/01T-0 on 780916:3 MSIVs(NS03A,NS04A,NS04B) Had Leakage Rates in Excess of Maximum Allowable Due to Stem Packing,Cracked Stellite Weld & Failure of Main Poppet to Seat Properly.Msiv Failures Being Investigated| ML20064B739 |
| Person / Time |
|---|
| Site: |
Oyster Creek |
|---|
| Issue date: |
09/29/1978 |
|---|
| From: |
Ross D
JERSEY CENTRAL POWER & LIGHT CO. |
|---|
| To: |
|
|---|
| Shared Package |
|---|
| ML20064B738 |
List: |
|---|
| References |
|---|
| LER-78-018-01T, LER-78-18-1T, NUDOCS 7810100075 |
| Download: ML20064B739 (4) |
|
|
|---|
Category:LICENSEE EVENT REPORT (SEE ALSO AO RO)
MONTHYEAR05000219/LER-1990-0081990-07-23023 July 1990
LER 90-008-00:on 900625,reactor Scram Occurred on Low Condenser Vacuum.Caused by Failure of Backwash Outlet Valve to Open,Causing Loss of Circulating Water to Condenser A. Plant Stabilized & Taken to Cold shutdown.W/900723 Ltr
05000219/LER-1984-020, Errata to LER 84-020-01,consisting of Page 4 Re Corrective Actions.Preventive Maint Request Initiated to Store Overload Devices in Vertical Position1985-09-23023 September 1985
Errata to LER 84-020-01,consisting of Page 4 Re Corrective Actions.Preventive Maint Request Initiated to Store Overload Devices in Vertical Position
05000219/LER-1982-0261982-06-0707 June 1982
LER 82-026/03L-0:on 820507,thermocouple for Safety Valve NR28J Found Open During Surveillance Testing.Caused by Broken Lead Wire at Thermocouple.One of Adjacent Valve Acoustic Monitor Setpoints Reduced.Thermocouple Replaced
05000219/LER-1981-061, Updated LER 81-061/03X-1:on 811223,of 11 Deficient Hydraulic Snubbers,Nine Found to Be Inoperable During Functional Testing.Caused by Seal Degradation &/Or Worn Poppet Valves. Snubbers Replaced W/Operable Spares1982-05-25025 May 1982
Updated LER 81-061/03X-1:on 811223,of 11 Deficient Hydraulic Snubbers,Nine Found to Be Inoperable During Functional Testing.Caused by Seal Degradation &/Or Worn Poppet Valves. Snubbers Replaced W/Operable Spares
ML20064J2011978-12-22022 December 1978
/03L-0 on 781126:during Testing Containment Spray Pumps 51C & 51A Failed to Start on Signal Due to Excess Friction in Circuit Breaker Trip Bar Bearings.Bearings Were Cleaned & Lubricated
ML20147H5331978-12-21021 December 1978
/03L-0 on 781125:switch RE03C Was Found to Trip at Pressure Less conservative(1075 Psig)Than Specified in Tech Specs 2.3.3.Cause of Pressure Switch Tripping Is Attributed to Sensor Repeatability
ML20064H3441978-12-14014 December 1978
/01T-0 on 781129:during Test,Noted That Containment Leakage Rate Greater than Allowable.Discovered That Vent Valves Were Leaking.Caused by Leakage Past the Seating Surfaces
ML20064H1871978-12-13013 December 1978
/01T-0 on 781129:during Test,Main Steam Line Drain Valve V-1-106 Did Not Close 2 Incore Probes,Did Not Fully W/Draw & Reactor Coolant Valve Did Not Indicate Closed.Due to Jammed Armature,Faulty Fuse Holder & Incorrect Settings
ML20064F0691978-11-16016 November 1978
/03L-0 on 781018:isolation Valve V-7-31 Had Damaged Rubber Seat & Drain Valve V-7-29 Obstructed by Piece of Cork & Filter F-1-58 Medium Disintegrated.Cause of V-7-31 & F-7-58 Degradations Due to Detonations in Sys
ML20064E7071978-11-15015 November 1978
/03L-0 on 781019:IRM 12 Became Inoperative While Cable to IRM 14 Was Disconnected.Irm 12 Failure Attributed to Maint Activities Under Vessel.Subesquent Inspec Showed IRM 12 Cable Damaged
ML20064E5561978-11-15015 November 1978
/03L-0 on 781017:during Routine Surveillance Test of Standby Gas Treatment Sys,The Sys II Discharge Valve Failed to Open,Due to Stuck Solenoid in Control Air Supply to Operator for V-28-30
