
/4 l
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)
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Last modified: January 31,1997 l
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http /scss.ornl. gov / scripts /scss/results/rLERDetil.cfm?lernmbr=23786019 1
LER HeaJer Listing for 23786019 lit1 M tttstitlMl M M f
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LER Header Listingfor LER Number: 23786019 Docket Number Year Report #
Rev #
DCS Number Event Date j
237 1986 019 1
8711200112 08/11/86 Nuclear Plant: Dresden Unit #2 NRC Region:
3 Facility Operator: Com.monwealth Edison Co.
Architect Engineer: Sargent & Lundy NSSS Vendor: General Electric Reactor Type: Boiling Water Reactor Comm. Operations Date: 06/09/70 fM IE6iiiifHiiEd Elidfdfdh3 ABSTRACT POWER LEVEL - 091%. ON 8/l1/86 AT 1433 HOURS, WITH UNIT 2 OPERATING AT 91%
POWER, THE MAIN TURBINE TRIPPED ON +55 INCHES OF REACTOR WATER LEVEL WHICil RESULTED IN A FULL REACTOR SCRAM FROM A TURBINE GENERATOR LOAD REJECTION SIGNAL. WlIILE TIIE NUCLEAR STATION OPERATOR (NSO) WAS REMOVING PIECES OF A BROKEN CONTROL ROOM INDICATING LIGHT BULB ON REACTOR RECIRCULATING PUMP LUBE OIL PUMP 'Bl', A SHORT CIRCUIT OCCURRED AND CAUSED TIIE 'B' RECIRCULATION PUMP TO TRIP. THE PUMP TRIP CAUSED THE REACTOR WATER LEVEL TO INCREASE TO +55 INCIIES AND CAUSED THE MAIN TURBINE TO TRIP AND A SUBSEQUENT REACTOR FULL SCRAM. WHEN TIIE NSO TRIED TO MANUALLY CLOSE THE 'A' FEEDWATER REGULATING VALVE, THE NSO NOTICED THE VALVE POSITION INDICATION SHOWED NO MOVEMENT. INSPECTION OF THE FAILED REGULATOR VALVE DISCOVERED A KEEPER FROM A FEEDWATER PUMP CHECK VALVE JAMMED UNDER THE VALVE'S SEAT. THE KEEPER WAS REMOVED AND Tile REGULATING VALVE WAS CYCLED SUCCESSFULLY. THE ROOT CAUSES OF THIS EVl?NT ARE: 1) FAILURE OF TILE NSO TO COMPLY WITH DRESDEN OPERATING PROCEDURE (DOP) 040-4, AND 2) FAILURE OF THE 'A' REACTOR FEEDWATER DISCilARGE CHECK VALVE CORRECTIVE ACTIONS INCLUDE: 1) OPERATOR TRAINING,
- 2) INVESTIGATION OF IMPROVED SEAT HOLD DOWN RING RETAINING METIIODS, AND
- 3) AMENDMENT TO DOP 3200.5, REACTOR FEED PUMP SHUTDOWN.
Unit Conditions at Time of Reportable Event:
Power Level: 91 %
Operating Mode: Unknown Reportability Reason (s) for This LER:
10 CFR 50.73(a)(2)(iv)- ESF Actuations. [13]
Primary Cause(s) for This LER:
Iof2 07/02/97 09:56:32
LER Header Listing for 23786019 -
http://scss.ornl. gov /rcripts/scss/results/rLERDetll.cfm?lernmbr=23786019 There is no Primary Cause for This LER Emergency Classification (s) for This LER:.
There is no Emergency Classification for This LER Referenced LERs:
j LER Event Number Reactor Unit Name '
Date LER Title j
23784009 Dresden Unit #2 06/21/84 The title for this LER is not currently available.
Contact Mike Poore at ORNL with uestions or comments concerning SCSS Web site content.
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Copyright C 1996 Gak Ridge National Laboratory (ORNL)
Last modified: January 31,1997 2ef2' 07/02/97 09:56:33
LER Header Listing for 23790009 h np://scss.orni. gov / scripts /scsvres u lts/rL E RDetl l.c fm ?!ernmbr=23790009 h.....)N h[')\\
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LER Header Listingfor LER Number: 23790009 Docket Number Year Report #
Rev #
DCS Number Event Date 237 1990 009 2
9208240177 09/23/90 Nucicar Plant: Dresden Unit #2 NRC Region:
3 Facility Operator: Commonwealth Edison Co.
Architect Engineer: Sargent & Lundy NSSS Vendor: General Electric Reactor Type: Boiling Water Reactor Comm.' Operations Date: 06/09/70
!%M EMiiiiW9M MMBERi ABSTRACT i
POWER LEVEL - 000%. On September 23,1990, with Unit 2 in a refuel outage during performance of Main Steam Isolation Valve (MSIV) Local Leak Rate Testing (LLRT), the A, C, and D main steam line volumes were found to be leaking in excess of the Technical Specification limit of 11.5 SCFH.
Further diagnostic testing indicated that MSIVs 2-203-1 A,1D, and 2C were the leaking valves. The cause of the leakage has been determined to be wear of the valve seating surfaces. On September 25, 1990, while performing further LLRTs, outboard primary containment drywell spray valve 2-1501-27A leaked an undetermined amount. This caused the as-found total type B and C leakage rate to be in excess of the Technical Specification limit. After flushing of the valve seat, the leak rate was reduced to a minimallevel. The safety significance for both events was minimal since in each case the in line isolation valves were not observed to be leaking. Of the remaining type B and C tests which demonstrated unsatisfactory leakages, the actual through leakage was determined to be minimal. On April 2,1991, while Unit 2 was shutdown for a short maintenance outage, six drywell bellows penetrations were inspected for cracks. Of the six, through wall cracks were found at penetration X144.
Ilowever, the safety significance is considered minimal due to the satisfactory performance of these bellows during the recent integrated Leak Rate Test (ILRT). A previous event mvolving the as found type B and C test results exceeding the 0.6 La limit is outlined in LER #88-004 on Docket #050249.
Unit Conditions at Time of Reportable Event:
Power Level: 0%
Operating Mode: Unknown Reportability Reason (s) for This LER:
10 CFR 50.73(a)(2)(i) - Shutdowns or Technical Specification Violations. [10]
Primary Cause(s) for This LER:
Equipment Failure [20 ]
Emergency Classification (s) for This LER:
I of 2 07/02/97 09:57:08
- LER Header Listing for 23790009 http://scss.ornl. gov / scripts /scss/results/rL E RDetl l.cfm71ernmbr=23790009 There is no Emergency Classification for This LER Referenced LERs:
LER^
Event Number Reactor Unit Name Date LER Title Leakage Path Discovered During Primary-23790018 Dresden Unit #2 12/18/90 Containment ILRT Due to Management 3
Deficiency Type B And C Local Leak Rate Test Limit 24988004 Dresden Unit #3 04/04/88 Exceeded Due to Leakage through Pnmary Containment Isolation Valve j
Local Leak Rate Testing "As Found" Limit i
24989009 Dresden Unit #3 12/07/89 Exceeded Due to Leakage From Primary Containment Valves i
Contact Mike Poore at ORNL with questions or comments concerning SCSS Web site content.
-l Contact Dale Yellding at NRC with questions concerning SCSS Web site access.
Copyright C 1996 Dak Ridge National Laboratory (ORNL)
-]
Last modified: January 31,1997 1
l 2 of 2 07/02/97 09:57:08
http hess.ornl. gov / scripts /scss'results/rLERDetll.cfm?lernmbr=23792031 l
LER lleader Listing for 23792031
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- M LER Header Listinefor LER Number: 23792031 Docket Number Year Report #
Rev#
DCS Number Event Date 237 1992 031 2
9306020332 09/28/92 Nuclear Plant: Dresden Unit #2 NRC Region:
3 Facility Operator: Commonwealth Edison Co.
Architect Engineer: Sargent & Lundy NSSS Vendor: General Electric Reactor Type: Boiling Water Reactor j
Comm. Operations Date: 06/09/70 f6iiitsiisiliis] 'FisiiiiiNisiM PMe'liliiFM ABSTRACT POWER LEVEL - 094%. At approximately 1910 hours0.0221 days <br />0.531 hours <br />0.00316 weeks <br />7.26755e-4 months <br /> on September 28,1992 with Unit 2 at 94%
power, the performance of Special Procedure 92-9-115 Revision 0 (Local Leak Rate Testing of Primary Containment Isolation Valves During Reactor operation) identified the Outboard Drywell Air Sample Valve 2-8501-5B to be leaking 74.28 scfh. This leakage, when added to the total primary containment maximum pathway leakage rate, brought the total leakage for Type B and C testable penetrations to 435.64 scfh. Revision 0 of this report was submitted since the value exceeded an administrative leakage liniit of 85% of 0.6L. (415.18 scfh) which was established as a condition of being granted a schedular exemption from the interval required by 10CFR 50, Appendix J. Inboard Drywell Air Sample valve 2-8501-5 A was immediately closed and challenged with a local leak rate test, which yielded a leakage rate of 2.20 scfh. Valve 2-8501-5B was declared inoperable, and valve 2-8501-5A was taken Out-of-Service in the closed msition. Additional testing perfonned on December 11,1992 during a Unit 2 maintenance outage ic entified valves 2-4799-530,2-2599-23B, and 2-2599-24B to be leaking an undetermined amount. As a result of this leakage, the Technical Specification limit for primary containment leakage,0.6La, was exceeded.
Unit Conditions at Time of Reportable Event:
Power Level: 94%
Operating Mode: Unknown Reportability Reason (s) for This LER:
10 CFR 50.73(a)(2)(i)- Shutdowns or Technical Specific rion Violations. [10]
Primary Cause(s) for This LER:
Equipment Failure (20 ]
Emergency Classification (s) for This LER:
There is no Emergency Classification for This LER I of 3 07/02/97 09:57:46
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LER licader Listing for 23792031 http://scss.orn l.go v/sc ripts/scss/resu lts/rL E RDetl l.c fm71ernm bm23792031 Referenced LERs:
LER Event Number.
Reactor UnitName Date LER Title
)
In Perable Control Rod Drive (CRD) E-9 Found 23788004 Dresden Unit #2 02/21/88 Electncally Armed Due to Personnel Error Leak Rate Limits Exceeded in Drywell Head Seal 23788018 Dresden Unit #2 10/30/88 and MSIV 2-203-1D Tests Due to Misalignment and Seat Wear i
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Violation of Core Thermal Power Limits Due to 23791007 Dresden Unit #2 04/11/91 Unplanned 2B Reactor Recirculation Pump Speed
- Increase Contact Mike Poore at ORNL with questions or comments concerning SCSS Web site content.
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Copyright O 1996 Oak Ridge National Laboratory (ORNL)
Last modified: January 31,1997 4
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http /scss.ornl. gov / scripts /scss/results/rLERDetil.cfm?lernmbr=23793002 LER Header Listing for 23793002 1
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mm LER Header Listingfor LER Number: 23793002 Docket Number Year Report #
Rev#
DCS Number Event Date 237 1993 002 2
9411290270 01/21/93 Nuclear Plant: Dresden Unit #2 NRC Region:
3 Facility Operator: Commonwealth Edison Co.
Architect Engineer: Sargent & Lundy NSSS Vendor: General Electric Reactor Type: Boiling Water Reactor Comm. Operations Date: 06/09/70 finKesmedl FiMiiM3 EhiiiiiiiEl ABSTRACT OWER LEVEL - 000%. On January 21,1993 with Unit 2 in a refuel outage, the performance of Dresden Technical Surveillance (DTS) 1600-01, Local Leak Rate Testing Of Primary Containment Isolation Valves, identified the Head Cooling Inlet Isolation Vajve 2-205-2-4 to be leaking an undetermined amount. This exceeded the maximum pathway leakage rate for Type B and C primary containment leakage,488.452 scfh (0.6L sub a), as listed in Technical Specification 3.7.A.2.b.(2)(a).
Once the leakage rate was recorded, the valve was again verified to be in the fully closed position. The measured leakage rate dropped to 3.0 scfh upon increasing the seating force. The valve operator was repaired under Work Request D10353, which reduced the leakage to 2.84 scih. The safety significance of the leakage past valve 2-205-2-4 has been considered to be minimal since the redundant Head Cooling Isolation Valve 2-205-27 leaked 3.31 scfh; therefore, the total through leakage out of the penetration, on a minimum pathway basis, was 3.31 scfh. The total as-found minimum pathway leakage (Type A test) was 2.3718 wt%/ day which exceeded the Technical Specification 3.7.A.2 limit of 1.2 wt%/ day. Calculations have been performed to prove this leakage did not exceed 10 CFR Part 100 j
limits. \\\\\\\\\\\\
Unit Conditions at Time of Reportable Event:
Power Level: 0%
Operating Mode: Unknown Reportability Reason (s) for This LER:
10 CFR 50.73(a)(2)(i) - Shutdowns or Technical Specification Violations. [10]
Primary Cause(s) for This LER:
Equipment Failure [20 )
Emergency Classification (s) for This LER:
There is no Emergency Classification for This LER 1 of 2 07/02/97 09:58:25
http /scss.ornl. gov / scripts /scss/results'rLERDetil.cfm71ernmbr=23793002 1
LER lleader Listing for 23793002 Referenced LERs:
i LER Event i
Number Reactor Unit Name Date LER Title
- Type B and C Primary Containment Local Leak 23790009 Dresden Unit #2 09/23/90 Rate Test Requirements Exceeded Due to Leaking Isolation Valves 2
Leakage Path Discovered During Primary 1
23790018 Dresden Unit #2 12/18/90 Containment ILRT Due to Management Deficiency Contact Mike Poore at ORNL with uestions or comments concerning SCSS Web site content.
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J Copyright C 1996 Gak Ridge National Laboratory (ORNL) i I,ast modified: January 31,1997 1
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LER llGader Listing for 23795018 http1/scss.ornl. gov / scripts /scss/resultvrLERDetll.cfm71ernmbr=23 i95018 C%
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(h, LEB Header Listingfor LER Number: 23795018 Docket Number Year Report #
Rev#
DCS Number Event Date 237 1995 018 1
9605200165 06/10/95 Nuclear Plant: Dresden Unit #2 NRC Region:
3 Facility Operator: Commonwealth Edison Co.
Architect Engineer: Sargent & Lundy NSSS Vendor: General Electric Reactor Type: Boiling Water Reactor Comm. Operations Date: 06/09/70 inENiEHiirr1 MiiiiiiWilisEl FMsMiEl ABSTRACT POWER LEVEL - 000%. At approximately 0830, on June 10,1995, with Unit 2 shutdown for Refuel Outage D2R14, the performance of a Dresden Technical Surveillance identified the High Pressure Coolant injection (HPCI) System Turbine Exhaust to Suppression Pool Check Valve 2-2301-45 to be leaking more than the test equipinent could measure. When the valve's leakage was added to the existing maximum pathway leakage rate, the maximum pathway leakage rate Technical Specification limit for Type B and C primary containment leakage was exceeded. The safety significance of the leakage past the 2-2301-45 was considered to be minimal since the additional leakage out of containment, on a minimum pathway basis, was 0 scfh from the inboard isolation Stop Check Valve 2-2301-74 and would not cause the maximum off-site dose rates established in 10 CFR 100 to be exceeded. The check valve was removed, inspected, replaced and Local Leak Rate Tested prior to unit startup. This supplement contains the root cause, corrective actions taken and results of an NPRDS search for all test volumes that exceeded administrative local leakage rate limits during Refuel Outage D2R14.
Unit Conditions at Time of Reportable Event:
Power Level: 0%
Operating Mode: Not Applicable Reportability Reason (s) for This LER:
10 CFR 50.73(a)(2)(i) - Shutdowns or Technical Specification Violations. [10]
10 CFR 50.73(a)(2)(ii)- Unanalyzed Conditions. [11]
Primary Cause(s) for This LER:
Equipment Failure [20 ]
Emergency Classification (s) for This LER:
There is no Emergency Classification for This LER Iof2 07/02/97 09:59:17
. LER lleader Listing for 23795018 http://scss.ornl. gov / scripts /scss/resul ts/rL ERDetl l.c fm71ernm br=23795018 f
a Referenced LERs:
LER Event Number Reactor Unit Name Date LER Title Type B and C Leakage Limit Exceeded due to 23794022 Dresden Unit #2 08/08/94 Worn Seating Surface of HPCI Check valve Local Leak Rate Testing "As Found" Limit l
24989009 Dresden Unit #3 12/07/89 Exceeded Due to Leakage From Primary Containment Valves Type B and C Primary Containment Local Leak 24991007 Dresden Unit #3 09/10/91 Rate Testing Limit Exceeded due to HPCI Turbine Exhaust Check Valve Leakage Type B and C Leakage Limit Exceeded Due to 24995011 Dresden Unit #3 05/29/95: Ineffective Corrective Actions for Past Valve Failures Contact Mike Poore at ORNL with questions or comments concerning SCSS Web site content.
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Copyright C 1996 Oak Ridge National Laboratory (ORNL)
Last modified: January 31,1997 j
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LER Header Listing for 24589021 httpd/sc ss.oml. gov / scripts /scss/results/rL E RDetl l.c fm 71erntabr=24589021
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Rev #
DCS Number Event Date 245 1989 021 0
8911220182 10/19/89 Nuclear Plant: Millstone Unit #1 NRC Region:
1 Facility Operator: Northeast Nuclear Energy Co.
Architect Engineer: Ebasco Services, Inc.
