:on 920508,containment Third Door Was Blocked Open During Refueling Operations.Caused by Interpretation That Movement of Core Components Per TS Definitions Rather than Literal Wording.Routine Maintenance Procedure Revised| ML20147J449 |
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| Site: |
Point Beach  |
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| Issue date: |
05/01/1997 |
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| From: |
Flentje F WISCONSIN ELECTRIC POWER CO. |
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| To: |
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| Shared Package |
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| ML20147J446 |
List: |
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| References |
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| LER-97-017, LER-97-17, NUDOCS 9705070170 |
| Download: ML20147J449 (4) |
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text
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l NRC 7ORM 366' U.S. NUCLEAR REOULATORY COMMISSION APPROVED BY OM8 NO. 3150-0104 (4-95)
EXPlRES 04/30/98 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS.
LICENSEE EVENT REPORT (LER)
REPORTED LESSONS LEARNED ARE INCORPORATED INTO THE LICENSING PROCESS AND FED BACK TO INDUSTRY.
FORWARD COMMENTS REGARDING BURDEN ESTIMATE (See reverse for required number of TO THE INFORMATION AND RECORDS MANAGEMENT digits / characters for each block)
BRANCH (T-6 F33),
U.S.
NUCLEAR REGULATORY COMMISSION, W ASHINGTON, DC 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT FACILITV NAME (1)
DOCKET NUMBER (2)
PAGE (3)
Point Beach Nuclear Plant, Unit 1 05000266 1 OF 4 TITLE (4)
Containment Third Door Blocked Open During Refueling Operations j
EVENT DATE (5) l LER NUMBER (8)
REPORT DATE (7)
OTHER FACILITIES INVOLVED (8) l YEAR SEQUENTIAL REVISION FACILITY NAME DOCKET NUMBER MONTH DAY YEAR NUMBER NUMBER MONTH DAY YEAR 08I92I97 FACILITY NAME DOCKET NUMBER 05 017 --
00 05 01 97 05000 OPERATINO l THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR 8: (Check one or more) (11)
MODE (9)
N l 20 220im 20.22031aH2Hvl X
60.73(a)(2His 50.73(aH2)(viiii l
20 2203(a)(1) 20.2203(aH3)(il 50.73(a)(2)(ii) 50.73(a)(2)(x)
POWER LEVEt. (10) 000 20.2203(aH2)(i) 20 2203(a)(3Hii) 50.73(aH2)(in) 73.71 20.2203(a)(2)(ii) 20.2203(aH4) 50.73(a)(2)(iv) oTHitR 20.2203(a)(2Hiii) 50.36(c)(1) 50.73(a)(2Hv) specify m Abstract below 20.2203(a)(2)(iv) 60.36(c)(2) 50.73(a)(2Hv4) e' m hlRC Form 366A LICEN CONTACT FOR THIS LER (12)
N A ME TELEPHONE NUMBER (include Area Cods)
Fritzie A.
Flentje, Regulatory Specialist (414) 755-6221 COMPLETE ONE LINE FOR E ACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)
CAUSE
SYSTEM COMPONENT MANUFACTURER REPORTABLE
CAUSE
SYSTE M CGMPONENT MANUFACTURER REPORT ABL E TO NPR9S TO NPRDS B
VI UDMP N
B VI DUCT N
SUPPLEMENTAL REPORT EXPECTED (14)
EXPECTED MON TH DAY YEAR YES SUBM:SSION (if yes, comp 6ete EXPECTED SUBMISSION DATE).
X NO DATE (15)
ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines) (16)
On March 31, 1997, during a review of condition reports relating to potential Technical Specifications violations, it was discovered that a condition report was initiated on May 9, 1992, that described an event which had been screened as not being a Technical Specification violation.
Accordingly, no 30-day licensee event report was submitted to the NRC. On on May 8, 1992, immediately following installation of the Unit 1 reactor vessel upper internals package, it was discovered that ramps used to transport equipment into and out of the containment blocked open the temporary third door on the upper personnel containment access.
TS 15.3.8.1 requires that,
....a temporary third door on the outside of the personnel lock shall be in place whenever both doors in a personnel lock are open."