ML20064C2501978-10-12012 October 1978
/03L-0:on 780914,circuit Breaker for motor- Operated Valve Tripped When Valve Closed by Isolation Signal.Caused by Component Failure Allowing Switch Setting to Shift.Setting Corrected & Set Screw Tightened
ML20064B7391978-09-29029 September 1978
/01T-0 on 780916:3 MSIVs(NS03A,NS04A,NS04B) Had Leakage Rates in Excess of Maximum Allowable Due to Stem Packing,Cracked Stellite Weld & Failure of Main Poppet to Seat Properly.Msiv Failures Being Investigated
1990-07-23
[Table view]
Category:TEXT-SAFETY REPORT
MONTHYEARML20217K4451999-09-30030 September 1999
Monthly Operating Rept for Sept 1999 for Oyster Creek Nuclear Generating Station.With
ML20211P6731999-08-31031 August 1999
Monthly Operating Rept for Aug 1999 for Oyster Creek Nuclear Generating Station.With
ML20211A7051999-07-31031 July 1999
Monthly Operating Rept for July 1999 for Oyster Creek Nuclear Station.With
ML20209G0631999-06-30030 June 1999
Monthly Operating Rept for June 1999 for Oyster Creek Nuclear Generating Station.With
ML20212H5491999-06-18018 June 1999
Non-proprietary Rev 4 to HI-981983, Licensing Rept for Storage Capacity Expansion of Oyster Creek Spent Fuel Pool
ML20195E7961999-05-31031 May 1999
Monthly Operating Rept for May 1999 for Oyster Creek Nuclear Generating Station.With
ML20206N7431999-04-30030 April 1999
Monthly Operating Rept for Apr 1999 for Oyster Creek Nuclear Generating Station.With
ML20205P5401999-03-31031 March 1999
Monthly Operating Rept for Mar 1999 for Oyster Creek Nuclear Generating Station.With
ML20204C8201999-02-28028 February 1999
Monthly Operating Rept for Feb 1999 for Oyster Creek Nuclear Generating Station.With
ML20199E4671998-12-31031 December 1998
Monthly Operating Rept for Dec 1998 for Oyster Creek Nuclear Generating Station.With
ML20195E8321998-12-31031 December 1998
10CFR50.59(b) Rept of Changes to Oyster Creek Sys & Procedures, for Period of June 1997 to Dec 1998.With
ML20198D2091998-11-30030 November 1998
Monthly Operating Rept for Nov 1998 for Oyster Creek Nuclear Generating Station.With
ML20195J8591998-11-12012 November 1998
Rev 11 to 1000-PLN-7200.01, Gpu Nuclear Operational QA Plan
ML20195C4271998-11-0606 November 1998
Safety Evaluation Supporting Proposed Ocnpp Mod to Install Core Support Plate Wedges to Structurally Replace Lateral Resistance Provided by Rim Hold Down Bolts for One Operating Cycle
ML20155J3021998-10-31031 October 1998
Monthly Operating Rept for Oct 1998 for Oyster Creek Nuclear Generating Station.With
ML20154R4981998-10-20020 October 1998
Core Spray Sys Insp Program - 17R
ML20154L3051998-10-14014 October 1998
Safety Evaluation Accepting Licensee Request to Defer Insp of 79 Welds from One Fuel Cycle at 17R Outage
ML20154Q3371998-09-30030 September 1998
Rev 8 to Oyster Creek Cycle 17,COLR
ML20154L5571998-09-30030 September 1998
Monthly Operating Rept for Sept 1998 for Oyster Creek Nuclear Generating Station.With
ML20151V6311998-08-31031 August 1998
Monthly Operating Rept for Aug 1998 for Oyster Creek Nuclear Generating Station.With
ML20237D5691998-08-31031 August 1998
Rev 0 to MPR-1957, Design Submittal for Oyster Creek Core Plate Wedge Modification
ML20237D5711998-08-18018 August 1998
Rev 0 to SE-000222-002, Core Plate Wedge Installation
ML20237B0131998-07-31031 July 1998
Monthly Operating Rept for July 1998 for Oyster Creek Nuclear Generating Station
ML20236R0511998-06-30030 June 1998
Monthly Operating Rept for June 1998 for Oyster Creek Nuclear Generating Station
ML20249B2981998-05-31031 May 1998