NSSS Vendor: General Electric Reactor Type: Boiling Water Reactor Comm. Operations Date: 03/01/71
[aliiisammesi W.Waiisii9 FXiiiiviii 5"l ABSTRACT POWER LEVEL - 070%. ON 10/19/89 AT 1515 IIOURS WITH THE PLANT AT 70% POWER (530F AND 1000 PSIG) A FULL REACTOR SCRAM OCCURRED AS A RESULT OF A MAIN TURBINE TRIP (TURBINE STOP VALVES GREATER THAN 10% CLOSURE). THE MAIN TURBINE TRIP WAS THE RESULT OF A HIGIl REACTOR WATER LEVEL TURBINE TRIP SIGNAL (+48'). THE lilGH REACTOR WATER CONDITION WAS Tile RESULT OF 'N FEEDWATER REGULATING VALVE BECOMING STUCK IN TIIE OPEN POSITION WHICH OCCURRED WHILE RETURNING 'B' FEEDWATER REGULATING VALVE TO SERVICE.
DURING Tile LEVEL DECREASE WlIICH FOLLOWED THE SCRAM, STANDBY GAS TREATMENT SYSTEM INITIATED AS EXPECTED. ALL SYSTEMS FUNCTIONED AS EXPECTED. NO SAFETY CONSEQUENCES RESULTED FROM THIS EVENT.
Unit Conditions at Time of Reportable Event:
Power Level: 70%
Operating Mode: Unknown Reportability Reason (s) for This LER:
10 CFR 50.73(a)(2)(iv)- ESF Actuations. [13]
10 CFR 50.73(a)(2)(v)- Event That Could Have Prevented Fulfillment of a Safety Function. [14]
Primary Cause(s) for This LER:
Design Error or inadequacy {34 )
Emergency Classification (s) for This LER:
There is no Emergency Classification for This LER Referenced LERs:
1of2 07/02/97 09:59:51
LERlleader Listing for 24589021 http://scss.ornl. gov / scripts /scss/results/rLERDetll.cfm71ernmbr=24589021 There are no Referenced LERs for This LER Contact Miktfmts at ORNL with questions or comments concerning SCSS Web site content. '
Contact Dale Yeilding at NRC with questions concerning SCS5 Web site access.
Copyright C 1996 Gak Ridge National Laboratory (ORNL) -
Last modified: January 31,1997 l
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LER l{eader Listing for 24593009 http://scss.om i. gov / scripts /scss/resu lts/rL E RDetl l.c fm?lernmbr=24593009 pntrMMrki$fstzar.ar7ttugitrainsM "
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LER Header Listingfor LER Number: 24593009 Docket Number Year Report #
Rev#
DCS Number Event Date 245 1993 009 0
9308130062 07/06/93 Nuclear Plant: Millstone Unit #1 NRC Region:
1 Facility Ooerator: Northeast Nuclear Energy Co.
Architect Engineer: Ebasco Services, Inc.
NSSS Vendor: General Electric Reactor Type: Boiling Water Reactor Comm. Operations Date: 03/01/71 riiiiiisyliisil Feiiiiiiiiiii=51 PWiiniiiW1 ABSTRACT POWER LEVEL - 100%. On July 6,1993, at 1005 hours0.0116 days <br />0.279 hours <br />0.00166 weeks <br />3.824025e-4 months <br />, with the plant operating at 100% power (530 degrees F and 1030 psig), during performance of a routine surveillance test of the plant's Containment isolation Valves, the outboard Low Pressure Coolant Injection (LPCI) to Recirculation Loop Valve 1-LP-9B failed to open electrically after being manually closed. During the performance of the surveillance, the motor operator's declutch fork jammed the motor clutch in mid-position, precluding manual or electrical operation of the valve. This resulted in the inoperability of the LPCI system.
Repairs were initiated and the declutch fork and clutch were repositioned to permit both manual and electrical operation of the rnotor operator. At 1415 hours0.0164 days <br />0.393 hours <br />0.00234 weeks <br />5.384075e-4 months <br /> repairs were completed and the valve was opened electrically. The LPCI system was then returned to service. All other safety systems were operable at the time of the event and there were no consequences.
Unit Conditions at Time of Reportable Event:
Power Level: 100%
Operating Mode: Unknown Reportability Reason (s) for This LER:
10 CFR 50.73(a)(2)(v)- Event Hat Could Have Prevented Fulfillment of a Safety Function. [14]
Primary Cause(s) for This LER:
Equipment Failure [20 ]
Emergency Classification (s) for This LER:
There is no Emergency Classification for This LER Referenced LERs:
There are no Referenced LERs for This LER I of 2 07/02/97 10:00:33
1.ER Header Listing for 24593009
' http://scss.ornl. gov / scripts /scss/results/rLERDetil.cfm?lemmbro24593009
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Contact Daktleilding at NRC with questions concerning SCSS Web site access.
Copyright 01996 Oak Ridge National Laboratory (Of'.NL)
Last modified: January 31,1997 l
2 OII 07/02/97 10:00:33-
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http /scss.ornl. gov / scripts /scss/results/rLERDetll.cfm?lernmbr=24596012 1
LER Ifeader Listing for 24596012 N
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LER Header Listinefor LER Number: 24596012 Docket Number Year Report #
Rev #
DCS Number Event Date 245 1996 012 0
9603200039 12/03/95 Nuclear Plant: Millstone Unit #1 NRC Region:
1 Facility Operator: Northeast Nuclear Energy Co.
Architect Engineer: Ebasco Services, Inc.
NSSS Vendor: General Electric Reactor Type: Boiling Water Reactor Comm. Operations Date: 03/01/71 MallHiMEl MilinWEi2 EDKHMiEE ABSTRACT POWER LEVEL - 000%. On December 3,1995, with the plant shutdown and the reactor in the COLD SilUTDOWN condition, it was determined that a reactor water cleanup system containment isolation check valve,1-CU-29, had exceeded its maximum leak rate while it was m operation. The valve was replaced during refueling outage 15 (RF015) because it could not be tested pursuant to 10CFR50 Appendix J criteria. Subsequent to its removal, a local leak rate test (LLRT) was performed on the valve, and it was determined to exceed the maximum allowable leak rate. Since the valve was leak rate tested after its replacement, this event was not immediately identified as reportable. The event was determined to be reportable on February 7,1996. Excessive leakage of a containment isolation valve is contrary to the requirements of Millstone Unit No.1 Technical Specification 4.7.A.3.e.(1)(a).
Therefore, this event is reportable pursuant to 10CFR50.73(a)(2)(i) as a condition prohibited by the plant's Technical Specifications. There were no adverse safety consequences as a result of this event, since the redundant valve in the penetration was capable of performing the containment isolation function.
Unit Conditions at Time of Reportable Event:
Power Level: 0%
Operating Mode: Not Applicable Reportability Reason (s) for This LER:
10 CFR 50.73(a)(2)(i)- Shutdowns or Technical Specification Violations. [10]
Primary Cause(s) for This LER:
Design Error or inadequacy [34 ]
Emergency Classification (s) for This LER:
There is no Emergency Classification for This LER
'iof2 07/02/97 10:01:10
http /scss.ornt. gov / scripts /scss/results/rLERDetll.cfm71ernmbr=24596012
- LER Header Listing for 24596012 1
Referenced LERs:
There are no Referenced LERs for This LER 4
4 J
Contact Mike Poore at ORNL with questions or comments concerning SCSS Web site content.
j Contact Dale Yellding at NRC with questions concerning SCSS Web site access.
l" Copyright C 1996 Oak Ridge National Laboratory (ORNL)
Last modified: January 31,1997 i
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LER lisader Listing for 24786036 http://scss.ornl. gov / scripts /scss/results/rL E RDetl l.c fm?lernmbr=24786036 O\\
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LER Header Listingfor LER Number: 24786036 Docket Number Year Report #
Rev #
DCS Number Event Date 247 1986 036 0
8612040092 10/23/86 Nuclear Plant: Indian Point Unit #2 NRC Region:
1 Facility Operator: Consolidated Edison Co.
Architect Engineer: United Engineers & Construction NSSS Vendor: Westinghouse Reactor Type: Pressurized Water Reactor Comm. Operations Date: 08/01/74 M MiiiMaiisM IEiiitTalEd ABSTRACT POWER LEVEL - 038%. AT 1330 ON 10-23-86, A MANUAL REACTOR TRIP WAS INITIATED FROM A POWER LEVEL OF 38% FOLLOWING AN AUTOMATIC TRIP OF THE #21 MAIN BOII ER FEED PUMP ON HIGli DISCHARGE PRESSURE. TIIE HIGH DISCHARGE PRESSURE WAS A RESULT OF THE CLOSURE OF A FLOW PATH BACK THROUGH IDLE FEED PUMP
- 22. SUBSEQUENT EXCESS FEEDWATER FLOW CAUSED BY THE CONTINUED OPERATION OF FEED PUMP #21. THE AUXILIARY FEEDWATER SYSTEM WAS ACTUATED FOLLOWING Tile LOSS OF THE MAIN BOILER FEED PUMP. THE 2 MOTOR DRIVEN AUXILIARY FEEDWATER PUMPS STARTED AS REQUIRED AND SUPPLIED SUFFICIENT FLOW, HOWEVER TIIE TURBINE DRIVEN AUXILIARY FEED PUMP TRIPPED AFTER RECEIVING ITS START SIGNAL. TURBINE DRIVEN AUXILIARY FEED PUMP FLOW WAS NOT REQUIRED, SINCE Tile MOTOR DRIVEN AUXILIARY FEED PUMPS PROVIDED ADEQUATE FLOW. SUBSEQUENT INVESTIGATION REVEALED PROBLEMS WITH THE MAIN FEEDWATER PUMP DISCHARGE VALVES AND THE STEAM CONTROL VALVE TO TIIE TURBINE DRIVEN AUXILIARY FEED PUMP. THESE PROBLEMS WERE CORRECTED PRIOR TO RETURNING TO POWER.
Unit Conditions at Time of Reportable Event:
Power Level: 38%
Operating Mode: Unknown Reportability Reason (s) for This LER:
10 CFR 50.73(a)(2)(iv)- ESF Actuations. [13]
Primary Cause(s) for This LER:
There is no Primary Cause for This LER Emergency Classification (s) for This LER:
1of2 07/02/97 10:01:52
http /scss.ornl. gov / scripts /scss/results/rLERDetil.cfm?lernmbr=24786036
/
- LER licader Listing for 24786036 There is no Emergency Classification for This LER Referenced LERs:
There are no Referenced LERs for This LER Contact Mike Poore at ORNL with questions or comments concerning SCSS Web site content.
Contact Dale Yellding at NRC with questions concerning SCSS Web site access.
Copyright O 1996 Oak Ridge National Laboratory (ORNL)
- Last modified: January 31,1997 0
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http /scss.ornl. gov / scripts /scss/results/rLERDetti.cfm?lernmbr=24788006 LER Header Listing for 24788006 1
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LER Header Listinefor LER Number: 24788006 Docket Number Year Report #
Rev#
DCS Number Event Date 247 1988 006 0
8807250371 06/17/88 Nuclear Plant: Indian Point Unit #2 NRC Region:
1 Facility Operator: Consolidated Edison Co.
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Architect Engineer: Unit.ed Engineers & Construction NSSS Vendor: Westinghouse Reactor Type: Pressurized Water Reactor Comm. Operations Date: 08/01/74 E M E m L.Ymm ABSTRACT i
POWER LEVEL - 100%. ON JUNE 17,1988, WHILE Tile PLANT WAS AT 100% POWER, AN UNPLANNED REACTOR TRIP OCCURRED, A CONTRACT EMPLOYEE PERFORMING CLEANING FUNCTIONS ON Tile CONVENTIONAL SIDE OF THE PLANT INADVERTENTLY DEPRESSED TIIE MANUAL TRIP BUTTON FOR 21 MAIN BOILER FEEDWATER PUMP.
DURING Tile SUBSEQUENT ATTEMPTED RECOVERY FROM THE LOSS OF ONE FEEDWATER PUMP Tile MAIN GENERATOR TRIPPED ON HIGH WATER LEVEL IN 22 STREAM GENERATOR. Tile GENERATOR TRIP CAUSED A MAIN TURBINE, AND SUBSEQUENT REACTOR TRIP. THE REACTOR PROTECTION SYSTEM FUNCTIONED NORMALLY AND TIIE PUBLIC HEALTH AND SAFETY WERE NOT AFFECTED. THE USE OF CONTRACT EMPLOYEES FOR CLEANING IN Tile VICINITY OF CRITICAL OPERATING EQUIPMENT AT THE STATION IS BEING REVIEWED.
Unit Conditions at Time of Reportable Event:
Power Level: 100%
Operating Mode: Unknown Reportability Reason (s) for This LER:
10 CFR 50.73(a)(2)(iv)- ESF Actuations. [13]
Primary Cause(s) for This LER:
Accidental Action (31 ]
Emergency Classification (s) for This LER:
There is no Emergency Classification for This LER Referenced LERs:
I of 2 07/02/97 10:04:09
LER Header Listing for 24788006 http://scis.ornl. gov / scripts /scss'results/rLERDett i.c fm71ernmbr=24788006
-LER Event Number Reactor Unit Name Date LER Title Manual Reactor Trip Due to Loss of Main Boiler 24786036' Indian Point Unit #2 10/23/86 Feed Ihunp Contact Mike Poore at ORNL with questions or comments concerning SCSS Web site content.
i Contact Dale Yeilding at NRC with questions concerning SCSS Web site acces2.
j Copyright O 1996 Oak Ridge National Laboratory (ORNL)
Last anodified: January 31,1997 i
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i 2 of 2 07/02/97 10:04:09
http /scss.ornl. gov / scripts /scss/results/rLERDetil.cfm71ernmbr=24989009 j
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LER licader Listing for 24989009 7's hggggggg.g.ggigayggagem r'g
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Rev #
DCS Number Event Date 249 1989 009 1
9209240032 12/07/89
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Nuclear Plant: Dresden Unit #3 NRC Region:
3 Facility Operator: Commonwealth Edison Co.
Architect Engineer: Sargent & Lundy NSSS Vendor: General Electric Reactor Type: Boiling Water Reactor Comm. Operations Date: 11/16/71 M M TJ.3 M ABSTRACT POWER LEVEL - 000%. On December 7,1989, with Unit 3 in a refueling outage and during the performance of Dresden Technical Staff Surveillance Procedure (DTS) 1600-1, Local Leak Rate Testing (LLRT) of Primary Containment Isolation Valves, feedwater check valve 3-0220-58A exhibited a !cakage of 1062.82 Standard Cubic Feet per Hour (SCFII). This leak rate, along with unsatisfactory leak rates from Reactor Building to pressure suppression chamber vacuum breaker check valve 3-1601-31B and High Pressure Coolant Injection steam exhaust to pressure suppression chamber check valve 3-2301-45, combined to bring the total as-found leakage, using the maximum pathway method for type 'B' and 'C' testing, to 1510.4526 SCFH. This leakage exceeded the Technical Specification 3.7.A.2.b.(2)(a) limit of 488.452 SCFH. The safety significance of this event was minimal because in line valves were not observed to be leaking. Therefore, the 'through' leakage, which represents actual containment leakage, was minimal. A previous occurrence of this type is outlined in LER-88-27 on Docket 050-249.
Unit Conditions at Time of Reportable Event:
Power Level: 0%
Operating Mode: Unknown Reportability Reason (s) for This LER:
10 CFR 50.73(a)(2)(i)- Shutdowns or Technical Specification Violations. [10]
Primary Cause(s) for This LER:
Equipment Failure [20 ]
Iluman Error [35 ]
- Emergency Classification (s) for This LER:
There is no Emergency Classification for This LER Iof2 07/02/97 10:05:05
http /scss.ornl. gov / scripts /scss/results/rLERDetil.cfm71ernmbr=24989009 1
LER ileader Listing for 24989009 Referenced LERs:
LER Event Number Reactor Unit Name Date LER Title High Pressure Coolant Injection System Declared 24989004 Dresden Unit #3 10/22/89 Inoperable Due to Failed Room Cooler Fan Drive Belts j
Contact Mike Poore at ORNL with uestions or comments concerning SCSS Web site content.
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Copyright O 1996 Oak Ridge National Laboratory (ORNL)
Last modified: January 31,1997
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l 2 of 2 07/02/97 10:05:05
LERlleader Listing for 24991007 http://scss.ornl. gov / scripts /scss/results/rLERDet!1 cfm?lernmbr=24991007 bygm. m
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Rev #
DCS Number Event Date 249 1991 007 1
9411010307 09/10/91 Nuclear Plant: Dresden Unit #3 NRC Region:
3 Facility Operator: Commonwealth Edison Co.
Architect Engineer: Sargent & Lundy NSSS Vendor: General Electric Reactor Type: Boiling Water Reactor Comm. Operations Date: 11/16/71 M MiiERiiBEl #1IU3 ABSTRACT POWER LEVEL - 000%. On September 10,1991 with Unit 3 in a refueling outage, while performing Dresden Technical Surveillance (DTS) 1600-01, Local Leak Rate Testing of Primary Containment Isolation Valve B, the leakage between the 3-2301-74, High Pressure Coolant Injection (HPCI) [BJ]
Turbine Exhaust To Suppression Chamber Stop Check Valve, and the 3-2301-45, HPCI Turbine Exhaust Check Valve, was unable to be determined. Further diagnosis and previous LLRT history indicated that the 32301-45 was the leaking valve. This valve leakage exceeded the Technical Specification 3.7.A.2.b(2)(a) limit of 488.452 scfh. Inspection of the check valve revealed a torn seat.
