text
%-
- - l CCN-90-14091 p
..+
n < g PHILADELPHIA ELECTRIC COMPANY PEACil llCirTOM AIDMIC POWER STATION R. D.1, llox 208 Delta, lYnnsylvania 17314
. ruce mornm-me rowia or excnuncs (717) 456-7014 I
1 May 7, 1990
.j Docket No. 50-277 Docu;nent Control Desk l
U S. Nuclear Regulatory Commission Washington, DC 20555
SUBJECT:
Licensee Event Report Peach Bottom Atomic Power Station - Unit 2 This LER concerns a daily instrument check that was not being performed as 4
required by Technical Specifications. This LER has been revised to provide l
the results of a review of previous similar events.
i
Reference:
Docket No. 50-277 Report Number:
2-90-001 Revision Number:
01 Event Date:
01/24/90 Report Date:
05/07/90 Facility:-
Peach Bottom Atomic Power Station
)
RD 1, Box 208, Delta, PA 17314 i
l This LER is being submitted pursuant to the requirements of 10 CFR 50.73(a)(2)(i)(B).
i i
Sincerely, I
Plant Manager cc:
J. J. Lyash, USNRC Senior Resident Inspector T. T. Martin, USNRC, Region I
$/
9005100223 900507 ADOCK0500g7 DR k
Y
N C Form 3e4 U 5. NUCLEAR KEGULATO3Y COMMISSION APPROVED OMB NO 3CS0104 I
LICENSEE EVENT REPORT (LER)
EXPIRi$ 8'31/98 FACILITY NAME 01 DOCtLET NUMBER (2)
P AUI
'3, Peach Bottom Atomic Power Station - Unit 2 0 l6 l 0 l0 [ 0 l 217l7 1 lorl ol 3
- * Shift Surveillance Log Did Not Meet The Requirements of Technical Specifications Due-To A nefi cient Procedure, E VENT DATE til LIR NUMBER (GI REPORT DATI 171 OTHER F ACitiTit5 INVOLVED ten MONTH DAY YEAR YEAR
7,*,Q MONTH DAY 4 EAR F AC'bliv h Aves DOckt T NUMet Rist PBAPS - Unit 3 0 i s j o l 0 l 0 l 2 l718 ol1 2 l4 9 0 9l o 0l 0 l1 0l1 ol 5 0l7 9l 0
~
0 t 5l0 l0 i 0; l l Twis RiPORT is sueMitfio PuRauaNT TO Tui ReouiREMiNT: O, iO C,R g
<ca.<...,-,..,,,, so.
,l iiii
o.one no.osm so,3aH H i Tui nii r
2f 406talt1HO 60 SeteH11 60.73telt2H.I 73 714el l00 a.o.
in H.,
ggg,g g,.;
i i
.0 =<ein,
.0 73,.ioH.,
no, 20 406teH1Hdo
)(
BO.73Eell2HJ (0.73(all2 Hventl Al J66Al 20 406teHt New) 60 T3teH2H.)
50,73isil29t.mHal 20 4061eH1H.)
60 73teH2Hedl 60 73ssH2Han tl CENSE E CONTACT FOR THl$ Lt R litt NAME TE LEPHONE NuYS(R ARkACODE A.
A.
Fulvio, Regulatory Engineer 7l117 4 15 61-t 710 11 14 COMPLETE ONE Liht POR E ACM COMPONENT F AlttJRt DEECRISED IN THis REPORT (131 "I.h0" M
F C REPORTA (
' O
CAUSE
t /$T EV COMPONgNy C AUS E sv 8T EM COYPONENT PRD5 l
1 I i i i I I
I I I I I i 1
1 1 I l' I I I
I I I l-1 I SUPPLEMENT AL REPORT E XPECTED 114)
MONTH DAY YEAR Sv8 mis 5 TON Yts III ves cemetere EXPtCHO SUO6HSSION Dew NO l
l l
usT R AcT a,.., s uv c., <,.,.. -,,,, ur.
c,,o,.~,-,o n ei On January 24, 1990, it was discovered that the daily instrument check of the main stack flow rate monitor was not being performed as required by Technical Specification (TS) 4.8.C.4.b.