Monthly Operating Rept for May 1998 for Oyster Creek Nuclear Station
ML20248F3531998-05-21021 May 1998
Part 21 Rept Re Electronic Equipment Repaired or Reworked by Integrated Resources,Inc from Approx 930101-980501.Caused by 1 Capacitor in Each Unit Being Installed W/Reverse Polarity. Policy of Second Checking All Capacitors Is Being Adopted
ML20247F1891998-05-0505 May 1998
Risk Evaluation of Post-LOCA Containment Overpressure Request
ML20247G0581998-04-30030 April 1998
Monthly Operating Rept for Apr 1998 for Oyster Creek Nuclear Generating Station
ML20216K0341998-03-31031 March 1998
Monthly Operating Rept for Mar 1998 for Oyster Creek Nuclear Generating Station
ML20151Y4651998-03-31031 March 1998
Non-proprietary Version of Rev 1 to GENE-E21-00143, ECCS Suction Strainer Hydraulic Sizing Rept
ML20217A4631998-03-23023 March 1998
Safety Evaluation Accepting Use of Three Heats/Lots of Hot Rolled XM-19 Matl in Core Shroud Repair Assemblies Re Licenses DPR-16 & DPR-59,respectively
ML20212E2291998-03-0404 March 1998
Rev 11 to 1000-PLN-7200,01, Gpu Nuclear Operational QAP, Consisting of Revised Pages & Pages for Which Pagination Affected
ML20216J0841998-02-28028 February 1998
Monthly Operating Rept for Feb 1998 for Oyster Creek Nuclear Generating Station
ML20203B2781998-02-16016 February 1998
10CFR50.59(b) Rept of Changes to Oyster Creek Systems & Procedures
ML20203A3801998-01-31031 January 1998
Monthly Operating Rept for Jan 1998 for Oyster Creek Nuclear Generation Station
ML20198P1791997-12-31031 December 1997
Monthly Operating Rept for Dec 1997 for Oyster Creek Nuclear Station
ML20217C7591997-12-31031 December 1997
1997 Annual Environmental Operating Rept for Oyster Creek Nuclear Generating Station
ML20197E9131997-11-30030 November 1997
Monthly Operating Rept for Nov 1997 for Oyster Creek Nuclear Station
ML20199E4561997-11-13013 November 1997
Safety Evaluation Accepting Ampacity Derating Analysis in Response to NRC RAI Re GL-92-08, Thermo-Lag 330-1 Fire Barriers, for Plant
ML20199D4381997-10-31031 October 1997
Monthly Operating Rept for Oct 1997 for Oyster Creek Nuclear Station
ML20202E8511997-10-21021 October 1997
Rev 0 to Scenario 47, Gpu Nuclear Oyster Creek Nuclear Generating Station Emergency Preparedness (Nrc/Fema Evaluated) 1997 Biennial Exercise. Pages 49 & 59 of Incoming Submittal Were Not Included
ML20211M9481997-10-0303 October 1997
Supplemental Part 21 Rept Re Condition Effected Emergency Svc Water Pumps Supplied by Bw/Ip Intl Inc to Gpu Nuclear, Oyster Creek Nuclear Generation Station.No Other Nuclear Generating Stations Effected by Notification
ML20198J7361997-09-30030 September 1997
Monthly Operating Rept for Sept 1997 for Oyster Creek Nuclear Generating Station
ML20211B7461997-09-24024 September 1997
Part 21 Rept Re Failure of Emergency Service Water Pump Due to Threaded Flange Attaching Column to Top Series Case Failure.Caused by Dissimilar Metals.Pumps in High Ion Svc Will Be Upgraded to 316 Stainless Steel Matl
ML20210V0181997-08-31031 August 1997
Monthly Operating Rept for Aug 1997 for Oyster Creek Nuclear Generating Station
ML20210L2961997-07-31031 July 1997
Monthly Operating Rept for Jul 1997 for Oyster Creek Nuclear Station
ML20149F9961997-07-18018 July 1997
Safety Evaluation Re Gpu Nuclear Operational Quality Assurance Plan,Rev 10 for Three Mile Island Nuclear Generating Station,Unit 1 & Oyster Creek Nuclear Generating Station
ML20196H0111997-07-11011 July 1997