The valve was replaced. The valve seat failed as a result of excessive valve cycling. The operating procedures for the turbine have been modified to prevent low turbine exhaust pressure operations which will in turn prevent cycling of check valve 3-2301-45. The safety significance of this event was tr.inimal because the redundant in-line isolation valve leaked 84 scfh. The total as-found minimum pathway leakage (Type A test) was 1.4595 wt%/ day which exceeded the Technical Specification limit of 1.2 wt%/ day. Calculations have been performed to prove this leakage did not exceed 10 CFR 100 limits.
Unit Conditions at Time of Reportable Event:
Power Level: 0%
Operating Mode: Unknown Reportability Reason (s) for This LER:
10 CFR 50.73(a)(2)(i) - Shutdowns or Technical Specification Violations. [10]
Primary Cause(s) for This LER:
Equipment Failure [20 )
Emergency Classification (s) for This LER:
There is no Emergency Classification for This LER Iof2 07/02/97 10:07:19
http /scss.ornl. gov / scripts /scss/results'rLERDetll.cfm71ernmbr=24991007 1
LER Ileader Listing for 24991007 1
Referenced LERs:
LER Event Number Reactor UnitName Date LER Title
- High Pressure Coolant injection System Declared 24989004-Dresden Unit #3 10/22/89 Inoperable Due to Failed Room Cooler Fan Drive l
Belts Local Leak Rate Testing "As Found" Limit i
24989009 Dresden Unit #3 12/07/89 Exceeded Due to Leakage From Primary Containment Valves Contact Mike Poore at ORNL with questions or comments concerning SCSS Web site content.
Contact Daltlellding at NRC with questions concerning SCSS Web site access.
Copyright C 199' Oak Ridge National Laboratory (ORNL)
Last modified: January 31,1997 2 of 2 07/02/97 10:07:19
LER licader Listing for 24993006 htt p 1/scss.ornl.go v/ scripts /scss/results/rL E RDetl l.c fm71emm br=24993006 O\\
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LER Header Listingfor LER Number: 24993006 Docket Number Year Report #
Rev#
DCS Number Event Date 249 1993 006 1
9310060159 03/13/93 Nuclear Plant: Dresden Unit #3 NRC Region:
3 Facility Operator: Conunonwealth Edison Co.
Architect Eagineer: Sargent & Lundy NSSS Vendor: General Electric Reactor Type: Boiling Water Reactor Comm. Operations Date: 11/16/71 liliRMBIEEl BliiiBliiiEd EliniiG41 ABSTRACT POWER LEVEL - 000%. With Unit 3 in a forced maintenance outage, the performance of Dresden Technical Surveillance (DTS) 1600-01, Local Leak Rate Testing Of Primary Containment Isolation Valves, identified the inboard 'A' Feedwater Line Check Valve 3-220-58A and the Traversing Incore Probe (TIP) Purge Check Valve 3-4799-514 to be leaking 196.67 scfh and 45 scfh respectively. These values when added to the existing maximum pathway leakage rate for Type B and C primary containment leakage,488.452 scfh (0.6L sub a), as listed in Technical Specification 3.7.A.2.b.(2)(a).
The Inboard 'A' Feedwater Line Check Valve 3-220-58A was disassembled and an inspection of the valve internals revealed a worn hinge pin / bushing. The seat, disk and hinge pin / bushing were replaced under Work Request (WR) 16938. The TIP Purge Check Valve 3-4799-514 was not inspected but the most likely cause for failure was debris on the seat. This check valve was replaced under WR 17211.
The safety significance of the leakage has been considered to be minimal, since the additional leakage out of containment, on a minimum pathway basis, was 48.65 scfh and would not cause the maximum off site dose rates established in 10 CFR 100 to be exceeded. Final as-left local leak rate tests (3-220-58A = 0.1 scfh,3-4799-514 = 1.8 scfh) were performed in accordance with DTS 1600-01 to verify the valves' seating integrity prior to placing them back into service.
L:\\8360\\8301\\249\\l80\\93\\006.R01 Unit Conditions at Time of Reportable Event:
Power Level: 0%
Operating Mode: Unknown Reportability Reason (s) for This LER:
10 CFR 50.73(a)(2)(i)- Shutdowns or Technical Specification Violations. [10]
Primary Cause(s) for This LER:
Equipment Failure [20 ]
Emergency Classification (s) for This LER:
1of2 07/02/97 10:07:52
LER lleader Listing for 24993006 httpd/scss.ornl. gov / scripts /scss/results/rLERDetll.c fm?lernmbr=24993006 There is no Emergency Classification for This LER l
Referenced LERs:
LER Event Number Reactor UnitName Date LER Title Type B and C Primary Containment Local Leak 23790009 Dresden Unit #2 09/23/90 Rate Test Requirements Exceeded Due to Leaking i
Isolation Valves Local Leak Rate Testing "As Found" Limit 24989009 Dresden Unit #3 12/07/89 Exceeded Due to Leakage From Primary Containment Valves Contact Mike Poore at ORNL with questions or comments concerning SCSS Web site content.
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Copyright C 1996 Oak Ridge National Laboratory (ORNL)
Last modified: January 31,1997 1
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LER lleader Listing for 24994009 http1/sc ss.ornl. gov / scripts /scss/results/r L ERDetl l.c fm7 lern m br=24994009 (N.
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LER Header Listingfor LER Number: 24994009 Docket Number Year Report #
Rev#
DCS Number Event Date 249 1994 009 1
9507270184 03/12/94 Nuclear Plant: Dresden Unit #3 NRC Region:
3 Facility Operator: Commonwealth Edison Co.
Architect Engineer: Sargent & Lundy NSSS Vendor: General Electric Reactor Type: Boiling Water Reactor Comm. Operations Date: 11/16/71 filiiliti976W1 PEsifiaidW IB99liMPMI ABSTRACT l
POWER LEVEL - 000%. At approximately 1800, on March 12,1994, with Unit 3 in Refuel Outage D3R13, the performance of Dresden Technical Surveillance, Local Leak Rate Testing Of Primary Containment isolation Valves, identified the Isolation Condenser System Condensate Return Throttling Valve 3-1301-3 to be leaking an undetermined amount. This value when added to the existing maximum pathway leakage rate resulted in the maximum pathway leakage rate limit for Type B and C primary containment leakage being exceeded. Trouble-shooting determined that the Motor Operated i
Valve had not closed completely when stroked for draining the system. After the valve was closed using primary containment isolation logic another LLRT was performed and leakage was determined to be 5.3 scfh. The safety significance of not fully closing the 3-1301-3 valve for the LLRT is considered minimal since when the valve was fully closed with primary containment isolation logic an acceptable LLRT was obtained. This supplement contains the root cause, corrective actions taken and results of an NPRDS search for all test volumes that exceeded administrative local leakage limits during Refuel Outage D3R13.
Unit Conditions at Time of Reportable Event:
Power Level: 0%
Operating Mode: Not Applicable Reportability Reason (s) for This LER:
10 CFR 50.73(a)(2)(i)- Shutdowns or Technical Specification Violations. [10]
Primary Cause(s) for This LER:
Equipment Failure [20 ]
Design Error or inadequacy [34 ]
Iluman Error [35 ]
Emergency Classification (s) for This LER:
'Iof2 07/02/97 10:08:35
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http /scss.ornl. gov / scripts /scss/results/rLERDetll.cfm71ernmbr=24994009 LER ll:ader Listing for 24994009 1
There is no Emergency Classification for This LER Referenced LERs:
LER Event Number Reactor UnitName Date LER Title Type B and C Primary Containment Local Leak 23793002 Dresden Unit #2 01/21/93 Rate Testing Limit Exceeded Due To Leakage
- Past IIcad Cooling !nlet Isolation Valve 2-205-2-4 Type B and C Primary Containment Local Leak 24991002 Dresden Unit #3 09/10/91 Rate Testing Limit Exceeded due to HPCI Turbine Exhaust Check Valve Leakage Leakage Limit Exceeded Due to Valve Internal 24995007 Dresden Unit #3 06/30/95 Damage Caused by Manual Operation of MOV Type B and C Leakage Limit Exceeded Due to 24995011 Dresden Unit #3 05/29/95 Ineffective Corrective Actions for Past Valve Failures Contact Mike Poore at ORNL with uestions or comments concerning SCSS Web site content.
Contact Dalcltilding at NRC with quest)ons concerning SCSS Web site access.
Copyright C 1996 Oak Ridge National Laboratory (ORNL) lest modified: January 31,1997 1
2 of 2 07/02/97 10:08:36
LER lleader Listing for 24995003 h tt p 1/scss.ornl. gov /scriptvsc ss/res ults/rL E RDet t i.c fm?lernmbr=24995003 dEggg!gfgE.9filialAFM
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LER Header Listinefor LER Number: 24995003 Docket Number Year Report #
Rev #
DCS Number Event Date 249 1995 003 0
9503090147 02/03/95 Nuclear Plant: Dresden Unit #3 NRC Region:
3 Facility Operator: Commonwealth Edison Co.
Architect Engineer: Sargent & Lundy NSSS Vendor: General Electric Reactor Type: Boiling Water Reactor Comm. Operations Date: 11/16/71 NsiiiiMilillirl Fa6iiialHiiiEM FNW1 ABSTRACT POWER LEVEL - 099%. At approximately 1820, on February 3,1995, with unit 3 operating at 99%
power, the performance of Dresden Technical Surveillance (DTS) 1600-01, Local Leak Rate Testing Of Primary Containment Isolation Valves, identified the Torus to Reactor Building Vacuum Breaker
[BF] Check Valve 3-1601-31B to be leaking an undetermined amount. This leakage rate resulted in the pnmary containment Type B and C maximum pathway leakage rate limit being exceeded. Building Vacuum Breaker Check Valve 3-1601-31 A was also found to be leaking an undetermined amount. A Technical Specification required nuclear plant shutdown was started at 1905. Upon completion of torquing the hinge pin flange bolting, leak rate tests yielded leakage rates of 5.97 scfh and 2.03 scfh for the A and B vacuum breakers respectively and the nuclear plant shutdown was halted at 2000 on February 3,1995. Calculations show that the leakage past valves 3-1601-31 A and 3-1601-31B would not cause the maximum off-site dose rates established in 10 CFR 100 m be exceeded. The event was caused by an inadequate surveillance procedure which was performed a month earlier on January 6, 1995. The procedure will be revised.
Unit Conditions at Time of Reportable Event:
Power Level: 99%
Operating Mode: Unknown Reportability Reason (s) for This LER:
10 CFR 50.73(a)(2)(ii)- Unanalyzed Conditions. [11]
Primary Cause(s) for This LER:
- Inadequate Training [36 )
Procedural Deficiency [40 ]
Emergency Classification (s) for This LER:
There is no Emergency Classification for This LER 1of2 07/02/97 10:09:10
l l
http://scss.orni. gov / scripts /sc ss/results/rL E RDeti l.c fm?le rnmbr=24995003 l. LERlicader Listing for 24995003 Referenced LERs:
LER Event Number Reactor UnitName Date LER Title Leakage Path Discovered During Primary I
23790018 Dresden Unit #2 12/18/90 Containment ILRT Due to Management Deficiency Local Leak Rate Testing "As Found" Limit 24989002 Dresden Unit #3 12/07/89 Exceeded Due to Leakage From Primary
. Containment Valves Type B and C Primary Containment Local Leak Rate Testing Limit of 0.6L sub a Exceeded Due to 24993016 Dresden Unit #3 09/29/93 Leakage Past Atmospheric Containment Atmosphere Dilution (ACAD) System Check Valve 3-2599-23A.
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Copyright O 1996 Oak Ridge National Laboratory (ORNL)
Last modified: January 31,1997 I
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2 of 2 07/02/97 10:09:10 l
l LER IIcader Listing for 25096002 http://scss.orn t gov /sc ripts/scss/results/r L E RDetl l.c fm71ernmbr=25096002 Eff1!T&tD4%4(lt@.stt(14:nYtttiMEM4MM
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LER Header Listinefor LER Number: 25096002 Docket Number Year Report #
Rev #
DCS Number Event Date 250 1996 002 0
9603120224 02/09/96 Nuclear Plant: Turkey Point Unit #3 NRC Region:
2 Facility Operator: Florida Power & Light Co.
Architect Engineer: Bechtel Power Corp.
NSSS Vendor: Westinghouse Reactor Type: Pressurized Water Reactor Comm. Operations Date: 12/14/72 fiillibililWil51 IAliiiiiiisiiiiliG'd MiliiiiiiiRW1 ABSTRACT POWER LEVEL - 060%. On February 09,1996, Florida Power & Light Company's Turkey Point Unit 3 was operating in mode 1 at 60% power to support condenser waterbox cleanmg. At 2329 the 'B' Steam Generator Feed Pump (SGFP) was stopped to monitor its discharge check valve closing stroke.
The discharge check valve did not stroke closed as expected. At 2334 the resulting feed flow transient caused the 'C' Steam Generator (S/G) level to increase, resulting in a turbine trip. A reactor trip by turbine trip occurred immediately thereafter. The cause of the turbine trip / reactor trip was cognitive personnel error. The operator failed to effectively control the 'C' S/G level during the feed flow transient. The NRC operations center was notified at 0035 on February 10,1996, in accordance with 10 CFR 50.72 (b)(2)(ii), Reactor Protection System Actuation.
Unit Conditions at Time of Reportable Event:
Power Level: 60%
Operating Mode: Not Applicable Reportability Reason (s) for This LER:
10 CFR 50.73(a)(2)(iv)- ESF Actuations. [13]
Primary Cause(s) for This LER:
lluman Error [35 ]
Emergency Classification (s) for This LER:
There is no Emergency Classification for This LER Referenced LERs:
There are no Referenced LERs for This LER lof2 07/02/97 10:09:49
LER Header Listing for 25096002 http://scss.ornl.g:v/ scripts /scss/results/rLERDetll.cfm?lernmbr=25096002 Contact Mike Poets at ORNL with questions or comments concerning SCSS Web site content.
Contact Dale Yellding at NRC with questions concerning SCSS Web site access.
Copyright C 1996 Oak Ridge National Laboratory (ORNL)
Last modified: January 31,1997 -
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'i 2 of 2 -
. 07/02/97 10:09:49
LER lleader Listing for 25486001 littp://scss.ornl. gov / scripts /scss/results/rLE RDetl l.c fm?lernmbr=25486001
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LER Header Listinefor LER Number: 25486001 Docket Number Year Report #
Rev #
DCS Number Event Date 254 1986 001 1
8607250259 01/06/86 Nuclear Plant: Quad Cities Unit #1 NRC Region:
3 Facility Operator: Commonwealth Edison Co.
Architect Engineer: Sargent & Lundy NSSS Vendor: General Electric Reactor Type: Boiling Water Reactor Comm. Operations Date: 02/18/73 p.*l r m M 13 3i g s fid ABSTRACT POWER LEVEL - 000%. ON JANUARY 6,1986, WHILE PERFORMING LOCAL LEAK RATE TESTING THE MEASURED COMBINED LEAKAGE RATE FOR ALL VALVES AND PENETRATIONS, EXCEPT MAIN STEAM ISOLATION VALVES, WAS FOUND TO LEAK IN EXCESS OF 293.75 SCFH (.060 LA) WHICil IS ALLOWED BY THE PLANT TECHNICAL SPECIFICATIONS. UNIT ONE WAS SHUTDOWN FOR THE END OF CYCLE EIGIIT REFUELING AND MAINTENANCE OUTAGE. THIS REPORT DOCUMENTS THE REPAIRS MADE TO VALVES AND PENETRATIONS WITIl UNACCEPTABLE LEAK RATES AND THE FINAL RESULTS OF THE LOCAL LEAK RATE TESTING PROGRAM. THIS REPORT IS SUBMITTED IN ACCORDANCE WITH THE REQUIREMENTS OF 10 CFR 50.73 (A)(2)(II),
WHICH REQUIRES THE REPORTING OF ANY EVENT OR CONDITION THAT RESULTED IN Tile CONDITION OF THE NUCLEAR POWER PLANT, INCLUDING ITS PRINCIPLE SAFETY BARRIER, BEING SERIOUSLY DEGRADED.
Unit Conditions at Time of Reportable Event:
Power Level: 0%
Operating Mode: Unknown Reportability Reason (s) for This LER:
10 CFR 50.73(a)(2)(ii)- Unanalyzed Conditions. [11]
Primary Cause(s) for This LER:
There is no Primary Cause for This LER Emergency Classification (s) for This LER:
There is no Emergency Classification for This LER Referenced LERs:
I of 2 07/02/97 10:10:20
- LER Header Listing for 23486001 http://scss.ornl. gov / scripts /scss/results/rLERDetil.cfm71ernmbr=25486001 LER Event Number '
' Reactor Unit Name Date LER Title
- 25484002 Quad Cities Unit #1 03/07/84 The title for this LER is not currently available.
26585007
. Quad Cities Unit #2 03/18/85 The title for this LER is not currently available.
Contact Mike Poore at ORNL with questions or comments concerning SCSS Web site content.
Contact Dalc1tilding at NRC with questions concerning SCSS Web site access.