NRC FORM 366 (4 95) 9705070170 970501 PDR ADOCK 05000266 S
PDR
- U.S. NUCLE A;4 RELULATORY COMMISSION (4-C5)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION f ACILITY NAMI (1)
DOCKET NUMBER (2)
LER NUMBER (65 PAGE (3)
YEAR SEQUENTIAL REVISION Point Beach Nuclear Plant, Unit 1 05000266 NUMMR NUMMR 2OF4 97 017 00 TEKT !!!rnere space is required, use additionalcopks of NRC Forrn 366A) (17) i
Event Description
i l
On March 31, 1997, during a review of potential Technical Specification j
violations, it was discovered that a violation of plant Technical Specifications occurred on May 8, 1992.
On May 8, 1992, immediately l
j following installation of the the Unit 1 reactor vessel upper internals, it was noted that portable ramps that are used to transport equipment into and out of containment were blocking open the temporary third door on the upper personnel containment access.
At the time of discovery, both doors J
of the personnel lock were open.
A condition report was submitted on May 9,
1992, documenting this event.
1 The condition report, CR 92-237, indicated that the ramps had been in place, thus blocking open the temporary third door, during the installation of the reactor vessel upper internals.
The condition report had been determined to not be a violation of Technical Specifications.
Specification 15.3.8.1 states, "A temporary third door on the outside of the personnel lock shall be in place whenever both doors in a personnel lock are open (except for initial core loading."
During installation of the upper internals, both doors of the personnel lock were open.
1 The reason provided for this event not considered to be a violation of Technical Specifications was that TS 15.1 defines refueling as, "Any operation involving movement of core components (those that could affect the reactivity of the core) within the containment..."
Replacement of the reactor vessel upper internals was interpreted to not be within the
~
context of this definition.
A retrospective review of the determination made in 1992 concluded that the event should have been reported as a violation of Specification 15.3.8.1.
cause
The initial reportability review based its conclusion upon an interpretation that movement of core components per the definitions section of Technical Specification rather than the literal wording of Specification 15.3.8.1.
Specification 15.3.8.1 requires that the temporary third door on the outside of the personnel lock be in place whenever both doors in a personnel lock are open.
f4RC FORM 366A (4-95)
eU.C. NUCLEA;; REOULATORY COMMSSSIM (4-osi LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME fil DOCKET NUMBER (2)
LER NUMBER 16)
PAGE (3)
YEAR SEQUENTIAL REVISION Point Beach Nuclear Plant, Unit 1 05000266 NUMMR NUMMR 3OF4 97 017 00 iEXT (11more space is requwed, use additional copies of NRC Form 366A) (17)
Corrective Actions
1.
The routine maintenance procedure for removing and installing the reactor vessel head and upper internals was revised to verify that i
refueling requirements are met prior to replacing the reactor vessel upper internals.
This action was completed with issuance of the i
revised procedure on September 4, 1992.
2.
The operations refueling procedure was revised to require the temporary third door to be closed prior to installation of the upper internals.
This action was completed with issuance of the revised procedure on August 18, 1992.
Subsequent to this event, a similar event occurred during the Unit i 1993 spring refueling outage.
The event was determined to be reportable in accordance with 10 CFR 50.73 (a) (2) (i).
LER 266/93-004-00 was submitted on May 10, 1993.
It was further determined that the corrective actions take as a result of the previous event (described herein) had not been completely offective in preventing recurrence of this type of event.
Accordingly, additional short and long-term corrective actions, as described in the referenced report were taken.
These corrective actions included:
1.
Refueling procedures were revised to ensure that no equipment is placed into or extended through the personnel locks during refueling operations.
Refueling operations include unlatching / latching of control rod drive shafts, removing / replacing the upper internals, and removing / replacing the reactor vessel head in addition to fuel movement.
2.
An alarm has been installed on thu third door such that any activity or movement of equipment and personnel through the door which causes the temporary door to be open greater than 2 minutes results in generation of an alarm.
The alarm is received by Security and the control room is immediately notified.
The control room contacts the core loading supervisor, who in turn, ensures immediate response by either Security or Operations to the open temporary third door, and/or suspension of refueling operations.
Reportability
This licensee event report is being submitted in accordance with the requirements of 10 CFR 50.73 (a) (2) (i), "Any operation or condition prohibited by the plant's Technical Specifications."
PRC FORM 366A 14 95)
~.