Special Rept 97-001:on 970620,removed High Range Radioactive Noble Gas Effluent Monitor (Stack Ragems) from Service to Allow Secondary Calibr IAW Master Surveillance Schedule. Completed Calibr on 970628 & Returned Stack Ragems to Svc
ML20210L3081997-06-30030 June 1997
Corrected Page to MOR for June 1997 for Oyster Creek Nuclear Generating Station
ML20141H2051997-06-30030 June 1997
Monthly Operating Rept for June 1997 for Oyster Creek Nuclear Station
1999-09-30
[Table view] |
Text
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47 77) y? . LICENSEE EVENT REPORT
' CONTROL BLOCx: l l l l l l l (PLEAsE FRINT oR TYTE ALL REQulRED INFonMATioNi 1 6 2, l o t i8 l9l N JUCENSEE
' J ) 0 lCODE C l P l141 lgl5010 l - l 0 L4l CENSE 1
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7 8 60 68 69 EVENT Q ATE 74 75 REPORT DATE 80 i
EVENT DESCRIPTION AND PROBABLE CONSEQUENCES h l lo12l l As a resul t of leakage rate measurement tests performed on MSIV's NS03A, ;
~
,o,3, ; NSO4A, and NSO4B during the days immediately following plant shutdown ;
,o,,,; for refueling on September 16, 1978, all three valves were found to have ;
, ,g,3, ,
leakage rates in excess of their maximum allowable as specified in (T.S.
4 Valve NS038 could not be tested because of excess leakage I o l e i L . 5. F.1.d) . ,
j3,7; i through valves NS03A an NSO4A.
s ria i 80 i
7 8 9 SYSTEM CAUSE CAUSE COMP. VALVE CCCE CODE SUSCCCE COMPONENT CODE SU8CCCE SU8 CODE
) @] 7 8 l Cl Dl@ l E l@' 12l M l@ l VlA lL lV lE lX [@ lF l@ l 0l@
9 10 11 13 18 19 20 SEQUENTIAL OCCURRENCE REPORT REVISION 5 gR mo E R EPORT NO. CODE TYPE NO.
2
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_ 21 22 23 24 26 27 28 29 30 31 32 K N AC CN CN PLANT ME HOURS 22 58 i FCR 8. SUP Li MANUFAC RER j! l33B lgl34Xl@ l Z lg 35 l36Z[@ l0 l0 l0 l0 l 37 40 y@
41 l42 Yl@ lN 43 l@ l44A 15 l8 l547l@
CAusE DESCRIPTION AND CORRECTIVE ACTIONS
! I i ) o l l The primary sources of leakage have been attributed to stem packing in (
,, ,,,; NSO48, stem packing and a cracked stellite weld in the body seat ring in g ggg NSO4A, and the failure of the main poppet to seat properly in NS03A. ,
,, ,3g l Necessary repairs will be performed during the ongoing outage. A compl- ;
,, ,,; ete investigation of the MSIV failures is being conducted. ;
8 9 80 STA % POWER OTHER STATUS On v RY DISCOVERY CESCRIPTION l1 l 5 l l Hl@ l 0 l 0 l 0'2l@l" NA l l C l@l Leakage rate test l
'JTivirv Ca'ETE~T REtEiSEo cP aEtEASE AMouNr NA 0, ACriviTv @ l l LOCATION CF RELEASE @
NA l i l 68 l9l Z l @ l10Z l@l 7 11 44 45 l
80
- ERSONNEL EXPCSURES NUP/ 9 E R TYPE DESCRIPTION NA l i l 71 1 01 o f 01@l Z l@l l
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.c ESCR ,TiO~@ l lt l . l l O f.,gERl 0 [@l: NA l
l 7 3 9 11 12 80 LCS$
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30 iSS DESCRIPTION *
!: l o l I N@l Weekly press release -- Oernber 3.1978
)
I I!llIIItittll1 P DA. A D B C K- #fV 117 5 /PFf 2-p e8 e, acq
' 7g1-/4(fghggggggg Donald A. h s P"CN E :
~
, -![ "C @ [ $ ,, ( r yM MadYs n Avenue unca8 Pcad Mornstown, New Jersey 07960 (201)455-8200 OYSTER CREEK NUCLEAR GENERATING STATION Forked River, New Jersey 08731 Licensee Event Report Reoortable Occurrence No. 50-219/78-18-1T Report Date September 29, 1978 Occurrence Date September 16, 1978 Identification of Occurrence Violation of the ' Technical Specifications, paragraph 4.5.F.1.d, when Main Steam isolation Valves NSO4A, NS03A, and NSO4B failed to meet their maximum allowable leakage rate. This event is considered to be a reportable occurrence as defined in the Technical Specifications, paragraph 6.9 2.a.3 Conditions Prior to Occurrence Start of a refueling shutdown.