Copyright C 1996 Oak Ridge National Laboratory (ORNL)
Last modified: January 31,1997 2 of 2 '
07/02/97 10:10:20 l
http /scss.ornl. gov / scripts /scss/rzsults/rLERDetll.cfm71ernmbr=25487001 1
LER thader Listing for 25487001 O
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LER Header Listingfor LER Number: 2548700L Docket Number Year Report #
Rev #
DCS Number Event Date 254 1987 001 0
8702040421 01/05/87 Nuclear Plant: Quad Cities Unit #1 NRC Region:
3 Facility Operator: Commonwealth Edison Co.
Architect Engineer: Sargent & Lundy NSSS Vendor: General Electric Reactor Type: Boiling Water Reactor Comm. Operations Date: 02/18/73 f~#17 KeiiEliiiiillill EHiMENilEEN ABSTRACT POWER LEVEL - 100%. ON 1-5-87, UNIT I WAS IN Tile RUN MODE AT 100% OF RATED i
CORE TiiERMAL POWER. Tile REACTOR WATER CLEANUP (RWCU) SYSTEM liAD BEEN l
SIIUTDOWN EARLIER TIIAT DAY FOR INSTRUMENT REPAIRS. AT 1635 HOURS, TIIE RWCU SYSTEM ISOL.ATION VALVES CLOSED DUE TO A HIGil TEMPERATURE SIGNAL ON Tile OUTLET OF Tile NON-REGENERATIVE IIEAT EXCIIANGER. DUE TO MISCOMMUNICATION BETWEEN THE NUCLEAR STATION OPFRATOR A'ND THE SillFT ENGINEER,IT WAS NOT REALI7ED TIIAT AN ESF ACTUATION IIAD OCCURRED. IT SilOULD BE NOTED THAT Tills IS NOT A GROUP 111 ISOLATION SIGNAL LISTED IN TECli SPEC TABLE 3.7-1. DUE TO Tile MISCOMMUNICATION, NRC NOTIFICATION WAS NOT MADE UNTIL 2100 IIRS; 25 MINS LATER TIIAN THE 4 IIR NOTIFICATION LIMIT. THE CAUSE FOR THE RWCU SYSTEM lllGH TEMPERATURE VALVE CLOSURE WAS THE RESULT OF LEAKAGE TIIROUGli Tile CilECK VALVE IN THE RETURN TO TIIE FEEDWATER SYSTEM AND Tile RWCU RECIRCULATION PUMP DISCilARGE CHECK VALVES. TIIE CAUSE FOR THE TARDY NRC NOTIFICATION WAS Tile RESULT OF MISCOMMUNICATION BETWEEN OPERATING PERSONNEL. THE CllECK VALVES WILL BE REPAIRED DURING AN OUTAGE OF SUFFICIENT DURATION. Tile OPERATING DEPARTMENT, INCLUDING Tile INDIVIDUALS INVOLVED,IIAS BEEN ADVISED THAT Tills RWCU SYSTEM VALVE CLOSURE DUE TO lilGH 'lEMPERATURE IS TO BE CONSIDERED AN ESF ACTUATION AND NRC NOTIFICATION IS REQUIRED. THIS REPORT IS SUBMITTED TO COMPLY WITH THE REQUIREMENTS OF 10CFR50.73(A)(2)(IV).
Unit Conditions at Time of Reportable Event:
Power Level: 100%
Operating Mode: Unknown Reportability Reason (s) for This LER:
10 CFR 50.73(a)(2)(iv)- ESF Actua' ions. [13]
Iof2 07/02/97 10:11:35
http /scss.ornl. gov / scripts /scss/results/rLERDetll.cfm?lernmbr=25487001
!' LERlieader Listing for 25487001 1
Primary Cause(s) for This LER:
Equipment Failure [20 ]
Iluman Error [35 ]
Emergency Classification (s) for This LER:
There is no Emergency Classification for This LER
' N'aierenced LERs:
~~l2R Event Nunsber Reactor UnitName Date LER Title Y8585017' Quad Cities Unit #2 08/15/85 The title for this LER is not currently available.
4 l
Contact blike Poore at ORNL with uestions or comments concerning SCSS Web site content.
Contact Dale Yellding at NRC with quest ns concerning SCSS Web site access.
Copyright C 1996 Oak Ridge National Laboratory (ORNL)
Last modified: January 31,1997 i
I 2 of 2 07/02/97 10:11:35
LER lbder Listing for 25487016
@://s css.ornl. gov / scripts /scss/re sults/r L ERDeti l.c fm?lern ra br=25487016 h,M,M7M
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LER Header Listingfor LFR Number: 25487016 1
Docket Number Year Report #
Rev #
DCS Number Event Date 254 1987 016 2
9206300136-09/12/87 Nuclear Plant: Quad Cities Unit #1 NRC Region:
3 Facility Operator: Commonwealth Edison Co.
Architect Engineer: Sargent & Lundy NSSS Vendor: General Electric Reactor Type: Boiling Water Reactor Comm. Operations Date: 02/18/73 M EMEl EFSiliGEIE1 J
ABSTRACT POWER LEVEL - 000%. SEPTEMBER 12,1987 QUAD CITIES UNIT ONE WAS SHUTDOWN FOR TIIE END OF CYCLE REFUELING MAINTENANCE OUTAGE. AT 1600 HOURS,IT WAS DETERMINED THAT THE MEASURED COMBINED LEAKAGE RATE FROM ALL i
PENETRATIONS AND VALVES, EXCLUDING THE MAIN STEAM ISOLATION VALVES, EXCEEDED TlIE TECHNICAL SPECIFICATION (3.7.A.2) LIMIT OF 293.75 SCFH (0.60 LA).
tills WAS IDENTIFIED WHILE LOCAL LEAK RATE TESTING THE MAIN STEAM LINE t
DRAIN VALVES, MO 720-1 AND 2. THE FAILURE MODE OF THE PENETRATIONS AND VALVES WAS FOUND TO BE GENERALLY DUE TO NORMAL WEAR. REPAIRS AND REPLACEMENTS WERE COMPLETED AS NECESSARY AND THE RESULTS OF THE LOCAL LEAK RATE TESTING PROGRAM ARE PROVIDED. THIS REPORT IS SUBMITTED TO COMPLY WITH THE REQUIREMENTS OF 10CFR50.73(A)(2)(II).
Unit Conditions at Time of Reportable Event:
Power Level: 0%
Operating Mode: Unknown Reportability Reason (s) for This LER:
1 10 CFR 50.73(a)(2)(ii)- Unanalyzed Conditions. [l1]
i Primary Cause(s) for This LER:
Equipment Failure [20 )
]
Emergency Classification (s) for This LER:
There is no Emergency Classification for This LER Referenced LERs:
i I
-1of2 07/02/97 10:12:10
http /scss.ornl. gov / scripts /scss/results/rLERDetll.cfm71ernmbr=25437016 1
LER 11eader Listing for 25487016 LER Event Number Reactor Unit Name Date LER Title 03/07/84 The title for this LER is not currently available.
25484002 Quad Cities Unit #1 LEAK RATE FROM ALL VALVES AND 25486001 Quad Cities Unit #1 01/06/86 PENETRATIONS ON UNIT ONE IN EXCESS
' OF TECHNICAL SPECIFICATION LIMIT Unit Leak Rate from all valves and pe,netrations on,t
' 25487016 Q'uad Cities Unit #1 09/12/87 One m excess of Technical Specification Limi Exceedence of Technical Specification Local Leak 25490029 Quad Cities Unit #1 11/15/90 Rate Test Limit 0.6 La While Testing The Containment Isolation Valves And Penetrations Contact Mike Pmis at ORNL with questions or comments concerning SCSS Web site content.
Contact Dale Yellding at NRC with questions concerning SCSS Web site access.
Copyright C 1996 Oak Ridge National Laboratory (ORNL)
Last modified: January 31,1997 1
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1 2 0f 2 07/02/97 10:12:10 Y
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- LER lieader Listing for 25490010 http://scss.ornl. gov / scripts /scss/r suits /rLERDetll.cfm?lernmbr=25490010 ry
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. k LER Header Listingfor LER Number: 25490010 Docket Number Year Report #
Rev#
DCS Number Event Date 254 1990 010 0
9006260006 05/22/90 Nuclear Plant: Quad Cities Unit #1 NRC Region:
3 Facility Operator: Commonwealth Edison Co.
Architect Engineer: Sargent & Lundy NSSS Vendor: General Electric Reactor Type: Boiling Water Reactor Comm. Operations Date: 02/18/73 BIERNAlililB1 F4HlfMii6El BBRifiiriq ABSTRACT POWER LEVEL - 097%. ON MAY 22,1990 AT 1242 HOURS, UNIT ONE WAS IN THE RUN MODE AT 97 PERCENT OF RATED CORE TIIERMAL POWER. WlIILE RETURNING THE REACTOR WATER CLEAN-UP (RWCU) SYSTEM TO SERVICE, A NON-REGENERATIVE IIEAT EXCIIANGER (NRIIX) HIGil TEMPERATURE ALARM WAS RECEIVED. ALTHOUGH NOT AN ENGINEERED SAFETY FEATURE (ESF), GROUP III ISOLATION, THIS RESULTED IN A CIIALLENGE TO THE ESF LOGIC AND A SYSTEM ISOLATION. AN EMERGENCY NOTIFICATION SYSTEM (ENS) PHONE NOTIFICATION WAS MADE AT 1405 HOURS IN ACCORDANCE WITH 10CFR50.72(B)(2)(II). THE CAUSE WAS DETERMINED TO BE DUE TO SYSTEM CllECK VALVES LEAKING. WORK REQUESTS WILL BE WRITTEN TO INSPECT AND REPAIR THE RWCU RECIRC PUMPS DISCHARGE CHECK VALVES, 1-1201-87A/87B, AND THE FEEDWATER RETURN CIIECK VALVE, 1-1201-81. THIS REPORT IS BEING SUBM11TED TO COMPLY WITH THE REQUIREMENTS OF 10CFR50.73(A)(2)(IV).
Unit Conditions at Time of Reportable Event:
Power Level: 97%
Operating Mode: Unknown Reportability Reason (s) for This LER:
10 CFR 50.73(a)(2)(iv)- ESF Actuations. [13]
Primary Cause(s) for This LER:
Equipment Failure [20 ]
Emergency Classification (s) for This LER:
There is no Emergency Classification for This LER Referenced LERs:
1of2 07/02/97 10:12:40
http /scss.ornl. gov / scripts /scss/results/rLERDetll.cfm71ernmbr=25490010 1
LER Header Listing for 25490010 LER Event Number Reactor Unit Name Date LER Title Reactor Water Cleanup System Valve Closure 25487001 Quad Cities Unit #1 01/05/87 Due to High Non-Regen Heat Exchanger Outlet lTemperature Contact Mike Poore at ORNL with uestions or comments concerning SCSS Web site content.
Contact Dale YeindLie at NRC with quest ns concerning SCSS Web site access.
Copyright C 1996 Oak Ridge National Laboratory (ORNL)
Last modified: January 31,1997 i
i l
2 ef 2 '-
07/02/97 10:12:40
http /scss.ornl. gov /scriptdscss/resultdrLERDetil.cfm?lernmbr=25490029 LER licader Listing for 25690029 J
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LER Header Listingfor LER Number: 25490029 Docket Number Year Report #
Rev#
DCS Number Event Date 254 1990 029 1
9207070111 11/15/90 Nuclear Plant: Quad Cities Unit #1 NRC Region:
3 Facility Operator: Commonwealth Edison Co.
Architect Engineer: Sargent & Lundy NSSS Vendor: General Electric Reactor Type: Boiling Water Reactor Comm. Operations Date: 02/18/73 M ninjl% WE W ABSTRACT POWER LEVEL - 000%. ON NOVEMBER 12,1990, QUAD CITIES UNIT ONE WAS SilUTDOWN FOR THE END OF CYCLE I1 REFUELING AND MAINTENANCE OUTAGE. ON NOVEMBER 15,1990, AT 0700 liOURS WillLE PERFORMING LOCAL LEAK RATE TESTING (LLRT) OF TIIE DRYWELL PERSONNEL AIR LOCK,IT WAS DETERMINED THAT THE TECHNICAL SPECIFICATION 3.7.A.2.D. LEAKAGE LIMIT OF 18.4 STANDARD CUBIC FEET PER HOUR (SCFil),0.0375LA, WAS EXCEEDED. ON NOVEMBER 15,1990, AT 2325 HOURS WillLE PERFORMING LLRT ON THE FEEDWATER CHECK VALVES, VALVE l-220-62B COULD NOT BE PRESSURIZED. THE TECHNICAL SPECIFICATION 3.7.A.2.A.2, LIMIT OF 293.75 SCFH (0.6LA) WAS EXCEEDED. AN EMERGENCY NOTIFICATION SYSTEM (ENS)
PilONE CALL WAS COMPLETED ON NOVEMBER 16,1990 AT 0318 HOURS IN ACCORDANCE WITH 10CFR50.72(B)(2)(I). THE CAUSE OF THE EXCESSIVE LEAKAGES WAS IDENTIFIED AND REPAIRS HAVE BEEN COMPLETED. THIS REPORT IS BEING SUBMITTED TO COMPLY WITH 10CFR50.73(A)(2)(I)(B). SUBSEQUENT TESTING HAS IDENTIFIED ADDITIONAL FAILURES.
Unit Conditions at Time of Reportable Event:
Power Level: 0%
Operating Mode: Unknown Reportability Reason (s) for This LER:
10 CFR 50.73(a)(2)(i) - Shutdowns or Technical Specification Violations. [10]
Primary Cause(s) for This LER:
Equipment Failure [20 ]
Poor Ergonomics or Human Environment (38 )
Emergency Classification (s) for This LER:
Iof2 07/02/97 10:13:12
http /scss.ornl. gov / scripts /scss/results/rLERDetll.cfm?ltrnmbr=25490029 LER lleader Listing for 25490029 l
There is no Emergency Classification for This LER Referenced LERs:
LER Event Number Reactor Unit Name Date LER Title 25487016 Quad Cities Unit #1 09/12/87 Leak Rate from all valves and penetrations,on, Unit One in excess of Technical Specification Limit Exceeding Technical Specification Leakage Limits 25489014 Quad Cities Unit #1 09/10/89 for Containment Isolation Valves and Main Steam Isolation Valves - Causes to be Determined Unit Two m,From All Valves and Pen,etrations on Leak Rate 26586014 Quad Cities Unit #2 10/12/86 -
Excess of Tech Spec Limit LEAK RATE FROM ALL AND 26588007 Quad Cities Unit #2 04/20/88, PENETRATIONS IN EXCESS OF SPECIFICATION LIMITS 1
Conta'ct Mike Poore at ORNL with uestions or comments concerning SCSS Web site content.
Contact Dale Yeilding at NRC with quest ons concerning SCSS Web site access.
Copyright C 1996 Oak Ridge National Laboratory (ORNL)
Last modified: January 31,1997 2cf2 07/02/97 10:13:13-
http /scss.ornt. gov / scripts /scss/results/rLERDetil.cfm71ernmbr=25492020 1
l LER Header Listing for 25492020 I
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LER Header Listinefor LER Number: 25492010 Docket Number Year Report #
Rev#
DCS Number Event Date 254 1992 020 1
9302220281 09/21/92 Nuclear Plant: Quad Cities Unit #1 NRC Region:
3 Facility Operator: Commonwealth Edison Co.
Architect Engineer: Sargent & Lundy NSSS Vendor: General Electric Reactor Type: Boiling Water Reactor Comm. Operations Date: 02/18/73 u;T T;3 F 6 BHi4BiEMI ABSTRACT POWER LEVEL - 000%. On September 20,1992, Quad Cities Unit One was shutdown for refueling maintenance (Q1R12). On September 21,1992 at 11:30 hours, while performing Local Leak Rate Testing (LLRT) of the Atmospheric Containment, Atmospheric Dilution (ACAD) to Standby Gas Treatment (SBGT) Containment isolation Valve AO-1-2599-5B [lK] [Bil it was determined that the
]
Technical Specification 3.7.A.2.a.2 limit of 293.75 SCFil(0.6 La) was exceeded. An Emergency Notification System (ENS) phone call was completed on September 21,1992 at 1306 (EST) hcurs in accordance with 10CFR50.72(b)(2)(i). The cause of the excessive leakages was identified and repairs have been completed. Additional failed valves were identified during subsequent testing of the remaining isolation valve volumes. This report is being submitted to comply with 10CFR50.73(a)(2)(ii). DVR 373 Unit Conditions at Time of Reportable Event:
Power Level: 0%
Operating Mode: Unknown Reportability Reason (s) for This LER:
10 CFR 50.73(a)(2)(ii)- Unanalyzed Conditions. [l1]
Primary Cause(s) for This LER:
Equipment Foilure [20 ]
Emergency Classification (s) for This LER:
There is no Emergency Classification for This LER Referenced LERs:
i 1 cf 2 07/02/97 10:13:45
1 ER licader 1.isting for 15492020 http://scss.ornl. gov / scripts /scss/results/rLE RDetl l.c fm?lernmbr=25492020 LER Event Number Reactor Unit Name Date LER Title 25487016 Quad Cities Unit #1 09/12/87 - Leak Rate from all valves and penetrations,on,t Unit One in excess of Technical Specification Limi Exceeding Technical Specification Leakage Limits 25489014 Quad Cities Unit #1 09/10/89 for Containment isolation Valves and Main Steam isolation Valves - Causes to be Determined Exceedence ofTechnical Specification Local Leak 21490029 Quad Cities Unit #1 11/15/90 Rate Test Limit 0.6 La While Testing The Containment Isolation Valves And Penetrations 26586014 Quad Cities Unit #2 10/12/86i Leak pte,1; rom All Valves and Pen,etrations on Unit Two in Excess of Tech Spec Limit Contact Mikehors at ORNL with questions or comments concerning SCSS Web site content.
Contact Dale Yellding at NRC with questions concerning SCSS Web site access.