The cause of this event was an incomplete procedure which occurred as a result of an inadvertent omission of this instrument check during a comprehensive revision to the shift surveillance logs. No actual safety consequences occurred as a result of this event. The shift surveillance log ST 9.1-3Z was revised to include the daily instrument check. A systematic review of TS required instrument checks has been performed to ensure TS requirements are being I adequately addressed. This event has been reviewed with the appropriate procedure I writers and the ST sponsor involved. There were three previous similar LERs.
1 i
NR,C r. u.
NRC Fenet 3864 -
U.S. NUCLEO G 81UL Af oRY COMMISSION M
LICENSEE EVENT REPORT (LER) TEXT CONTINUATION uPaovio ous No 3m oio.
(XP1RIS' 8/31/O F ACILfTV hAadgigt DOCKET NUM88R (2)
LE R NUMBE R 106 PAGE(3)
Peach Bottom Atomic Power Station vsaa "MOl?'
",'#.0
- - Unit 2 ol oll oll ol 2 or ol3 0 [5 l010 l 0 l 2 l 717 9l 0 f tKT ll* more space h reewed, seen eadeonelHRC le*m.9NKall17)
Requirements'for the Report This LER is being submitted pursuant to 10 CFR 50.73(a)(2)(1)(B) because a surveillance required by Technical Specifications (TS) was not being performed.
Unit Status at Time of the Discovery of the Event Unit 2 was in the Run Mode at 100% rated thermal power.
Unit 3 was in the Run Mode at 100% rated thermal power.
There were no structures, components, or systems that were inoperable at the start of l
the event that contributed to the event.
l Description of the Event On January 24, 1990, at 1200 hours0.0139 days <br />0.333 hours <br />0.00198 weeks <br />4.566e-4 months <br /> while performing a review of TS required _
instrument checks, it was discovered that the daily instrument check of the main stack flow rate monitor (Ells: MON) was not being performed as required by TS
.4.8.C.4.b.
The main stack flow rate monitor instrument is common to both units 2 and 3.
Prior to an October 14, 1989 revision of an operating shift surveillance log, this parameter was being logged daily.
Cause of the Event
The cause of this event was an incomplete procedure which occurred as a result of an inadvertent omission of this instrument check during a comprehensive revision to the I shift surveillance logs due to a personnel error by the Surveillance Test Sponsor.
Revision 15 of Surveillance Test (ST) 9.1-3Z (one of the shif t surveillance logs) contained a step requiring an instrument check of the main stack flow rate monitor.
- - but it had been deleted by revision 16 dated October 14, 1989.
The operating shift surveillance logs under went an extensive revision prior to Unit a
3 restart (November 1989). The procedures were extensively reformatted to improve human factors (i.e., rearranged to better coordinate the instrument order list in-the ST with the outlay of Control Room instruments), and the procedures were revised to incorporate the labeling enhancements made during Control Room modifications.
The daily check of the main stack flow rate monitor was inadvertently omitted from ST 9.1-3Z during this revision.
Analysis of the Event
No actual safety consequences occurred as a result of this event.
Although the main stack flow rate monitor was not included on the shift log surveillance test, this instrument has a history of being reliable and functioning properly. Additionally, ST 7.6.1.J " Determination of Total Noble Gas Release Rate"
.uses the main stack flow as read from the main stack flow rate monitor in determining g,0x. a.
.u... cro, m s m m - o
NRC 7.em 304A U.S. NUCLEfA $ E1ULf. TORY COMMIEStON LICENSEE EVENT REPORT (LER) TEXT CONTINUATION m ovfo cus No 3m-otu i
EXP6RES-S'3U98 9 ACiplT] ff4% 11)
DOCKET NUMBER 82)
LtR NUMS$R (s)
PAGE (31 "UtG
"T.E Peach Bottom Atomic Power Station Unit'2 0 l6 l0 l0 l0 l 2l 7l7 91 o ol0l1 ol l
'o l 3 oF ol3 iext in e.c. m,
v..=.nw w we i = sasa w me the total noble. gas release rates in site gaseous effluent.
The total noble gas release rate calculation 1s performed on average five times each week. Although the i
daily instrument check had not been performed since November 1989, routine performance of ST 7.6.1.J demonstrated that the main stack flow rate monitor was operational.