Reactor mode switch: Refuel Moderator temperature: 166*F Reactor vessel head in place
\
Descriotion of Occurrence On September 16, 1978, at approximately 1600 hours0.0185 days <br />0.444 hours <br />0.00265 weeks <br />6.088e-4 months <br />, an unsuccessful attempt was made to pressurize the reactor vessel to perform a leakage rate measurement test on inboard main steam isolation valves NSO3A and NS03B. The maximum pressure reached was approximately four (4) psig. Subsequent investigations indicated that valve NSO3A was leaking excessively through the seat area and that valve NSO4A, which is in series with NS03A, had excessive leakage through the stem packing. Because of the stem packing leak, it could not be immediately determined if the seat sealing surface of NSO4A was also leaking.
On September 18, 1978, at approximately 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br />, to further explore the sealing capability of NSO4A, the main steam lines were flooded, with valves -
NSO3A and NSO4A closed. W a ter was observed flowing from a drain line down-l l
Jersey Central Power & Light Company is a Memcer of the Ger.eral Puche Ut; lit:es System
4
'Rsportable Occurrcnca No 50-219/78-18-IT
. Page 2 September 29, 1978 stream of both valves, thus confirming excessive seat leakage in addition to stem packing leakage. Main steam isolation valve NSO4B was tested in a conventional manner and was observed to have a leakage rate of approximately 14.5 SCFH. The maximum allowable is 11 9 SCFH. The stem packing was identified as the major source of leakage for this valve. It should be noted that valve NS03B could not be tested for leakage because of the excessive leakage through the 3A and 4A valves.
On September 20, 1978, another attempt was made to quantify the amount of leakage through valves NS03A and NSO4A. This was done by attempting to measure the air flow required to pressurize between the valves using a 0-100 SCFM rotometer connected to a 100 psig air source. The maximum air flow was measured to be 24 SCFM which corresponds to the maximum flow possible through the air supply hose and the associated piping connections.
Since a similar air supply system was used in the initial attempt to pressurize the vessel, it may be concluded, assuming that all of the leakage was through the MSIV's, that a differential pressure of 4 psig across the series valves corresponds to a flow rate of 24 SCFM through the valves.
Accarent Cause of Occurrence The stem packing was identified as the primary source of leakage for NSO4B and a contributing source of leakage for NSO4A. The cause of excessive seat leakage through NS03A and NSO4A has been preliminarily determined to be from two separate and distinct causes.
For inboard main steam isolation valve NS03A, the cause of the leakage has been attributed to failure of the main poppet-to seat properly. An inspection of the valve internals indicates excessive wear between the main poppet alignment pad and the lower guide on which it slides. This condition caused a shift in the main poppet path of travel causin'g the poppet not to seat properly.
For outboard main steam isolation valve NSO4A, the cause of leakage has been attributed to a cracked stellite weld in the body seat ring. The weld was made in 1974 to repair an axial crack across the stellite inlay which was discovered during a planned valve inspection that year.
Analysis of Occurrence Under design basis loss of coolant accident conditions (LOCA), the containment isolation valves function to limit the leakage out of the containment structure to less than one (1) weight percent of the contdinment atmosphere per day.
Conformance to this criteria assures that the release limits set forth in 10CFR100 will not be exceeded. The test results demonstrated that the leakage through valves NS03A and NS038 was at least 3 5 weight percent per day of the containment volume at approximately a four (4) psig differential pressure, i
l
, i
'Aspdr^ table Occurrence No. 50-219/78-18-1T Paga 3 Septcmbar 29, 1978 4
Corrective Action Necessary repairs will be performed during the ongoing refueling maintenance outage. A complete investigation of the MSIV failures is being conducted.
At the conclusion of the investigation, a detailed report will be submitted to the Plant Operations Review Committee who will determine if further corrective action is necessary.
Failure Data Manufacturer: Atwood & Morrill Co.
Type: Angle mounted globe valve Size: 24" diameter e
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