Copyright C 1996 Oak Ridge National Laboratory (ORNL)
Last modified: January 31,1997 l
j
'l 2 of 2 07/02/97 10:13:46
LER ileader Listing for 25494005 http //scss.ornl. gov / scripts /scss/results/rLERDetil.cfm?lernmbr=25494005 1
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1 LER Header Liningfor LER Number: 25494005 Docket Number Year Report #
Rev#
DCS Number Event Date 254 1994 005 1
9412290205 03/14/94 Nuclear Plant: Quad Cities Unit #1 NRC Region:
3 Facility Operator: Commonwealth Edison Co.
Architect Engineer: Sargent & Lundy NSSS Vendor: General Electric Reactor Type: Boiling Water Reactor Comm. Operations Date: 02/18/73 M %.TH:s8 W J%
ABSTRACT POWER LEVEL - 000%. At approximately 2330 on March 14,1994, Quad Cities Unit-1 was shutdown for the cycle 13 refueling and maintenance outage. After Local Leak Rate Testing (LLRT) the 'A' Main Steam Isolation Valves (h'SIV), it was determined that the measured leakage rate of 15.8 standard cubic feet per hour (SCFil) in the inboard MSIV AO-1-203-1 A exceeded the individual MSIV Technical Specification (3.7.A.2.a.3) leakage limit of 11.5 SCFII. The outboard MSIV AO-1-203-2A had a measured leakage rate of 0.35 SCFit On March 17,1994, at 0500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> while performing LLRT on the Unit-1 Feedwater (FW) check valve 1-220-58B, the valve failed with an undetermined leakage.
The unquantified leakage exceeds the 0.6 La (293.75 SCFil) combined leakage limits specified in Technical Specification 3.7.A.2.a.2. The cause of the excessive leakages have been determined where mssible, repairs have been completed and the valves / penetrations have been retested. This report is xing submitted to comply with 10CFR50.73(a)(2)(ii)(B).
i Unit Conditions at Time of Reportable Event:
Power Level: 0%
Operating Mode: Unknown i
Reportability Reason (s) for This LER:
j 10 CFR 50.73(a)(2)(ii)- Unanalyzed Conditions. (11]
Primary Cause(s) for This LER:
Equipment Failure [20 )
Emergency Classification (s) for This LER:
There is no Emergency Classification for This LER I
Referenced LERs-l l
1of2 07/02/97 10:14:33
http /scss.omi. gov / scripts /scss/results/rLERDetti.cfm?lernmbr=25494005 LCR ileader Listing for 25494003 1
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l LER Event Number Reactor Unit Name Date LER Title 25492020 Quad Cities Unit #1 09/21/9 {Tech Spec Containment Leakage Limit 0.6 La Exceeded Toxic Gas Analyzer Inoperable Duc To Personnel 25493007 Ouad Cities Unit #1 06/24/93 Error Exceedance of Technical Specitication Local Leak 26592002 Quad Cities Unit #2 01/03/92 Rate Test Limit 0.6 La While Testing the Containment Isolation Valves.
tL p Main Steam Isolation Valves Exceed Tech 26593025 Quad Cities Unit #2 12/05/93 Spec Leakage L,mits i
Contact Mike Poore at ORNL with questions or comments concerning SCSS Web site content.
Contact Dale Yellding at NRC with questions concerning SCSS Web site access.
Copyright O 1996 Oak Ridge National Laboratory (ORNL)
Last modified: January 31,1997 2 of 2 07/02/97 10:14:33
http /scss.ornl. gov / scripts /scsa/results/rLERDetll.cfm?lernmbr=25594006 1
LER Header Listing for 25594006 WZiTM i
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LER H1ader Listingfor LER Number: 25594006 Docket Number Year Report #
Rev#
DCS Number Event Date 255 1994 006 0
9404010017 02/17/94 Nuclear Plant: Palisades Unit #1 NRC Region:
3 Facility Operator: Consumers Power Co.
Architect Engineer: Bechtel Power Corp.
NSSS Vendor: Combustion Engineering Reactor Type: Pressurized Water Reactor Comm. Operations Date: 12/31/71 ftisiidiihiiiiiff CtiiiiiiiiiiiiiM FNiliilijjiiigj
- tBSTRACT POWER LEVEL - 100%. On Febmary 17,1994, at 1538 hours0.0178 days <br />0.427 hours <br />0.00254 weeks <br />5.85209e-4 months <br />, with the plant operating at 100%
3ower, an accumulation of boric acid on the valve body of check valve CK-ES3166 was confirmed to 3e caused by a through wall defect. Tne 24-inch austenitic stainless steel check valve is located between the containment sump and the suction piping for one train of Engineered Safeguards System (ESS) pumps, component in the Section XI program. The valve was declared inoperable. A Plant shutdown was initiated at 1634, February 17,1994 and the plant was taken to cold shutdown.
. Inspection of the comparable check valve in the other ESS train, CK-ES3181, identified indications of a similar nature in the casting of that valve. The cause of this event was a through wall defect in the body of an ASME, Class 2, check valve, CK-ES3166, due to preferential corrosion at the grain boundaries in a weld repaired region of the valve casting. The code does not allow operation of an ASME Class 2 component with a through wall leak. Corrective action for this event will be the repair of both check valves, CK-ES3166 and CK-ES3181.
Unit Conditions at Time of Reportable Event:
Power Level: 100%
Operating Mode: Unknown Reportability Reason (s) for This LER:
10 CFR 50.73(a)(2)(i) - Shutdowns or Technical Specification Violations. [10]
Primary Cause(s) for This LER:
Equipment Failure [20 ]
Emergency Classification (s) for This LER:
There is no Emergency Classification for This LER Referenced LERs:
I of 2 07/02/97 10:15:12
http /scss.ornl. gov / scripts /scss/results/rLERDetil.cfm71ernmbr=25594006
-LER lleader Listing for 25594006 1
There are no Referenced LERs for This LER j
1 1
Contact Mike Poore at ORNL with questions or comments concerning SCSS Web site content.
Contact Dale Yedding at NRC with questions concerning SCSS Web site access.
Copyright C 1996 Dak Ridge National Laboratory (ORNL)
Last modified: January 31,1997 1
1 I
' 2 OI2 07/02/97 10:15:12
LER IIcader Listing for 26585007 h ttp://sc ss.ornl. gov / scripts /scss/re sults/rL E RDeti l.c fm?lernmbr=26585007 KsinhantMHr:d.srpt:JiniMn!!sM4M "
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LER Header Listingfor LER Number: 26585007 Docket Number Year Report #
Rev#
DCS Number Event Date 265 1985 007 1
8507310084 03/18/85 Nuclear Plant: Quad Cities Unit #2 NRC Region:
3 Facility Operator: Commonwealth Edison Co.
Architect Engineer: Sargent & Lundy NSSS Vendor: General Electric Reactor Type: Boiling Water Reactor Comm. Operations Date: 03/10/73
[siiiIdiliiiiiii~rl i"Wiiiiiiiiiid IGiiFiiiiiPM ABSTRACT POWER LEVEL - 000%. ON MARCII 18,1985, QUAD-CITIES UNIT TWO WAS SHUT DOWN FOR REFUELING. WHILE PERFORMING REFUELING OUTAGE LOCAL LEAK RATE TESTING, THE MEASURED COMBINED LEAKAGE RATE FOR ALL PENETRATIONS AND VALVES, EXCEPT MAIN STEAM ISOLATION VALVES, WAS FOUND TO LEAK IN EXCESS OF 293.75 SCFH (0.61 LA). VALVES THAT HAVE A HISTORY OF EXCESSIVE LEAKAGE AND/OR LARGE LEAKAGE RATES INCLUDE THE FOLLOWING (1) MAIN STEAM LINE DRAIN VALVE 220-1;(2) ALL FEEDWATER CHECK VALVES 220-58A,B AND 220-62A,B;(3)
DRYWELL PURGE BUTTERFLY VALVES 1601-23,24,60,61,62,63; AND (4) HPCI STEAM EXIIAUST VALVE 2301-45.
Unit Conditions at Time of Reportable Event:
Power Level: 0%
Operating Mode: Unknown Reportability Reason (s) for This LER:
10 CFR 50.73(a)(2)(ii)- Unanalyzed Conditions. [11]
Primary Cause(s) for This LER:
There is no Primary Cause for This LER Emergency Classification (s) for This LER:
There is no Emergency Classification for This LER Referenced LERs:
There are no Referenced LERs for This LER 07/02/97 10:[5:40 Iof2
http /scss.oml. gov / scripts /scss/r:,sults/rLERDetti.cfm71:rnmbr=26585007 1
'i LER11 ender Listing for 26585007 Contact Mike Poore at ORNL with q' uestions or comments concerning SCSS Web site content,
. Contact DakJ(ellding at NRC with questions concerning SCSS Web site access.
Copyright C 1996 Oak Ridge National Laboratory (ORNL)
Last modified: January 31,1997 1
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. 2 of 2 '
~ 07/02/9710:15:40
- LER lleader Listing for 26588007 h ttp://scss.ornl. gov / scripts /scss/resu lts/rL ERDeti l.c fm71ernmbr=26588007
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LER Header Listinefor LER Number: 26S88007 Docket Number Year Report #
Rev #
DCS Number Event Date 265 1988 007 1
8909250264 04/20/88 Nuclear Plant: Quad Cities Unit #2 NRC Region:
3 Facility Operator: Commonwealth Edison Co.
Architect Engineer: Sargent & Lundy NSSS Vendor: General Electric Reactor Type: Boiling Water Reactor Comm. Operations Date: 03/10/73
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pi P E d iaiu M F y u irw E B1 ABSTRACT POWER LEVEL - 000%. ON 4/10/88, QUAD CITIES UNIT 2 WAS SHUT DOWN FOR THE END OF CYCLE 9 REFUELING AND MAINTENANCE OUTAGE. ON 4/13/88 AT 1630 HOURS,IT WAS DETERMINED TilAT THE MEASURED COMBINED LEAKAGE RATE FROM ALL PENETRATIONS AND VALVES, EXCLUDING THE MAIN STEAM ISOLATION VALVES, EXCEEDED TIIE TECH SPECS (3.7.A.2) LIMIT OF 293.75 STANDARD CUBIC FEET PER llOUR (SCFH)(0.60LA). TIIIS WAS IDENTIFIED WHILE LOCAL LEAK RATE TESTING THE 2-220 58B AND 2-220-62B FEEDWATER CHECK VALVES. THE FAILURE MODE OF THE PENETRATIONS AND VALVES WAS FOUND TO BE GENERALLY DUE TO NORMAL WEAR.
REPAIRS AND REPLACEMENTS WERE COMPLETED AS NECESSARY AND THE RESULTS OF TIIE LOCAL LEAK RATE TESTING PROGRAM ARE PROVIDED. THIS REPORT IS SUBMITTED TO COMPLY WITH THE REQUIREMENTS OF 10CFR50.73(A)(2)(II).
I Unit Conditions at Time of Reportable Event:
Power Level: 0%
Operating Mode: Unknown Reportability Reason (s) for This LER:
10 CFR 50.73(a)(2)(ii)- Unanalyzed Conditions. [l1]
Primary Cause(s) for This LER:
Equipment Failure [20 ]
Emergency Classification (s) for This LER:
There is no Emergency Classification for This LER Referenced LERs:
Iof2 07/02/97 10:16:45
)
LER Header Listing for 26588007 http://scss.ornl. gov /ecripts/scss/results/rLERDetl l.c fm71ernmbr=26588007
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LER Event Number Reactor UnitName Date LER Title 25484002)
Quad Cities Unit #1 03/07/84 The title for this LER is not currently available.
l LEAK RATE FROM ALL VALVES AND 25486001 Quad Cities Unit #1 01/06/86 PENETRATIONS ON UNIT ONE IN EXCESS OF TECHNICAL SPECIFICATION LIMIT
' 25487016' Quad Cities Unit #1 09/12/87 - One in excess of Techn,es and penetrations,on Un Leak Rate from all valv ical Specification Limit 26585007:
Quad Cities Unit #2 03/18/85-The title for this LER is not currently available.
1 Leak Rate From All Valves and Penetrations on 26586014 Quad Cities Unit #2 10/12/86 Unit Two, Excess of Tech Spec Limit m
Contact Mike Poore at ORNL with questions or comments concerning SCSS Web site content.
Contact Dale Yellding at NRC with questions concerning SCSS Web site access.
Copyright C 1996 Oak Ridge National Laboratory (ORNL)
Last modified: January 31,1997 2 of 3 07/02/97 10:I6:45
. LER IIcader Listing for 26590003 h ttp://scss.orn l. gov / scripts /scss/results/rL E RDeti l.c fm71ernmbr=26590003 f~'\\
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LER Header Listinefor LER Number: 26590003 Docket Number Year Report #
Rev#
DCS Number Event Date 265 1990 003 1
9207070098 02/05/90 Nuclear Plant: Quad Cities Unit #2 NRC Region:
3 Facility Operator: Commonwealth Edison Co.
Architect Engineer: Sargent & Lundy NSSS Vendor: General Electric Reactor Type: Boiling Water Reactor Comm. Operations Date: 03/10/73 MiliiNjBiliiiF1 r11sireiiiinuirm Femiiir?!
ABSTRACT POWER LEVEL - 100%. ON FEBRUARY 4,1990, QUAD CITIES UNIT TWO WAS SHUTDOWN FOR THE END OF CYCLE 10 REFUELING AND MAINTENANCE OUTAGE. ON FEBRUARY 5, 1990, AT 1730 HOURS, WHILE LOCAL LEAK RATE TESTING (LLRT) THE HIGH PRESSURE COOLANT INJECTION (HPCI) SYSTEM STEAM EXHAUST CHECK VALVE, IT WAS DETERMINED THAT Tile MEASURED LEAKAGE RATE OF 528.6 STANDARD CUBIC FEET PER HOUR (SCFH) EXCEEDED THE TECHNICAL SPECIFICATION 3.7.A.2.A.2 LIMIT OF 293.75 SCFH (0.60LA) FOR ALL VALVES AND PENETRATIONS EXCLUDING THE MAIN STEAM ISOLATION VALVES (MSIV). THE APPARENT CAUSE FOR EXCEEDING THE TECHNICAL SPECIFICATION 0.6 LA LIMIT WAS DUE TO THE EXCESSIVE LEAKAGE OF THE HPCI STEAM EXHAUST CIIECK VALVE. CORRECTIVE ACTION INCLUDED REPLACING THE VALVE AND MODIFYING THE SPARGER TO PREVENT SEAT EROSION, Tills REPORT IS BEING SUBMITTED TO COMPLY WITII 10CFR50.73(A)(2)(I)(B).
Unit Conditions at Time of Reportable Event:
Power Level: 0%
Operating Mode: Unknown Reportability Reason (s) for This LER:
10 CFR 50.73(a)(2)(i) - Shutdowns or Technical Specification Violations. [10]
Primary Cause(s) for This LER:
Equipment Failure [20 ]
Emergency Classification (s) for This LER:
There is no Emergency Classification for This LER
' Referenced LERs:
I of a 07/02/97 10:17:30
l 11R Header Listing for 26590003 http /scss.orn! g scripts /scss/results/rLERDetl l.c fm71ernm br=26590003 1
LER Event Number Reactor Unit Name Date LER Title j
Unit 25487016 Quad Cities Unit #1 09/12/87 - Leak Rate from all valv,es and penetrations on,t One in excess of Technical Specification Limi i
Exceeding Technical Specification Leakage Limits 25489014 Quad Cities Unit #1 09/10/89 for Containment Isolation Valves and Main Steam 1
Isolation Valves - Causes to be Determined Le k Rate From All Valves and Penetrations on
. 26586014 Quad Cities Unit #2
. 10/12/86 j
Unit Two in Excess of Tech Spec Limit LEAK RATE FROM ALL AND
. 26588007 Quad Cities Unit #2 04/20/88 PENETRATIONS IN EXCESS OF SPECIFICATION LIMITS Contact Mike Poore at ORNL with questions or comments concerning SCSS Web site content.
Contact flaltltilding at NRC with questions concerning SCSS Web site access.
Copyright C 1996 Oak Ridge National Laboratory (ORNL)
Last modified: January 31,1997 2 of 2 07/02/97 10:17:30
http /scss.ornl. gov / scripts /scss/results/rLERDctll.cfm71ernmbr=26592002 LER Header Listing for 26592002 1
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LER Header Listinefor LER Number: 26592002 Docket Number Year Report #
Rev#
DCS Number Event Date 265 1992 002 1
9207070090 01/03/92 Nuclear Plant: Quad Cities Unit #2 NRC Region:
3 Facility Operator: Commonwealth Edison Co.
Architect Engineer: Sargent & Lundy NSSS Vendor: General Electric Reactor Type: Boiling Water Reactor Comm. Operations Date: 03/10/73 M ITS!liiF#Fra EHaTdiiEd I
1 ABSTRACT POWER LEVEL - 000%. ON JANUARY 1,1992, QUAD CITIES UNIT TWO WAS SHUTDOWN FOR REFUELING AND MAINTENANCE (Q2RI1). ON JANUARY 3,1992 AT 0715 HOURS WHILE PERFORMING LOCAL LEAK RATE TESTING (LLRT) OF THE CONTAINMENT PURGE FOUR VALVE VOLUME 2-1601-21,22,55 AND 56. IT WAS DETERMINED THAT THE TECHNICAL SPECIFICATION 3,7.A.2.A.2 LIMIT OF 293. 75 SCFH (0.6 LA) WAS EXCEEDED.
AN EMERGENCY NOTIFICATION SYSTEM (ENS) PHONE CALL WAS COMPLETED ON JANUARY 3,1992 AT 1029 (EST) HOURS IN ACCORDANCE WITH 10CFR50.72(B)(2)(1).
ADDITIONAL FAILED VALVES WERE IDENTIFIED DURING SUBSEQUENT TESTING OF THE REMAINING ISOLATION VALVE VOLUMES THE CAUSE OF THE EXCESSIVE LEAKAGES WAS IDENTIFIED AND REPAIRS HAVE BEEN COMPLETED. THIS REPORT IS BEING SUBMITTED TO COMPLY WITH 10CFR50.73(A)(2)(II).
