05000277/LER-1990-033, Corrected LER 90-033-00:on 901108,discovered TS Limiting Condition of Operation Not Entered for Inoperable Containment Isolation Valve Due to Procedural Deficiency
Corrected LER 90-033-00:on 901108,discovered TS Limiting Condition of Operation Not Entered for Inoperable Containment Isolation Valve Due to Procedural Deficiency
05000278/LER-1999-005-03, :on 990920,uplanned Esfas During Planned Mod Activitives in Main CR Were Noted.Caused by Inattention to Detail by Individuals Performing Work.All CR Mods Were Ceased to Allow Review of Mod Work Packages.With
05000278/LER-1999-005-03, :on 990920,uplanned Esfas During Planned Mod Activitives in Main CR Were Noted.Caused by Inattention to Detail by Individuals Performing Work.All CR Mods Were Ceased to Allow Review of Mod Work Packages.With
LER-1990-033, Corrected LER 90-033-00:on 901108,discovered TS Limiting Condition of Operation Not Entered for Inoperable Containment Isolation Valve Due to Procedural Deficiency
CCN90-k4334 PHILADELPHIA ELECTRIC COMPANY PEAC}l BOTTO4 ATOMIC POWLk STATION L D 1, Box 208 Delu, Pennsyhsnia 17314 enum wrvou-vise Poe se of sacsuswcs (717) 4$G7044 J
December lo, 1990 1
Docket No. 50-277 Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555
SUBJECT:
Licensee Event Report Peach Bottom Atomic Power Station - Unit 2 This LER concerns a Technical Specification Limiting Condition of Operation not entered for an inoperable containment isolation valve due to procedural deficiency.
Reference:-
Docket No. 50-277 Report Number:
2-90-033 Revision Number:
00 Everit Date:
11/08/90 Report Date:
12/10/90 Facility:
Peach Bottom Atomic' Power Station RD 1 Box 208, Delta, PA 17314 This LER is being submitted pursuant to the requirements of 10 CFR 50.73(a)(2)(1)(B).
Sincerely, i
cc:
J. J. Lyash, USNRC Senior Resident Inspector T. T. Martin, USNRC, Region'l i
R. A. Burricelli, Public Service Electric & Gas Commitment Coordinator Correspondence Control Program T. M. Gerusky, Commonwealth of PennsylvaniaINPO Records Center R. 1. McLean, State of Maryland C. A. McNeill, Jr. - $26-1, PECo President and C00 D. B. Miller, Jr. - SMO-1, Vice President - PBAPS Nuclear Records - PBAPS H. C. Schwenn, VP - Atlantic Electric J. Urban. Delmarva Power r
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On 11/8/90, a resident NRC inspector discovered the bottle pressure for the compressed Hitrogen (N?) gas cylinder that suppl.ies backup gas pressure to the Air Operated ( A0)-2519, "Drywt il tr.d Torus inlet N2 Purge" valve operator and boot seal to be less than the acceptable value spccified in the daily surveillance test.
A review of the completed surveillance tests indicated that the leak rate appears to have increased above the allowable limit in May 1990. Tech Spec 3.7.0.2 should have been entered and the appropriate Limiting Condition for Operation taken when the leak rate first exceeded allowable limits. The cause of the event was due to procedure deficiencies.
No actual safety consequences occurred as a result of this event. The boot seal for A0-2519 was inflated during the entire event by the normal instrument air system. The backup N2 supply was available for the redundant inboard containment isolation valves in this penetration during the event. The leak was repaired. The surveillance test was temporarily changed to provide clear operability criteria.
Two previous similar LERs were identified.
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'This report is required per 10 CFR 50.73(a)(2)(1)(B) as a result of a condition prohibited by Tech Specs.
Unit Conditions at Time __of Event Unit 2 was in the RUN mode at 98% of rated thermal reactor power. There were no systemst structures, or components that were inoperable that contributed to this evont.
Description of Event
On 11/8/90, a resident NRC inspector discovered the bottle pressure for the Operated (AO)-2519 (Ells: gas cylinder that supplies backup gas pressure to the A compressed Nitrogen (N2) boot seal to be less than the acceptable value specified in the daily surveillance test. The Shift Supervisor (Utility, Licensed) was notified and the bottle was replaced. Due to the' leakage rate, the valve was declared inoperable and Tech Spec 3.7.0.2 Limiting Condition for Operation (LCO) vias entered for an inoperable containment isolation valve. The LCO remained in effect until modification work on A0-2519 was complete.
The leak was repaired and the LC0 was exited on 11/12/90.
A review of the completed surveillance tests indicated that the leak rate appears to have increased above the allowable limit in May 1990. Tech Spec 3.7,0.2 should have been entered and the appropriate Limiting Condition for Operation taken when the leak rate first exceeded allowable limits. The daily surveillance of bottle pressures instructs the operator to install a fresh N2 bottle and notify the System Engineer when bottle pressure drops below 1300 psig. The 1300 psig pressure criteria is based on a leak rate that ensures the the bottle could supply boot seal pressure for 20 days following design basis LOCA with a seismic event or a loss of off-site power.