Corrective Actions
On February 1, 1990, the operating shift surveillance log ST 9.1-3Z was revised to include the daily instrument check.
From January 24, 1990 to February 1, 1990 the daily instrument check of the main stack flow rate monitor was performed during ST 7.6.1.J.
l l
A systematic review of the shift surveillance logs has been conducted to ensure operating and shutdown TS required instrument checks are included.
No additional l discrepancies were identified. This event has been reviewed with the appropricte l procedure writers and the ST Sponsor involved.
Previous Similar Events
Three previous LERs were identified, LERs 2-88-18, 2-88-27, and 2-89-01 in which TS requirements were not met due to procedural inadequacies which resulted from a previous test revision.
Part of the corrective actions associated with a root cause evaluation, which was performed as a result of these three LERs was to assign a sponsor for each ST procedure. Additionally, Administrative (A) Procedure A-47
" Surveillance Test Procedures" was revised to administrative 1y control the sponsor's responsibility during review of ST revisions.
l An-investigation has been performed to determine why these corrective actions did not l prevent this event.
A-47 requires an ST sponsor to verify that Technical l Specification surveillance requirements are not jeopardized by an ST revision.
In I this event the sponsor failed to meet this requirement by not adequately comparing l the new revision to the old revision.
l l
l l
.e.... u.
. a. c., n u.u m., -
(9 $h
|
|---|
|
|
| | | Reporting criterion |
|---|
| 05000277/LER-1990-001, :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,shift Surveillance Log Did Not Meet Requirements of Tech Specs.Caused by Deficient Procedure. Surveillance Log Revised to Include Daily Instrument Check
| 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(1) | | 05000277/LER-1990-001-02, :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 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
| 10 CFR 50.73(a)(2)(1) | | 05000278/LER-1990-001-03, :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 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
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function | | 05000278/LER-1990-002, :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,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
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(iv), System Actuation | | 05000278/LER-1990-002-02, :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 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
| 10 CFR 50.73(a)(2)(iv), System Actuation | | 05000277/LER-1990-002-03, :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 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
| 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(1) | | 05000278/LER-1990-003-02, :on 900306,reactor Scram Occurred Due to Main Turbine Trip.Caused by Improper Setting of Timer Due to Deficient Test Procedure.Procedure Revised |
- on 900306,reactor Scram Occurred Due to Main Turbine Trip.Caused by Improper Setting of Timer Due to Deficient Test Procedure.Procedure Revised
| 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(1) | | 05000277/LER-1990-003-03, :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,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
| 10 CFR 50.72(b)(3)(ii), Degraded or Unanalyzed Condition 10 CFR 50.73(a)(2) | | 05000277/LER-1990-003-01, :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 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
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2) | | 05000277/LER-1990-004-02, :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 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
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function | | 05000277/LER-1990-004, :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 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
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function | | 05000278/LER-1990-004-03, :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 900430,Tech Spec Violation Occurred When MSIV Closure Timing Testing Not Performed in Required Surveillance Interval.Caused by Ambiguous Test Procedure. Surveillance Test Revised
| 10 CFR 50.73(a)(2)(1) | | 05000277/LER-1990-005-02, :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 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
| 10 CFR 50.73(a)(2)(1) | | 05000278/LER-1990-005-03, :on 900507,Group 2A Primary Containment Isolation Sys Isolation Occurred During Surveillance Test. Caused by Inadequate Worker Practices.Blown Fuse Replaced & Personnel Counseled |