Unit Conditions at Time of Reportable Event:
Power Level: 0%
Operating Mode: Unknown Reportability Reason (s) for This LER:
10 CFR 50.73(a)(2)(ii)- Unanalyzed Conditions. [11]
Primary Cause(s) for This LER:
Equipment Failure [20 ]
Emergency Classification (s) for This LER:
There is no Emergency Classification for This LER Referenced LERs:
1 of 2 07/02/97 10:18:07
LER IIcader Listing for 26592002 http1/scss.ornl. gov / scripts /scss/results/rLE RDetl l.cfm?lernmbr=26592002
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LER Event Number Reactor Unit Name Date LER Title Leak Rate from all valv Unit One in excess of Techm,es and penetrations,on,t 25487016 Quad Cities Unit #1 09/12/87 cal Specificat:on Limi
. Exceeding Technical Specification Leakage Limits 25489014 Quad Cities Unit #1 09/10/89' for Containment Isolation Valves and Main Steam Isolation Valves - Causes to be Determined Exceedence of Technical Specification Local Leak 25490029 Quad Cities Unit #1 11/15/90 Rate Test Limit 0.6 La While Testing The Containment Isolation Valves And Penetrations Leak Rate From All Valves and Pen,etrations on 26586014 Quad Cities Unit #2 10/12/86: Unit Two in Excess of Tech Spec Limit l
Contact Mike Poort at ORNL with questions or comments concerning SCSS Web site content.
Contact Daltltilding at NRC with questions concerning SCSS Web site access.
Copyright O 1996 Oak Ridge National Laboratory (ORNL)
Last modified: January 31,1997 i
'2of2 07/02/97 10:I8:07 L
LER licader Listing for 26986008 http /scss.ornl. gov / scripts /scss/results/rLERDetll.cfm?lernmbr=26986008 1
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LER Hg.gder Listingfor LER Number: 26986008 Docket Number Year Report #
Rev #
DCS Number Event Date 269 1986 008 0
8600000000 05/10/86 Nuclear Plant: Oconee Unit #1 NRC Region:
2 Facility Operator: Duke Power Co.
Architect Engineer: Duke Power Co. & Bechtel Power Corp.
NSSS Vendor Babcock & Wilcox Reactor Type: Pressurized Water Reactor Comm. Operations Date: 07/15/73 fe~@"~~1 FisiiWeiis?! PfinitiisT1 ABSTRACT POWER LEVEL - 048%. ON 5-10-86, AT 0616 HOURS OCONEE UNIT 1 EXPERIENCED AN ANTICIPATORY REACTOR TRIP WHEN 'l A' MAIN FEEDWATER PUMP (MFWP) TRIPPED WHILE OPERATIONS PERSONNEL WERE ATTEMPTING TO PUT 1D1 HEATER DRAIN PUMP IN SERVICE. THE EVENT WAS INITIATED WHEN IDI HEATER DRAIN PUMP j
TRIPPED. Tile CilECK VALVE lHD-205 STUCK OPEN CREATING A BACKFLOW THROUGH l
Tile ID1 HEATER DRAIN PUMP TO THE ID FLASH TANK TO THE CONDENSER. THE CONDENSATE BOOSTER PUMPS TRIPPED ON LOW SUCTION PRESSURE CAUSING THE MAIN FEEDWATER PUMP TO TRIP AND A SUBSEQUENT ANTICIPATORY REACTOR TRIP.
THE ROOT CAUSE OF THIS EVENT WAS TIIE FAILURE OF CHECK VALVE IDH-205 TO CLOSE AS DESIGNED AND ALLOWED BACK FLOW.THE IMMEDIATE CORRECTIVE ACTION WAS TO STABILIZE THE UNIT AT IIOT SIlUTDOWN CONDITIONS AND TO REPAIR THE STICKING CllECK VALVE IIID-205. TIIE PLANT RESPONSE WAS NORMAL.
THE UNIT WAS RETURNED TO CRITICAL BY 1153 HOURS. THE TURBINE / GENERATOR WENT ON LINE AT 1403 HOURS.
Unit Conditions at Time of Reportable Event:
Power Level: 48%
Operating Mode: Unknown Reportability Reason (s) for This LER:
10 CFR 50.73(a)(2)(iv)- ESF Actuations. [13]
No CFR.Special Report, Part 21 Report, etc. [21]
Primary Cause(s) for This LER:
There is no Primary Cause for This LER Emergency Classification (s) for This LER:
Iof2 07/02/97 10:18:37
I
= LER11cader Listing for 26986008 http /scss.ornl. gov / scripts /scss/results/rLERDetll.cfm?lernmbr=26986008 J
4 There is no Emergency Classification for This LER Referenced LERs:
There are no Referenced LERs for This LER Contact Mike.foort at ORNL with questions or comments concerning SCSS Web site content.
J Contact Dale Yellding at NRC with questions concerning SCSS Web site access.
Copyright C 1996 Oak Ridge National Laboratory (ORNL)
Last modified: January 31,1997 l
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3 of 2 07/02/97 10:18:37
http /scss.orni. gov / scripts /scss/results/rLERDetti.cfm71ernmbr=28186020 1
LER IIcader Listing for 28186020 I6Mmtfratrimmr.sr7JIJ2fgfg r '\\
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LER Header Listingfor LER Number: 28186020 Docket Number Year Report #
Rev#
DCS Number Event Date 281 1986 020 2
8704130361 12/09/86 Nuclear Plant: Surry Unit #2 NRC Region:
2 Facility Operator: Virginia Electric Power Co.
Architect Engineer: Stone & Webster Engr., Corp.
NSSS Vendor: Westinghouse Reactor Type: Pressurized Water Reactor Comm. Operations Date: 05/01/73 EONil FiEia6afiffr4 F#iiiiiisEl ARSTRACT POWER LEVEL - 100%. ON DECEMBER 9,1986 AT 1420, WITH UNITS 1 AND 2 AT 100%
POWER, TIIE UNIT 2 REACTOR TRIPPED DUE TO LOW WATER LEVEL IN 'C' STEAM GENEl?ATOR (Ells-SG). THE EVENT WAS INITI ATED BY THE CLOSURE OF 'C' MAIN STEAM TRIP VALVE (MSTV)(EIIS-ISV). APPROXIMATELY 40 SECONDS AFTER THE TRIP, A CARBON STEEL ELBOW IN THE 18 INCH SUCTION PIPE TO THE 'A' MAIN FEED PUMP (Ells-P) RUPTURED CAUSING A LOSS OF NORMAL FEEDWATER. WATER FLASHING FROM THE SEVERED PIPE ENGULFED EQUIPMENT AND PERSONNEL IN THE AREA.
SEVERAL WORKERS WERE SERIOUSLY BURNED. OPERATORS FOLLOWED APPROPRIATE PLANT PROCEDURES AND QUICKLY STABILIZED THE UNIT FOLLOWING THE TRIP. CLOSURE OF Tile 'C' MSTV WAS DUE TO IMPROPER REASSEMBLY OF THE VALVE FOLLOWING MAINTENAN'CE. THE RUPTURED FEEDWATER PIPING HAS BEEN ATTRIBUTED TO PIPE WALL THlhdlNG DUE TO AN EROSION / CORROSION PHENOMENON. UNIT 1 WAS SIIUT DOWN ON DECEMBER 10,1986 AS A PRECAUTIONARY MEASURE. BOTH UNITS MSTVS WERE INSPECTED AND TESTED PRIOR TO STARTUP. A SYSTEMATIC INSPECTION OF BOTH UNITS WAS CONDUCTED 1
AND PIPING EXHIBITING UNACCEPTABLE THINNING HAS BEEN REPLACED.
Unit Conditions at Time of Reportable Event:
Power Level: 100%
Operating Mode: Unknown Reportability Reason (s) for This LER:
10 CFR 50.73(a)(2)(iv)- ESF Actuations. [13]
Primary Cause(s) for This LER:
There is no Primary Cause for This LER Emergency Classification (s) for This LER:
1 of 2 07/02/97 10:19:08
LER lleader Listing for 28186020 h ttp //scss.ornl.go v/ scripts /scss/results/rL ERDet t i.c fm71ernmbr=28186020 Alert [943]
- Referenced LERs:
There are no Referenced LERs for This LER Contact Mike Poort at ORNL with guestions or comments concerning SCSS Web site content.
. Contact Dale Yellding at NRC with questions cancerning SCSS Web site access.
Copyright C 1996 Oak Ridge National Laboratory (ORNL)
Last modified: January 31,1997 l
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. 3of2 07/02/97 10:19:08 s
LER Header Listing for 28588010 http://scss.ornl. gov / scripts /scss/results/rLERDctil.cfm71ernmbr=28588010
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MM LER Header Listing for LER Number: 28588010 Docket Number Year Report #
Rev #
DCS Number Event Date 285 1988 010 0
8805190295 04/15/88 Nuclear Plant: Ft. Calhoun Unit #1 NRC Region: 4 Facility Operator: Omaha Public Power District Architect Engineer: Gibbs & Hill, Inc.
NSSS Vendor: Combustion Engineering Reactor Type: Pressurized Water Reactor Comm. Operations Date: 06/20/74 reif!pwmiiR1 Frnsiiiiiiiiid FEiiiiiiiiiiG1 ABSTRACT POWER LEVEL - 100%. ON APRIL 15,1988 AT 1455 HOURS WHILE OPERATING AT 100 PERCENT POWER, T6 STING REVEALED THAT CHECK VALVES IN INSTRUMENT AIR LINES TO BUBBLER LEVEL INSTRUMENTATION ON THE SAFETY INJECTION AND REFUELING WATER TANK (SIRWT) FAILED TO HOLD A BACK-PRESSURE, AS WOULD BE REQUIRED AFTER A LOSS OF INSTRUMENT AIR. IF A LOCA OCCURRED WITH A COINCIDENT LOSS OF INSTRUMENT AIR PRESSURE UNDER THIS CONDITION, IT IS POSSIBLE THAT A RECIRCULATION ACTUATION SIGNAL WOULD HAVE ACTUATED EARLIER IN THE TRANSIENT THAN DESIGNED, RESULTING IN A LOSS OF SAFETY INJECTION AND CONTAINMENT SPRAY FLOW. AT 1539 HOURS ON APRIL 15,1988, THE NRC WAS NOTIFIED IN ACCORDANCE WITH 10 CFR 50.72 B.I.II.B. UPON DISCOVERY OF Tile FAILURE, THE CilECK VALVES WERE REPLACED WITH A DIFFERENT TYPE CHECK VALVE. THE NEW VALVES WERE TESTED TO ENSURE OPERABILITY, AND THE SYSTEM WAS RETURNED TO NORMAL. TO ENSURE CONTINUED OPERABILITY, THE CHECK VALVES IIAVE BEEN INCORPORATED INTO TiiE STATION S IN-SERVICE-INSPECTION (ISI) PROGRAM. PRESENT PLANS CALL FOR THE VALVES TO BE TESTED AS PART OF Tile ISI PROGRAM DURING THE SCHEDULED 1988 OUTAGE.
Unit Conditions at Time of Reportable Event:
Power Level: 100%
Operating Mode: Unknown Reportability Reason (s) for This LER:
10 CFR 50.73(a)(2)(ii) - Unanalyzed Conditions. [11]
Primary Cause(s) for This LER:
Spurious / Unknown Cause [60 ]
' Emergency Classification (s) for This LER:
, I of 2 07/02/97 10:19:41
http /scss.ornl. gov / scripts /scss/results/rLERDetil.cfm?lernmbr=28588010 LER IIcader Li* ting for 28588010 ':
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' There is no Emergency Classification for This LER '-
Referenced LERs:
l There are no Referenced LERs for This LER Contact Mjke Poore at ORNL with questions or comments concerning SCSS Web site content.
Contact Dale Yelldine st NRC with questions concerning SCSS Web site access.
Copyright O 1996 Oak Ridge National Laboratory (ORNL)
Last modified: January 31,1997 i
i
. 2 of 2 07/02/97 10:19:42
LER licader Listing for 28686003 http ://scss.orn i. gov / scripts /sc ss/results/r LE RDeti l.c fm?lernm br=28686003 O'\\,
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DCS Number Event Date 286 1986 003 0
8606240342 05/19/86 Nuclear Plant: Indian Point Unit #3 NRC Region:
1 Facility Operator: Power Authority of the State of NY Architect Engineer: United Engineers & Construction NSSS Vendor: Westinghouse Reactor Type: Pressurized Water Reactor Comm. Operations Date: 08/30/76
!)WilliiiiiEM Feiiiiisiid PWailiinM1 ABSTRACT POWER LEVEL - 025%. ON MAY 19,1986 WITH THE REACTOR AT 25 PERCENT POWER AND NO. 31 MAIN BOILER FEEDWATER PUMP (MBFP)IN SERVICE, A TURBINE TRIP AND SUBSEQUENT REACTOR TRIP WERE INITI ATED AUTOMATICALLY BY HIGH WATER LEVEL IN NO. 34 STEAM GENERATOR (SG). DURING INITIAL STARTUP OF NO. 32 MBFP FEEDWATER PRESSURE AND FLOW DECREASED, CAUSING A SIMILAR DECREASE IN SG LEVELS. IN ORDER TO MAINTAIN LEVEL, THE SPEED OF THE OPERATING MBFP WAS INCREASED, AND NO. 32 MBFP WAS SHUTDOWN. A RAPID RISE IN SG LEVELS SUBSEQUENTLY OCCURRED, WITII NO. 34 SG REACHING THE HIGli LEVEL TRIP SETPOINT FIRST. INVESTIGATION DETERMINED THAT THE CHECK VALVE IN THE DISCilARGE LINE OF NO. 32 MBFP WAS OPEN, ALLOWING A FEEDWATER REVERSE FLOW PATH TO DEVELOP DURING PUMP START-UP. WIlEN 32 MBFP WAS TRIPPED, THE REVERSE FLOW PATH WAS SUDDENLY CLOSED, AND AN OVER-FEEDING CONDITION DEVELOPED WillCH SUBSEQUENTLY TRIPPED Tile UNIT ON HIGH SG LEVEL.
INSPECTION OF THE FAILED VALVE FOUND THAT THE VALVE DISC HAD BECOME LOOSE. TO PREVENT RECURRENCE, AN IMPROVED METHOD OF ANCHORING THE DISC PlVOT PINS WAS IMPl.EMENTED.
Unit Conditions at Time of Reportable Event:
Power Level: 25%
Operating Mode: Unknown Reportability Reason (s) for This LER:
10 CFR 50.73(a)(2)(iv)- ESF Actuations. [13]
Primary Cause(s) for This LER:
There is no Primary Cause for This LER Emergency Classification (s) for This LER:
Iof2 07/02/97 10:20:26
. http /scss.ornl. gov / scripts /scss/results/rLERDetti.cfm?lernmbr=28686003 LER Header Listing for 2!686003 J
There is no Emergency Classification for This LER 1
Referenced LERs:
There are no Referenced LERs for This LER Contact Mike Poort at ORNL with questions or comments concerning SCSS Web site content.
Contact Dale Yellding at NRC with questions concerning SCSS Web site access.
Copyright C 1996 Dak Ridge National Laboratory (ORNL)
Last modified: January 31,1997 i
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07/02/97 10:20:27
LER Header Listing for 28691005 http://scss.ornl. gov / scripts /scss/results/rLERDetl l.c fm?lernmbr=28691005 lsrtFrt/M d h rr.n e m
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LER Header Listingfor LER Number: 28691005 Docket Number Year Report #
Rev#
DCS Number Event Date 286 1991 005 0
9104240431 03/22/91 Nuclear Plant: Indian Point Unit #3 NRC Region:
1 Facility Operator: Power Authority of the State of NY Architect Engineer: United Engineers & Construction NSSS Vendor: Westinghouse Reactor Type: Pressurized Water Reactor Comm. Operations Date: 08/30/76 MiiiiiMiWiiili#~ 1 FWiiiiRNiiid EfiitViiiiW1 ABSTRACT POWER LEVEL - 025%. ON MARCH 22,1991, WITH THE REACTOR AT 25 PERCENT POWER, A UNIT TRIP WAS INITIATED AS Tile RESULT OF A STEAM GENERATOR LOW-LOW LEVEL TRIP. ALL PLANT SYS" EMS FUNCTIONED PROPERLY FOLLOWING THE TRIP. THE CAUSE OF THIS EVENT WAS DETERMINED TO BE CYCLIC FATIGUE FAILURE OF THE LOCKING PIN ON TIIE 31 MAIN BOILER FEED PUMP CHECK VALVE. THE DISCHARGE CHECK VALVES ON BOTil MAIN FEED PUMPS WERE OVERHAULED AND RETESTED.
Unit Conditions at Time of Reportable Event:
Power Level: 25%
Operating Mode: Unknown Reportability Reason (s) for This LER:
10 CFR 50.73(a)(2)(iv)- ESF Actuations. [13]
Primary Cause(s) for This LER:
Equipment Failure [20 ]
Emergency Classification (s) for This LER:
There is no Emergency Classification for This LER Referenced LERs:
LER Event Number Reactor Unit Name Date LER Title UNIT TRIP CAUSED BY FAULTY CHECK 28686003 Indian Point Unit #3 05/19/86 VALVE IN MAIN FEEDPUMP DISCHARGE
- LINE Iof2 07/02/97 10:20:57
. LER IIcader Listing for 28691005
. http://scss.ornl. gov / scripts /scss/results/rLERDetll.cfm71ernmbr=28691005 Contact Mike Poore at ORNL with uestions or comments concerning SCSS Web site content.
Contact Dale Yellding at NRC with quest as concerning SCSS Web site access.