Cause of Event
The cause of the event was procedure deficiencies in that clear direction is not provided to the shif t in the daily-surveillance test for determination of operability of the valve.
The surveillance test criteria concentrated on bottle pressure and not on the actual leak rate.
A contributing cause is personnel error. The Shift is directed by the surveillance test to call the System Engineer by the following working day after a N2 bottle is replaced.
This information was not being forwarded to the System Engineers.
If the System Engineers had been notified of the increased frequency of bottle replacements, the leakage problem may have been addressed.
Another contributing cause is that the importance of loss of N2 bottle supply on A0 valve operability was not clearly understood by the shift. The backup N2 system is l
not described in Tech Specs. An Engineering document exists which determined that the gas bottles are required to provide a 20 day supply of N2 to the valve boot seals j
in the event of a loss of the normal air supply. This document, in the form of a 4,s. ctm 1969 540 %89 900'#
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.u,m, JustificationforContinuedOperation(JCO)wasnotadequatelyaddressedinthe surveillance test nor were the shift personnel aware of it.
Analysis of Eve.n_t, No actual safety consequences occurred as a result of this event.
The valve operator and boot seal are supplied by the Instrument Air System (Ells:LD) under normal operation.
In the event of a loss of the instrument air H2 issupplied i
from a compressed gas cylinder to maintain boot seal pressure for up to 20 days.
The boot seal for A0-2519 was inflated during the entire event by the normal instrument air system. A0 2519 is an outboard containment isolation valve in the containment purge penetration.
The backup N2 supply was available for the inboard containment isolation valves in the containment purge penetration during this event, except for the short period detailed in LER 2-90-032.
For a Design B. asis LOCA with a seismic event or a loss of off-site power, the normal instrument air supply to the A0 would be lost, and the A0-2519 boot seal would be supplied by the gas cylinder to maintain containment integrity (EEIS JM). Historical data indicates that the leakage past the deflated boot seal would be within 10 CFR 50
- App J and Tech Spec limits.
Corrective Actions
The leak was repaired.
The surveillance test was temporarily changed to provid:
clear operability criteria. The Operators have been informed of the significance of N2 bottles and associated operability concerns. System Engineers will review the surveillance test data weekly in order to identify increased leakage which may result in exceeding the-20 day requirement.
Modification 1316 is scheduled to be installed during the upcoming Unit 2 Refueling Outage (1/91)
This modification replaces the backup bottles with nitrogen supplied directly from the Tontainment Atmospheric Dilution System.
If the daily N2 bottle Surveillance is still required following the modification it will be revised to include a per-day leak rate acceptance criteria and clear operability criteria.
The lack of awareness of the JC0 indicates a need for program improvements. Existing JCOs will be reviewed for required procedure revisions and training issues. A tracking mechanism will be established for action items resulting from JCOs.
Previous Similar Events
Two. previous similar LERs were identified. LER 2 80-030/1T-0 involved a' potentia'l i
on 900124,shift Surveillance Log Did Not Meet Requirements of Tech Specs.Caused by Deficient Procedure. Surveillance Log Revised to Include Daily Instrument Check
on 900124,discovered That Daily Instrument Check of Main Stack Flow Rate Monitor Not Performed.Caused by Incomplete Procedure.Operating Shift Surveillance Log Revised to Include Daily Instrument Check
on 900108,HPCI Sys Declared Inoperable During Surveillance Testing When Start Time Exceeded 25 S.Caused by Inadequate Calibr Procedure Which Allowed Setting of 18 S. Ramp Generator & Signal Converter Replaced
on 900128,ESF Sys Actuations Occurred Due to Reactor Vessel Level Fluctuations After Manual Scram.Caused by Failure of O-ring on Fluid Inlet Port to Servo Valve for Hydraulically Operated Valve.Valve Replaced
on 900128,reactor Manually Scrammed Due to Leak of Electrohydraulic Control Sys Fluid.Caused by Lock Nut on Interlock Dump Valve Setting Adjustment Bolt Becoming Unsecured Due to Sys Vibration.Leak Stopped
on 900402,shutdown Required to Comply W/Tech Specs W/One Inoperable Automatic Depressurization Sys (ADS) Valve Completed.Caused by Component Failure for K Automatic Depressurization Sys Valve
on 900310,excessive Primary Containment as Found Leakage Rate Discovered.Caused by Excessive Clearance Between Valve Disc & Seat Assemblies When in Closed Position.Valves Rebuilt Using New Discs
on 900310,primary Containment Leakage Rate Limit Exceeded Tech Spec Requirements Due to Excessive Through Seat Leakage on Main Steam Line Drain Isolation Valves MO-74 & MO-77