- on 900507,Group 2A Primary Containment Isolation Sys Isolation Occurred During Surveillance Test. Caused by Inadequate Worker Practices.Blown Fuse Replaced & Personnel Counseled
| 10 CFR 50.73(a)(2)(iv), System Actuation | | 05000277/LER-1990-006-02, :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 900402,actuation of Emergency Diesel Generator Occurred.Caused by Personnel Miscommunication. Shift Mgt Will Be Reminded of Necessity to Control Activities in Control Room
| 10 CFR 50.73(a)(2)(iv), System Actuation | | 05000278/LER-1990-006-03, :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 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
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(1) | | 05000277/LER-1990-007-02, :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 900412,evaluation Involving Seismic Qualification Performed Due to Postulated Failure of Condensate & Vacuum Pumps During Design Seismic Events. Caused by Design Oversight.Program Updated
| | | 05000278/LER-1990-007-03, :on 900620,RWCU Isolated While Pressurizing Sys Following Isolation Valve Insp.Caused by Design Weakness. Design of Instrumentation Reviewed & Procedural Controls Enhanced |
- on 900620,RWCU Isolated While Pressurizing Sys Following Isolation Valve Insp.Caused by Design Weakness. Design of Instrumentation Reviewed & Procedural Controls Enhanced
| 10 CFR 50.73(a)(2)(iv), System Actuation | | 05000277/LER-1990-008, :on 900414,testing of LPCI Pumps & Core Spray Subsystems Not Performed,Per Tech Spec 4.5.A.4.Caused by Personnel Error in Not Recognizing Inoperable Status of RHR Pump 2D When Changing to Refuel Mode |
- on 900414,testing of LPCI Pumps & Core Spray Subsystems Not Performed,Per Tech Spec 4.5.A.4.Caused by Personnel Error in Not Recognizing Inoperable Status of RHR Pump 2D When Changing to Refuel Mode
| 10 CFR 50.73(a)(2)(1) | | 05000278/LER-1990-008-02, :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 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
| 10 CFR 50.73(a)(2)(iv), System Actuation | | 05000277/LER-1990-009-02, :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 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
| 10 CFR 50.73(a)(2)(1) | | 05000278/LER-1990-009-03, :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 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
| 10 CFR 50.73(a)(2)(1) | | 05000278/LER-1990-010-03, :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 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
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function | | 05000277/LER-1990-010-01, :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 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
| 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2) | | 05000277/LER-1990-011-01, :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 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
| 10 CFR 50.73(a)(2)(1) | | 05000278/LER-1990-011-03, :on 900910,HPCI Sys Inoperable Due to Dirty Relay Contact in Auxiliary Oil Pump Motor Starter |
- on 900910,HPCI Sys Inoperable Due to Dirty Relay Contact in Auxiliary Oil Pump Motor Starter
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function | | 05000278/LER-1990-012-03, :on 891120,miscalibration of Reactor Level Transmitters Resulted in Tech Spec Violation |
- on 891120,miscalibration of Reactor Level Transmitters Resulted in Tech Spec Violation
| 10 CFR 50.73(a)(2)(1) | | 05000277/LER-1990-012-01, :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 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
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(1) | | 05000278/LER-1990-013-03, :on 901009,core Spray Pump Actuation Occurred During Testing Due to Personnel Error |
- on 901009,core Spray Pump Actuation Occurred During Testing Due to Personnel Error
| 10 CFR 50.73(a)(2)(iv), System Actuation | | 05000277/LER-1990-013, :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 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
| 10 CFR 50.73(a)(2)(1) | | 05000277/LER-1990-014, :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 900620,RWCU Isolation Occurred During Pressurization of Sys Following Isolation Valve Maint.Caused by Design Weakness.Design of Instrumentation Reviewed & Procedural Controls Enhanced
| 10 CFR 50.73(a)(2)(iv), System Actuation | | 05000278/LER-1990-014-03, :on 901103,unexpected ESF Actuation of Primary Containment Isolation Sys Occurred Due to Personnel Error |
- on 901103,unexpected ESF Actuation of Primary Containment Isolation Sys Occurred Due to Personnel Error
| 10 CFR 50.73(a)(2)(iv), System Actuation | | 05000278/LER-1990-015-03, :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 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