Copyright c 1996 Oak Ridge National Laboratory (ORNL) -
Last modified: January 31,1997 '.
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07/02/97 10:20:57-
- FIELD FIELD TITLEICODING INSTRUCTIONS 6
PLANT DOCKET NUMBER:
CODING INSTRUCTIONS:
For LERs marked "Y" in Field 2 or 3 above, enter the three digit docket number for the reporting plant. [ NOTE: if the LER is applicable to more than one unit, indicate all docket numbers in Field 6.]
7 NSSS VENDOR:
CODING INSTRUCTIONS:
For LERs marked "Y"in Field 2 or 3 above, enter the appropriate number depending on the vendor.
1 Westinghouse 2
Babcock and Wilcox 4
Combustion Engineering 8
ARCHITECT ENGINEER:
CODING INSTRUCTIONS:
For LERs marked "Y" in Field 2 or 3 above, enter the appropriate number (s) depending on the AE.
1 Bechtel 8
B&R 2
S&W 9
Other 3
Ebasco 4
S&L 5
UE&C 6
Gilbert
-7 Utility 3
4
LER fleader Listing for 28993007 http /scss.ornl. gov / scripts /scss/results/rLERDetti.cfm?lernrabr=28993007 1
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nummmmmme LER Header Listinefor LER Number: 28993007 Docket Number Year Report #
Rev#
DCS Number Event Date j
289 1993 007 0
9311120183 10/08/93 Nuclear Plant: Three Mile Island Unit #1 NRC Region:
1 Facility Operator: General Public Utilities Corp.
Architect Engineer: Gilbert Associates,!nc.
NSSS Vendor: Babcock & Wilcex Reactor Type: Pressurized Water Reactor Comm. Operations Date: 09/02/74 feC":Tl IT6EFNiiifM PirifiNiP'1 ABSTRACT POWER LEVEL - 000%. FAILURE OF PRESSURE ISOLATION VALVE (DH-V228) TO SEAT TIGIITLY DURING PLANT OPERATION DUE TO LOOSENESS OF THE HINGE MECIMNISM TMI-l was shutdown for the Cycle 10 Refueling Outage (10R) on October 8,1993 when it was discovered that a condition may have existed during power operation where Pressure Isolation Valve (PlV) DH-V22B in the Decay Heat Removal (DHR) System would exceed the Technical Specification (TS) surveillance leakage limits. This condition is reportable under 10 CFR 50.73(a)(2)(i)(B) as a condition or operation prohibited by the plant's TS. Repairs to DH-V22B were completed during 10R.
The hinge pins were found to have play (freedom of mcvement) such that the disk c id not consistently seat tightly. Following repairs the disk was verified to seat properly and when tested no leakage was observed. There were no safety consequences associated with this event. With the isolation valve (DH-V4B) open, leakage from the Reactor Coolant System (RCS) would be limited by the first check valve in series (CF-V5B). PlV leakage ?otential will be monitored during quarterly tests whe'.e the PIVs are not isolated. An evaluation wi i be performed if significant leakage is noted.
Unit Conditions at Time of Reportable Event:
Power Level: 0%
Operating Mode: Unknown Reportability Reason (s) for This LER:
10 CFR 50.73(a)(2)(i)- Shutdowns or Technical Specification Violations. [10]
Primary Cause(s) for This LER:
Equipment Failure [20 ]
Emergency Classification (s) for This LER:
There is no Emergency Classification for This LER 1 of 2 07/02/97 10:21:27
LERIIcader Listing for 28993007 http-1/scss.ornl. gov / scripts /scss/results/rLERDetil.cfm?lemmbr=28993007 L Referenced LERs: -
There are no Referenced LERs for This LER Contact Mike Poore at ORNL with guestions or comments concerning SCSS Web site content.
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Copyright C 1996 Oak Ridge National Laboratory (ORNL)
Last modified: January 31,1997 -
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- M LER Header Listingfor LER Number: 31888003 Docket Number Year Report #
Rev#
DCS Number Event Date 318 1988 003 0
8804220184 03/17/88 Nuclear Plant: Calvert Cliffs Unit #2 NRC Region:
1 4
Facility Operator: Baltimore Gas & Electric Co.
Architect Engineer: Bechtel Power Corp.
NSSS Vendor: Combustion Engineering Reactor Type: Pressurized Water Reactor Comm. Operations Date: 04/01/77 NEiaisiiiiiiiiil Fuiisirsiirs71 Engsi@l ABSTRACT POWER LEVEL - 000%. ON MARCH 17,1988. UNIT 2 WAS IN MODE 5, COLD SHUTDOWN, FOR A MAINTENANCE OUTAGE. Tile CHECK VALVE WillCH PROVIDES ISOLATION OF NO. 22 STEAM GENERATOR FROM NO. 21 STEAM GENERATOR IN THE EVENT OF A MAIN STEAM LINE BREAK UPSTREAM OF THE MAIN STEAM ISOLATION VALVES, WAS DISASSEMBLED FOR INSPECTION. UPON DISASSEMBLY, THE DISK WAS FOUND TO BE SEVERELY BENT, MAKING THE VALVE INCAPABLE OF PERFORMING ITS DESIGN FUNCTION. ANCllOR/ DARLING 6'- 900# TILTING DISK CHECK VALVE. Tile VALVE WAS REVERSE FLOW TESTED AND PLACED INTO SERVICE. THE FAILURE OF THE VALVE WAS A RESULT OF NORMAL WEAR OF THE HINGE PINS AND BUSHING AREA,IN CONJUNCTION WITH STEAM EROSION / CUTTING OF Tile DISK SEATING SURFACE.
LONG TERM PREVENTIVE ACTIONS WILL BE TO PERIODICALLY EITHER REVERSE-FLOW TEST THE VALVES OR TO DISASSEMBLE AND INSPECT THE VALVE INTERNALS. THE FREQUENCY OF THE TEST WOULD BE EVERY REFUELING OUTAGE, AND THE FREQUENCY OF INSPECTION WOULD BE TO INSPECT ONE VALVE EACH REFUELING OUTAGE. SIMILAR VALVES INSTALLED ON UNIT 1 WILL BE REVERSE FLOW TESTED DURING 'IHE APRIL 1988 UNIT 1 REFUELING OUTAGE. ANY CORIECTIVE ACTIONS FOR THE UNIT 1 VALVES WILL IlE BASED ON THE TEST RESULTS.
Unit Conditions at Time of Reportable Event:
Power Level: 0%
Operating Mode: Unknown Reportability Reason (s) for This LER:
10 CFR 50.73(a)(2)(ii)- Unanalyzed Conditions. (l1)
Primary Cause(s) for This LER:
Equipment Failure [20 ]
1of2 07/02/97 10:21:57
http /scss.ornl. gov / scripts /scss/results/rLERDett1.cfm?lernmbr=31888003 LERIIcader Listing for 31888003 1
Emergency Classification (s) for This LER:
~ There is no Emergency Classification for This LER
- Referenced LERs:
There are t.a Referenced LERs for This LER Contact Mike Poore at ORNL with uestions or comments concerning SCSS Web site content.
Contact Dale Yelldjag at NRC with quest ons concerning SCSS Web site access.
Copyright O 1996 Oak Ridge National Laboratory (ORNL)
Last modified: January 31,1997
'2ef2-07/02/97 10:21:57'
LER l{cader Listing for 32486017 http://scss.orn l. gov / scripts /scss/results/rL E RDetl l.c fm?lernm br=32486017 bM
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LE'R Header Listingfor LER Number: 32486017 Docket Number Year Report #
Rev#
DCS Number Event Date 324 1986 017 1
8707220498 06/18/86 Nuclear Plant: Brunswick Unit #2 NRC Region:
2 Facility Operator: Carolina Power & Light Co.
Architect Engineer: United Engineers & Construction NSSS Vendor: General Electric Reactor Type: Boiling Water Reactor Comm. Operations Date: 11/03/75 liiiiiiHiiiiiiiiEl FEdiisiiiP117iiiiMiiiM1 ABSTRACT POWER LEVEL - 055%. AT 0811 ON 6/18/86, WITH UNIT 2 AT 55% POWER, A REACTOR SCRAM OCCURRED DUE TO LOW LEVEL (LL) NO 1 WHILE PLACING REACTOR MAIN STEAM-DRIVEN FEED PUMP (RFP) 2B INTO SERVICE. IMMEDIATELY AFTERTHE F CRAM, TIIE CONTROL OPERATOR CLOSED THE RFP 2B DISCHARGE VALVE ALLOWING hFP 2A TO RESTORE NORMAL LEVEL CONTROL. A SCRAM RECOVERY WAS CARRIED OUT PRIMARY CONTAINMENT GROUPS 2,6, AND 8 ISOLATIONS OCCURRED DUE TO THE LL NO.1. REACTOR LEVEL WAS CONTROLLED WITH THE REACTOR FEEDWATER SYSTEM. REACTOR PRESSURE DECREASED TO 870 PSIG. THE LL NO.1 RESULTED FROM REVERSE FLOW OF REACTOR FEEDWATER THROUGH RFP 2B DISCHARGE CllECK VALVE FW.V2 PRIOR TO STARTING THE PUMP. A WORN VALVE DISC PIVOT PIN LOCKING STUD IN V2 ALLOWED ONE OF Tile TWO PIVOT PINS TO LOOSEN AND CAUSE Tile DISC TO BECOME IMPROPERLY SEATED. V2, CRANE VALVE COMPANY MODEL LIST 973A, WAS DISASSEMBLED AND THE DISC PIVOT PINS, PlVOT PIN BUSHINGS, AND PlVOT PIN LOCKING STUDS WERE REPLACED. FW-VI, THE RFP 2A DISCHARGE CHECK VALVE, WAS OVERHAULED DURING THE RECENT UNIT REFUELING / MAINTENANCE OUTAGE. APPROPRIATE PROCEDURES IIAVE BEEN REVISED TO MINIMIZE THE POTENTIAL FOR A DEFECTIVE CilECK VALVE TO CAUSE A LEVEL TRANSIENT WHILE PLACING A REACTOR FEED PUMP IN SERVICE.
Unit Conditions at Time of Reportable Event:
Power Level: 55%
Operating Mode: Unknown Reportability Reason (s) for This LER:
10 CFR 50.73(a)(2)(iv)- ESF Actuations. [13]
Primary Cause(s) for This LER:
There is no Primary Cause for This LER Iof2' 07/02/97 10:22:31
http /scss.orn). gov / scripts /scss/results/rLERDetll.cfm71ermnbr=32486017 LER lleader Listing for 32486017.!
1 Emergency Classification (s) for This LER:
There is no Emergency Classification for This LER Referenced LERs:
There are no Referenced LERs for This LER Contact Mike Poors at ORNL with questions or comments concerning SCSS Web site content.
Contact Dale Yell <lia! at NRC with questions concerning SCSS Web site access.
Copyright O 1996 Oak Ridge National Laboratory (ORNL).
Last modified: January 31,1997
-I i
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07/02/97 10:22:31 2Cf2 1
LER licader Listing for 33390012 http://scss.ornl. gov / scripts /scss/results/rLERDett i.c fm71ernmbr=33390012 N
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Rev #
DCS Number Event Date 333 1990 012 2
9303110290 04/04/90 Nuclear Plant: Fitzpatrick Unit #1 NRC Region:
1 Facility Operator: Power Authority of the State of NY Architect Engineer: Stone & Webster Engr., Corp.
NSSS Vendor: General Electric Reactor Type: Boiling Water Reactor Comm. Operations Date: 07/28/75
[WpYsiiqiO ifesseseiisN E%iNiiiiM1 ABSTRACT POWER LEVEL - 000%. UPDATED REPORT - PREVIOUS REPORT DATE 8/29/90 EIIS Codes are in [] During the 1990 Refuel Outage,61 check valves in the Normal and Emergency Service Water (ESW) [KG, BI] were opened and visually inspected. 37 valves were from the ASME Section XI IST Program and 24 were from the Preventive Maintenance (PM) Program. declared inoperable dse to failing the inspection criteria. The IST check valves were shown to be operable by actual flow test or calculation. Other efforts included internal inspection of 500 feet of piping and 10 safety-related coolers and Air llandling Units (AHUs). 2 AHUs were found to have 25 percent tube plugging with silt / sand, but were shown able to remove design basis heat load. Of the 500 feet of piping,200 feet were found 10 to 30 percent restricted in cross-sectional area, but a calculation demonstrated that flow control valves were hydraulically limiting. Except for the potential loss of control room / relay room
[NA] cooling, ESW was considered capable of performing the design safety function. The valves, coolers, and piping were cleaned / replaced as necessary and returned to service. Intake bays were also cleaned. Periodic flushing / performance testing prevents recurrence. LERs88-005,88009, and 89-015 are related.
Unit Conditions at Time of Reportable Event:
Power Level: 0%
Operating Mode: Unknown Reportability Reason (s) for This LER:
10 CFR 50.73(a)(2)(ii)- Unanalyzed Conditions. [11]
Primary Cause(s) for This LER:
. Equipment Failure [20 ]
Emergency Classification (s) for This LER:
There is no Emergency Classification for This LER Iof2 07/02/97 10:23:08
f LER lleader Listing for 33390012 http://scss.ornl. gov / scripts /scss/resul drLERDetil.cfm71erambr=33390012 t
i-i Referenced LERs:
LER Event Number ;
Reactor UnitName Date LER Title Failure of Ventilation Backup Cooling Water 33388005l Fitzpatrick Unit #1 05/25/88 i
Supply Check Valves P tentialInadequate Cooling of ECCS Due to 33388009 Fitzpatrick Unit #1 10/21/88 Procedure inadequacies Nine Air Operated Containment Isolation Valves Exhibit Operaticaal Deficiencies Due to Packing 33389015 Fitzpatrick Unit f1 09/18/89 Problems & Iron Build-Up in Reactor Building Closed Loop Cooling System Contact Mike Poore at ORNL with questions or comments concerning SCSS Web site content.
Contact Dalt.ltilding at NRC with questions concerning SCSS Web site access, Copyright C 1996 Oak Ridge National Laboratory (ORNL)
Last modified: January 31,1997 2I 07/02/97 10:23:09
LER IIcader Listing for 37385052 http://scss.ornl. gov / scripts /scss/results/rLE RDetl l.c fm71ern mbr=37385052 h Mggigitp m
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ammmmmmmmmmmmmmusammmu ammme LER Header Listingfor LER Number: 37385052 Docket Number Year Report #
Rev#
DCS Number Event Date 373 1985 052 0
8508070484 06/29/85 Nuclear Plant: La Salle Unit #1 NRC Region:
3 Facility Operator: Commonwealth Edison Co.
Architect Engineer: Sargent & Lundy NSSS Vendor: General Electric Reactor Type: Boiling Water Reactor Comm. Operations Date: 01/01/84 V.l1.~.n? MMONilP1 PEPMilMEl ABSTRACT POWER LEVEL - 092%. ON 6-29-85 AT 0815 HRS, OPERATIONS STARTED THE 'A' CONTROL ROD DRIVE PUMP AND THEN SHUTDOWN THE 'B' CONTROL ROD DRIVE PUMP IN PREPARATION FOR ROUTINE MAINTENANCE. THE 'A' CRD PUMP TRIPPED ON LOW SUCTION PRESSURE WHEN THE 'B' CRD PUMP WAS SHUTDOWN. THE 'A' CRD PUMP WAS SUBSEQUENTLY RESTARTED SEVERAL TIMES BUT TRIPPED EACH TIME ON LOW SUCTION PRESSURE. Tile 'B' CRD PUMP WAS RESTARTED BUT WAS UNABLE TO DELIVER ENOUGH FLOW TO REPRESSURIZE THE CRD HYDRAULIC CONTROL UNIT ACCUMULATORS PRIOR TO THE ACTUATION OF MORE THAN ONE CRD ACCUMULATOR LOW PRESSURE ALARM. IN ACCORDANCE WITH TECH SPEC 3.1.3.5 ACTION A.2.A THE UNIT 1 MODE SWITCH WAS PLACED IN THE SHUTDOWN POSITION.
TIIE UNIT RESPONDED AS EXPECTED AND WAS TAKEN TO COLD SHUTDOWN WITHOUT FURTHER INCIDENT. INVESTIGATION REVEALED THAT THE 'B' CRD PUMP DISCHARGE STOP CHECK VALVE FAILED TO CLOSE WHICH ALLOWED EXCESS FLOW WHEN THE 'A' PUMP WAS STARTED (FLOW VENT BACKWARDS THROUGH THE 'B' PUMP). TIIE EXCESS FLOW THROUGH THE 'A' PUMP CREATED A LOW SUCTION PRESSURE ON THE 'A' PUMP WHICH CAUSED THE PUMP TO TRIP. THE CONSEQUENCES OF THIS EVENT WERE MINIMAL BECAUSE ALL RODS WENT FULL IN UPON INITIATION OF THE SCRAM. THE 'A' AND 'B' DISCHARGE CHECK VALVES WERE INSPECTED AND REPAIRED.
Unit Conditions at Time of Reportable Event:
Power Level: 92%
Operating Mode: Unknowe i
Reportability Reason (s) for This LER:
10 CFR 50.73(a)(2)(i)- Shutdowns or Technical Specification Violations. [10]
Primary Cause(s) for This LER:
- Iof2 07/02/97 10:23:51
LERllender Listing for 37385052' http://scss.ornl. gov / scripts /scss/rssults/rL ERDeti l.c fm?lernmbr=37385052 There is no Primary Cause for This LER Emergency Classification (s) for This LER:
- There is no Emergency Classification for This LER Referenced LERs:
There are no Referenced LERs for This LER Contact Mike Poore at ORNL with questions or comments concerning SCSS Web site content.'