on 900321,discovered Potentially Inoperable Safety Sys Due to Inadequate Emergency Svc Water Cooling Flow Through Room Coolers.Caused by Gradual Buildup of Corrosion & Silt.Mod Completed
on 900430,Tech Spec Violation Occurred When MSIV Closure Timing Testing Not Performed in Required Surveillance Interval.Caused by Ambiguous Test Procedure. Surveillance Test Revised
on 900321,determined That Under DBA Conditions Some ECCS & RCIC Sys Room Coolers Would Not Receive Min Acceptable Emergency Svc Water Flow.Caused by Buildup of Corrosion Products.Piping Inspected
on 900326,Tech Spec Surveillance Not Performed within Required Interval.Caused by Personnel Error.Personnel Counseled & Will Periodically Review Omitted Test Rept to Ensure Performance of Surveillance Tests
on 900511,blown Fuse from Battery Charger 3B Resulted in Declaring HPCI Sys,Core Spray B Logic,Rhr B Logic,Core Spray Subsystem B,Rhr Subsystem B & E2 & E4 Emergency Diesel Generators Inoperable
on 900402,actuation of Emergency Diesel Generator Occurred.Caused by Personnel Miscommunication. Shift Mgt Will Be Reminded of Necessity to Control Activities in Control Room
on 900412,evaluation Involving Seismic Qualification Performed Due to Postulated Failure of Condensate & Vacuum Pumps During Design Seismic Events. Caused by Design Oversight.Program Updated
on 900620,RWCU Isolated While Pressurizing Sys Following Isolation Valve Insp.Caused by Design Weakness. Design of Instrumentation Reviewed & Procedural Controls Enhanced
on 900727,offgas Recombiner Isolation Occurred Causing Main Condenser Vacuum to Begin Decreasing.Caused by Component/Sys Failure in Offgas Recombiner Condensate Sys. Design of Sys Will Be Evaluated for Root Cause
on 900417,discovered That Testing of LPCI Pumps & Core Spray Subsystems Not Performed When LPCI Pump D Declared Inoperable on 900414.Caused by Personnel Error. Procedures Declaring Pump Inoperable Revised
on 900727,discovered That Recirculation Loop Temp Increased in Excess of Allowable Heat Up Rate & Temp Reading Not Recorded at Required Intervals.Caused by Personnel Error & Procedural Deficiency
on 900430,discovered That Rod Block Monitor Not Been Proven Operable Prior to Exceeding 30% Power as Required by Tech Specs.Caused by Programmatic Deficiency. General Plant Procedures Revised
on 900804,HPCI Declared Inoperable Due to Failure of Manual & Overspeed Trip Tappet Assembly.Caused by Design Problem Re Tappet Swelling.New Design Being Provided by GE Will Replace Tappet Assembly
on 900502,three Control Room Emergency Ventilation Actuations Occurred.Caused by Poor Electrical Continuity as Result of Oxidation Between plug-in Circuit Boards & Mating Electrical Connections
on 900503,discovered That Tech Spec 3.4.1 Not Performed on 6-wk Frequency as Required.Caused by Personnel Error.Tracking of Surveillance Activities Scheduled to Be Transferred to Improved Software Package
on 890815,discovered Valves Left Closed After Removal of Blocking Permit & on 890813,emergency Cooling Water Pump & Emergency Diesel Generator Removed from Svc. Caused by Inadequate Procedures
on 900530,failure to Perform Tech Spec Surveillance Tests Since 891130 Discovered.Caused by Incorrect Std Practice.Tests Revised to Clarify Testing Requirements of Tech Spec 4.14.C.1.c
on 900620,RWCU Isolation Occurred During Pressurization of Sys Following Isolation Valve Maint.Caused by Design Weakness.Design of Instrumentation Reviewed & Procedural Controls Enhanced
on 901125,operators Failed to Record Drywell Sump Flow Readings Every 4 H Per Tech Spec 3.6.C.Caused by Personnel Error.Timer Will Be Obtained to Alarm Every 4 H to Prompt Operator to Record Drywell Readings
on 900707,control Room Emergency Ventilation Actuation Occurred as Result of High Radiation Signal from Control Room Ventilation Sys Radiation Monitor B.Cause of Spurious Signal Under Investigation
on 901202,main Steam Relief Valve 71J Opened for Approx 3/4 Due to Installation of Jumper in Wrong Panel.Caused by Inadequate pre-job Briefing.Technicians Counseled & Procedures Will Be Revised
on 900718,two Control Room Emergency Ventilation Sys Actuations Resulted from Spurious High Radiation Signals from Radiation Monitor B.Cause Under Investigation.Diagnostic Equipment Installed
on 900724,HPCI Declared Inoperable Due to Battery Charger 2B Transient.Caused by Degradation of Foam Rubber Electrical Module Support Piece.Foam Replaced & Procedures Revised to Include Foam Replacement
on 900730,control Room Emergency Ventilation Actuation Occurred Due to High Radiation Signal from B Radiation Monitor.Cause Under Investigation.Sys Restored to Normal Standby Alignment
on 900821,discovered That Battery Charger 2D Inoperable for Undervoltage Alarm Relay Calibr.Caused by Personnel Error.Event Reviewed by Mgt W/Shift Supervisor
on 900830,reactor Level Instrumentation Associated W/Ref Leg 2B Condensing Chamber Declared Inoperable.Caused by Loss of Inventory in Ref Leg Due to Erroneous Reactor Water Level Readings