| 10 CFR 50.73(a)(2)(1) | | 05000278/LER-1990-016-03, :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 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
| 10 CFR 50.73(a)(2)(iv), System Actuation | | 05000277/LER-1990-017, :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 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
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function | | 05000277/LER-1990-020, :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 900821,discovered That Battery Charger 2D Inoperable for Undervoltage Alarm Relay Calibr.Caused by Personnel Error.Event Reviewed by Mgt W/Shift Supervisor
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function | | 05000277/LER-1990-021, :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 |
- 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
| 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(1) | | 05000277/LER-1990-022, :on 900909,RWCU Isolation Occurred During Performance of Surveillance Test |
- on 900909,RWCU Isolation Occurred During Performance of Surveillance Test
| 10 CFR 50.73(a)(2)(iv), System Actuation | | 05000277/LER-1990-023, :on 900910,engineering Personnel Discovered Unqualified External Temporary Seal Assembly |
- on 900910,engineering Personnel Discovered Unqualified External Temporary Seal Assembly
| 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.72(b)(3)(ii), Degraded or Unanalyzed Condition | | 05000277/LER-1990-024, :on 900910,personnel Failed to Perform Tech Spec Surveillance Due to Procedure Deficiency |
- on 900910,personnel Failed to Perform Tech Spec Surveillance Due to Procedure Deficiency
| 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(1) | | 05000277/LER-1990-025, :on 900913,high Pressure/Emergency Svc Water Ventilation Found Outside Design Basis Due to Deficient Design |
- on 900913,high Pressure/Emergency Svc Water Ventilation Found Outside Design Basis Due to Deficient Design
| 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.72(b)(3)(ii), Degraded or Unanalyzed Condition | | 05000277/LER-1990-026, :on 900913,HPCI Sys Declared Inoperable Due to Low Emergency Svc Water Flow Through Room Coolers |
- on 900913,HPCI Sys Declared Inoperable Due to Low Emergency Svc Water Flow Through Room Coolers
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function | | 05000277/LER-1990-028, :on 900801,discovered Missed Surveillance on Drywell Suppression Chamber Vacuum Breaker |
- on 900801,discovered Missed Surveillance on Drywell Suppression Chamber Vacuum Breaker
| 10 CFR 50.73(a)(2)(1) | | 05000277/LER-1990-029, :on 901016,E4 Diesel Generator Not Tested While RHR Loop a Inoperable |
- on 901016,E4 Diesel Generator Not Tested While RHR Loop a Inoperable
| 10 CFR 50.73(a)(2)(1) | | 05000277/LER-1990-030, :on 901024,control Room Emergency Ventilation Actuation Occurred Due to Inadvertent Opening of 480 Volt Breaker |
- on 901024,control Room Emergency Ventilation Actuation Occurred Due to Inadvertent Opening of 480 Volt Breaker
| 10 CFR 50.73(a)(2)(iv), System Actuation | | 05000277/LER-1990-031, :on 890314,missed Core Spray Surveillance Resulted in TS Violation Due to Inadequate Controls |
- on 890314,missed Core Spray Surveillance Resulted in TS Violation Due to Inadequate Controls
| 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(1) | | 05000277/LER-1990-032, :on 901106,potential Loss of Primary Containment Occurred |
- on 901106,potential Loss of Primary Containment Occurred
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.72(b)(3)(ii), Degraded or Unanalyzed Condition | | 05000277/LER-1990-033, :on 901108,Tech Spec Limiting Condition of Operation Not Entered for Inoperable Ci Valve |
- on 901108,Tech Spec Limiting Condition of Operation Not Entered for Inoperable Ci Valve
| 10 CFR 50.73(a)(2)(1) | | 05000277/LER-1990-034, :on 901112,discovered That Emergency Diesel Generator Droop Selector Switches in Parallel Position. Caused by Personnel Error.Personnel Involved Counseled on Importance of Attention to Detail |
- on 901112,discovered That Emergency Diesel Generator Droop Selector Switches in Parallel Position. Caused by Personnel Error.Personnel Involved Counseled on Importance of Attention to Detail
| 10 CFR 50.73(a)(2)(1) | | 05000277/LER-1990-035, :on 901115,discovered That Multiple Safety Sys Could Be Inoperable as Result of 4 Kv Emergency Bus Undervoltage Relays Outside Acceptable Calibr Limits.Caused by Inadequate Design.Relays Modified |
- on 901115,discovered That Multiple Safety Sys Could Be Inoperable as Result of 4 Kv Emergency Bus Undervoltage Relays Outside Acceptable Calibr Limits.Caused by Inadequate Design.Relays Modified
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.72(b)(3)(ii), Degraded or Unanalyzed Condition 10 CFR 50.73(a)(2)(1) |
|