Contact Dale YeMia= at NRC with questions concerning SCSS Web site access.
Copyright C 1996 Dak Ridge National Laboratory (ORNL)
Last modified: January 31,1997 1
2 of 2 07/02/97 10:23:51
LER il:ader Listing for 40986038 http://scss.or nl. gov / scripts /scss/results/rLERDett i.cfm?lernmbr40986038
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LER Header Listinefor LER Number: 40986038 Docket Number Year Report #
Rev#
DCS Number Event Date 409 1986 038 0
8701130311 12/16/86 Nuclear Plant: Lacrosse Unit #1 NRC Region:
3 Facility Operator: Dairyland Power Cooperative Architect Engineer: Sargent & Lundy NSSS Vendor: Allis Chalmers Reactor Type: Boiling Water Reactor Comm. Operations Date: 11/01/69
!NiWiiilliff ViiisilisiiikTH IEMiniiii&M ABSTRACT POWER LEVEL - 060%. ON DECEMBER 16,1986, THE CONCENTRATION IN THE BORON TANK WAS ANALYZED TO BE 17% BY WElGHT WHICH WAS BELOW THE TECHNICAL SPECIFICATION MINIMUM REQUIREMENT OF 17.8%. BORIC ACID AND BORAX WERE ADDED TO THE TANK AND THE SODIUM PENTABORATE DECAHYDRATE CONCENTRATION INCREASED ABOVE THE REQUIRED MINIMUM VALUE. THE TANK CONCENTRATION HAD BEEN DILUTED BY WATER FROM THE OVERHEAD STORAGE TANK. FOLLOWING PERFORMANCE OF THE STARTUP TEST OF BORON AND EMERGENCY CORE SPRAY CONTROLS AND VALVES ON DECEMBER 13, THE TANK LEVEL HAD STARTED SLOWLY INCREASING. THE WATER WAS DETERMINED TO BE LEAKING BACK THROUGH THE BORON INJECTION OUTLET CHECK VALVE,60-26-001 AND CORE SPRAY PUMP BORON SOLUTION INLET VALVE,60-25-005, INTO THE BORON TANK. VALVE 60-25-005 WAS MANUALLY ISOLATED, CYCLED, AND FLUSHED. THE LEAKAGE THEN STOPPED. A STEP WILL BE ADDED TO THE TEST PROCEDURE REQUIRING TilAT THE CORE SPRAY PUMP BORON SOLUTION INLET VALVES BE CHECKED FOR LEAKAGE FOLLOWING CYCLING. ALSO, THE BORON CHECK VALVE IS SCilEDULED TO BE DISSASSEMBLED NEXT REFUELING OUTAGE. THE ONE PREVIOUS REPORT INVOLVING LOW BORON TANK SOLUTION CONCENTRATION WAS RO 76-17.
Unit Conditions at Time of Reportable Event:
Power Level: 60%
Operating Mode: Unknown Reportability Reason (s) for This LER:
10 CFR 50.73(a)(2)(i)- Shutdowns or Technical Specification Violations. [10]
Primary Cause(s) for This LER:
There is no Primary Cause for This LER 1of2 07/02/97 10:24:19
http /scss.ornl. gov / scripts /scss/results/rLERDetll.cfm71ernmbr40986038
/
'l ER Header Listing for 40986038,-
Emergency Classification (s) for This LER:.
There is no Emergency Classification for This LER Referenced LERs:
.'lhere are no Referenced LERs for This LER i
Contact Mike Poore at ORNL with guestions or comments concerning SCSS Web site content.
Contact Dale Yellding at NRC with questens concerning SCSS Web site access.
Copyright C 1996 Oak Ridge National Laboratory (ORNL) i Last modified: January 31,1997 l
4 1
I 2CI2 07/02/97 10:24:20-1
LER Header Listing for 48385039 h ttp>//sc ss.ornl. gov / scripts /scss/results/rL E RDetl l.c fm?lernm br=48385039
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maammmmmmmmmmmmmmme mammmmmmme LER Header Listingfor LER Number: 48385039 Docket Number Year Report #
Rev #
DCS Number Event Date 483 1985 039 0
8509260002 08/20/35 Nuclear Plant: Callaway Unit #1 NRC Region:
3 Facility Operator: Union Electric Co.
Architect Engineer: Bechtel Power Corp.
NSSS Vendor: Westinghouse Reactor Type: Pressurized Water Reactor i
Comm. Operations Date: 12/19/84 IH!iiBiiiiEiiFl FiGiiiiiviiiiiM F4Rniif91 i
ABSTRACT POWER LEVEL - 025%. ON 8/20/85 AT APPROXIMATELY 2140 CDT A REACTOR TRIP OCCURRED DURING UNIT STARTUP. A TURBINE TRIP OCCURRED AS THE RESULT OF MOISTURE SEPARATOR REllEATER (MSR)'B' HIGli LEVEL SIGNAL DUE TO Tile LOSS OF Tile DRAIN PATli FROM IIEATER DRAIN TANK 'B.' TIIE NORMAL DRAIN PATil WAS LOST DUE TO CilECK VALVE AFV-160 WillCll HAD STUCK CLOSED. TIIE ALTERNATE PATH WAS LOST DUE TO A VALVE WillCH IIAD NOT BEEN PROPERLY REPOSITIONED FOR NORMAL OPERATION AFTER MAINTENANCE. TILE TRANSIENT CAUSED BY Tile TURBINE TRIP RESULTED IN A STEAM GENERATOR 'C' LO-LO LEVEL SIGNAL WIIICil CAUSED A REACTOR TRIP, FEEDWATER ISOLATION, AUXILIARY FEEDWATER ACTUATION AND A STEAM GENERATOR BLOWDOWN AND SAMPLE ISOLATION. ALL SAFETY COMPONENTS ACTUATED PER DESIGN. TIIE OPERATORS STABILIZED TIIE PLANT, DETERMINED Tile CAUSE OF TILE TRIP AND COMMENCED A RECOVERY IN ACCORDANCE WITH PLANT PROCEDURES. COR.RECTIVE ACTION FOR tills EVENT INCLUDED A VERIFICATION OF CORRECT VALVE POSITIONS FOR THE MSR DRAINS AND FREEING OF CilECK VALVE AFV-160. TO PREVENT RECURRENCE, A LETTER WAS ISSUED TO TIIE SillFT SUPERVISORS CONCERNING TilEIR RESPONSIBILITIES FOR CONTROL OF WORK ACTIVITIES.
Unit Conditions at Time of Reportable Event:
Power Level: 25%
Operating Mode: Unknown Reportability Reason (s) for This LER:
10 CFR 50.73(a)(2)(iv)- ESF Actuations. [13]
Primary Cause(s) for This LER:
There is no Primary Cause for This LER 1 of 2 07/02/97 10:24:52
' LER Header Listing for 48385039 http://scss.ornl. gov / scripts /scss/results/rLERDeti 1.c fm71rmmbr=48385039 Emergency Classification (s) for This LER:
i There is no Emergency Classification for This LER Referenced LERs:
LER Event Number Reactor Unit Name Date LER Title 48385038 Callaway Unit #i j 08/20/85 The title for this LER. is not currently available.
Contact Mike Poore at ORNL with guestions or comments concerning SCSS Web site content.
Contact Dalt.Yellding at NRC with questions concerning SCSS Web site access.
j Copyright C 1996 Oak Ridge National Laboratory (ORNL)
Last modified: January 31,1997 l
i l
2 of 2 07/02/97 10:24:52
I AEOD Annual Report,1989 AEOD IIT Action Tracking System 1
Action Source: IIT Report on San Onofre Nuclear Generating Station (SONGS) Unit 1 Event of November 21,1985 (Reference 1)
Note:
Following the loss-of-power and water-hammer event that occurred on November 21,1985, the NRC developed an action plan for the review of licensee actions required as a result of the event and the subseq'uent IIT report. This plan involved actions by Region V, NRR, and IE. De NRC concerns and the licensee commitments were largely addressed before the startup of Unit 1 (July 15,1986). This effort was documented in Inspection Report No. 50-206/86-34 (Reference 2).
Item 1: Adequacy of feedwater check valves to perform safety function.
Actions: Implement and coordinate the staff and industry actions necessary to ensure the reliability of safety-related check valves. (Responsible Office: IE; other offices to assist as requested)
Arco.,. be evaluated include Ucensee's engineering report on root cause analysis and proposed corrective actions e
Adequacy of check valve design for this application e
Adequacy of Inservice Testing (IST) Program and procedures to detect degraded and failed valves e
Adequacy of check valves (and related testing programs)in other systems such as the residual e
heat removal (RHR) system Disposition: Resolved.
Reference 3 documents the IE review of plant-specific aspects of the event in response to item 1 and concludes that plant-specific check valve problems have been adequately resolved. The generic ade-quacy of check valve designs and IST programs was coordinated with utility owners groups and the In-stitute of Nuclear Power Operations (INPO), and the results were disseminated to the industry.This activity is documented in References 4 and 5. Concurrent with this industry activity, the NRC staff conducted in-depth plant inspections to assess check valves in accordance with NRC Inspection Man-ual Inspection Procedure 73756, "Insers..a Testing of Pumps and Valves," dated March 16,1987, These inspections found that licensee programs to address check valves ranged from little attention to reasonably good attention. In April 1989, the Nuclear Industry Check Valve Group was formed, nis group plans to develop specific guidelines for plant actions pertaining to database information on check valve performance, non-intrusive examination procedures to assess check valve performance, and maintenance procedures. In addition, the NRC recently issued a generic letter (Reference 6),
which covers certain aspects of IST requirements for check valves.
Item 2: Completeness of resolved Unresolved Sa ety Issue (USI) A-1, " Water llammer."
r Actions: Assess the need to reevaluate USI A-1 to address specifically the potential for and prevention of con-densation-induced water hammers in feedwater piping (assuming the issue concerning check valve in-tegrity will be resolved in Item 1). (Responsible Office: NRR)
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NUREG-1272, Appendix F 10
Reactors-Staff Actions AEOD IIT Action Tracking System Item 2 '(cont.) -
Disposition: Resolved.
Reference 7 documents the results of the NRR reassessment of USI A-1 in response to Item 2 and concludes that the event at San Onofre was due to grossly failed check valves in the feedwater system and a basis did not exist for reopening the water-hammer safety issue. Additionally, resolution of USI A-1 recognized that water-hammer events would continue to occur and that such events would not produce unacceptably high contributions to core-melt frequency or public risk. The imposition of new requirements to reduce water-hammer events is not supportab!c by current cost-benefit guidelines and, thus, no further action is warranted at this time.
Item 3: Adequacy of San Onofre Unit I design.
Actions: Implement and coordinate the staff's actions to reevaluate the following San Onofre design features.
(Responsible Office: NRR)
Manual loading of the dicsci generators following a loss-of-power event Manual actuation of steamline isolation valves and assurance of steam generator availability to remove decay heat 12ck of steam generator blowdown status in control room Adequacy of the licensee's design change to eliminate spurious safety injection (SI) indication upon loss of power Dispsition: Resolved.
Reference 8 documents NRR's resolution of design issues in response to Item 3. Reference 9 docu-ments the remuuan of the last outstridmg item ide.itified in Reference 8 and, thus, all plant specific design issues have been resolved.
Item 4: Adequacy of post trip review.
Actions: (a) Evaluate NRC requirements for ensuring that sufficient event data are retrievable to reconstruct accurately the event following a loss of offsite power. (Responsible Office: NRR)
(b) Evaluate the licensee's process for post-trip review and evaluation, including the thoroughness of the review and oversight provided by the onsite and offsite nuclear safety review groups. (Respon-sible Office: Region V (RV))
Disposition: Resolved.
Reference 10 documents the NRR evaluation in response to Item 4a and concludes that this issue was adequately addressed as part of item 1.2 of NRC Generic letter 83 28 (" Required Actions Based on Generic implications of Salem ATWS Events," issued July 8,1983).
Reference 11 documents the RV assessment of the adequacy of the licensee's safety review process in response to item 4b and concludes that the licensee's process for post-trip review and evaluation meets regulatory iequirements.
I1 NUREG-1272, Appendix F
~AEOD Annual Report,1989 AEOD IIT Action Tracking System Item 5: Adequacy of licensee's recordkeeping practices.
Actions: Evaluate the adequacy of the licensce's maintenance records. (Responsible Office: RV)
Disposition: Resolved.
Reference 12 documents the RV assessment of the adequacy of the licensee's maintenance records in response to item 5 and concludes that the licensee's recordkeeping practices were satisfactory.
Item 6: Adequacy of operator training and/or procedures.
~ Actions: Review the implementation of the training program regarding operator understanding and actions in j
the areas of electrical systems and invoking Technical Specification action statements. (Responsible I
Office: RV)
)
i Disposition: Resobed.
Reference 11 documents the RV assessment of the licensee's training program in response to IOm 6 and concludes that operator training and/or procedures were edequate.
Item 7: Adequacy of emergency notifications and NRC res:*onse.
Actions: (a) Verify the adequacy of the licensce's procedures and trcining for reporting of events to the NRC 1
Operations Center. (Responsible Office: RV)
(b) Evaluate the need for changes in NRC policy or guidance regarding the use of the emergency notification system (ENS)line, the use of NRC personnel as ENS communicators, and possible approaches to improve the ability to determine the overall plant status. (Responsible Office:
AEOD)
Disposition: Ongoing.
Reference 13 documents the RV assessenent of the adequacy of the licensee's reporting of events to 19 NRC in response to item 7a and concludes that it was adequate.
References 14 and 15 document the IE evaluation in response to Item 7b and conclude that the NRC policy on the use of the ENS was clear.To ensure that the NRC policy regarding the use of NRC per-sonnel as ENS communicators would be followed, action was taken, as documented in Reference 15, to communicate NRC's policy and guidance regarding the resident inspector's role during licensee events to all regions. Actions that were taken to improve the ability to determine the overall plant status included development of the Emergency Response Data System (ERDS), increased emphasis on site-specific training for NRC Headquarters Operations Officers (HOOs), and development of various site-specific information systems for use. by the HOOs. These actions, as documented in References 15 and 16, are in various stages of implementation and/or are being upgraded, and with the exception of the ERDS, are expected to be fully implemented within a year. The ERDS is expected to be fully op-erational in about 3 years.
Item 8: Significance of backlog of license amendments.
Actions: Evaluate whether a backlog of license amendments and Technical Specification changes contributed to delays in approving the licensee's 151' program. (Responsible Office: NRR)
' NUREG-1272, Appendix F 12
Reactors-Staff Actions AEOD IIT Action Tracking System item 8 (cont.)
Disposillom Resolved.
Reference 17 documents the NRR evaluation of the backlog of licensee amendments in respor ie to item 8 and concludes that the backlog of license amendments did not delay approval of the licensee's IST program.
References:
1.
NUREG-1190, " Loss of Power and Water Hammer Event at San Onofre, Unit 1, on Novem-ber 21,1985," dated January 1986.
.2.
NRC Inspection Report No. 50-206/86 34, dated July 28,1986.
s 3.
Memorandum from J. Taylor to J. Martin (NRC), " Staff Actions Resulting From the Novem-ber 21,1985 San Onofre Unit 1 Water Hammer Event," dated June 6,1986.
"N 4.
Institute of Nuclear Power Operations (INPO), Significant Operating Experience Report (SOER) 86-03, " Check Valve Failures or Degradations," dated October 1986. (Proprietary)
\\ 5.
Electric Power Research Institute (EPRI), Report NP-5479," Check Valve Application Guide-lines," dated January 1988.
\\6. NRC Generic Ixtter 89-04, " Guidance on Developing Acceptable Inservice Testing Programs,"
. 3-dated April 3,1989.
7.
Memorandum from H. Denton to V. Stello, Jr. (NRC), " Review and Assessment of Waterham.
g
-b mer Occurrences Since CY 1981," dated July 7,1986.
- k 8.
letter from T. M. Novak (NRC) to K. P. Baskin (Southern California Edison Company),"Techni-cal Evaluation of Issues From the NRC Staff Action Plan in Response to the November 21,1985 Event at San Onofre Unit 1," dated November 12,1986.
9.
Ixtter from C. M. Trammell (NRC) to H. B. Ray (Southern California Edison Company), "long-Term Electrical Cable Monitoring Program-San Onofre Nuclear Generating Station, Unit 1 (TAC 73782)," dated December 8,1989.
- 10. Memorandum from H. Denton to V. Stello, Jr. (NRC)," Post Trip Data Following Imss of Offsite Power," dated August 4,1986.
- 11. NRC Inspection Report No. 50s206/86-20, dated June 9,1986.
- 12. NRC Inspection Report No. 50-206/86-22, dated June 9,1986.
- 13. NRC Inspection Report No. 50-206/86-16, dated June 2,1986,
- 14. Memorandum from K. Perkins to G. Zech (NRC), " Response to 01/04/86 Memo From Stello to Denten on the 11/21/86 San Onofre Investigation," dated February 10,1986.
- 15. Memorandum from G. Grimes to D. Kirsch (NRC),"IRB Response to San Onofre 1 IIT Action item List," dated June 11,1986.
- 16. Memorandum from K. Perkins to G. Zech (NRC),"IRB Response to San Onofre ! IIT Action Item List " dated February 13,1986.~
13 '
NUREG-1272, Appendix F w
s e
1
. AEOD Annual Rcport,1989 AEOD IIT Action Tracking System i
References (cont.)
- 17. Memorandum from H. Denton to V. Stello, Jr. (NRC), " San Onofre Nuclear Generating Station, Unit 1 (SONGS-1) Inservice Testing (IST) Review," dated April 28,1986.
i l
NURIIG-1272, Appendix F 14