1U.S. NUCLEAR REOULATLRY COMM64;lON il 951 LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACitlTY NAME (1)
DOCKET NUMBER (2)
LER NUMBER (6)
PAGE (3)
YEAR SEQUENTIAL HEVISION Point Beach Nuclear Plant, Unit 1 05000266 NUMBER NUMBER 4 OF 4 97 017 00 TEXT ll1 more space is required, use addttwnel copies of NRC form 366A) (17) s
safety Assessment
1 There are no safety consequences from this event. The temporary third door was capable of being shut with minimal effort by removal of the ramps, had this been necessary.
Additionally, Point Beach Nuclear Plant abnormal operating procedure AOP-8B, " Irradiated Fuel Handling Accident in Containment," requires the core loading supervisor or his designated alternate to verify the temporary third doors are closed following a fuel handling accident in containment.
The health and safety of plant personnel and the public were not endangered by this occurrence.
j l
Similar Occurrences:
There were two events that are similar to this event.
The first event, was documented in Licensee Event Report 91-003-00, dated November 8,
- 1991, for Point Beach Unit 2.
The second event, subsequent to this event, was documented in Licensee Event Report 93-004-00, dated May 10, 1993, for Unit 1.
No similar events that have occurred since the above noted licensee event report.
ceneric Implications None.
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| 05000266/LER-1997-001, :on 970108,safety Injection Delay Times Exceeded Design Basis Values.Caused by Degraded Voltage Conditions.Licensee Engineers Will Prepare FSAR Change Requests to Reflect LBLOCA Evaluation |
- on 970108,safety Injection Delay Times Exceeded Design Basis Values.Caused by Degraded Voltage Conditions.Licensee Engineers Will Prepare FSAR Change Requests to Reflect LBLOCA Evaluation
| 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition | | 05000301/LER-1997-001, Forwards LER 97-001-00,re Containment Structure Where Internal Containment Structural Members Could Have Damaged Containment Liner During Safe Shutdown Earthquake | Forwards LER 97-001-00,re Containment Structure Where Internal Containment Structural Members Could Have Damaged Containment Liner During Safe Shutdown Earthquake | | | 05000301/LER-1997-001-01, :on 970107,containment Liner Clearance Was Not IAW Plant Design Basis.Caused by Void Between Containment Liner & Concrete Containment Structure.Inspected Containment |
- on 970107,containment Liner Clearance Was Not IAW Plant Design Basis.Caused by Void Between Containment Liner & Concrete Containment Structure.Inspected Containment
| 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2) | | 05000266/LER-1997-002, :on 970109,potential to Overpressurize Piping Between Containment Isolation Valves Occurred.Caused by Original Design Not Providing Overpressure Protection for Piping.Review Completed |
- on 970109,potential to Overpressurize Piping Between Containment Isolation Valves Occurred.Caused by Original Design Not Providing Overpressure Protection for Piping.Review Completed
| 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition | | 05000301/LER-1997-002-01, :on 970415,potential Reactor Coolant Sys Branch Connection Stresses Beyond Design Basis,Indicated.Caused by Mod Initiated to Remove RTD Bypass Line Isolation Valves. Stress Analysis Conducted on RTD Bypass Piping |
- on 970415,potential Reactor Coolant Sys Branch Connection Stresses Beyond Design Basis,Indicated.Caused by Mod Initiated to Remove RTD Bypass Line Isolation Valves. Stress Analysis Conducted on RTD Bypass Piping
| 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition | | 05000266/LER-1997-003, :on 970109,did Not Perform Leak Test on Spare Containment Penetrations Per Ts.Caused by Lack of Routine Testing.Tested Penetrations W/Satisfactory Results |
- on 970109,did Not Perform Leak Test on Spare Containment Penetrations Per Ts.Caused by Lack of Routine Testing.Tested Penetrations W/Satisfactory Results
| 10 CFR 50.73(a)(2)(1) | | 05000301/LER-1997-004-01, :on 970729,declared RHR Loop Inoperable Due to CCW Leak.Caused by Failure of RHR Heat Exchanger CCW Piping. Repaired Piping & Declared RHR Loop Operable |
- on 970729,declared RHR Loop Inoperable Due to CCW Leak.Caused by Failure of RHR Heat Exchanger CCW Piping. Repaired Piping & Declared RHR Loop Operable
| 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | | 05000266/LER-1997-004, :on 970113,potential for Particular Common Mode Failure That Could Affect Opposite Trains of Unit 2 Safeguards Equipment Was Noted.Caused by Lack of Physical Separation.Replaced Subject Circuit Breakers |
- on 970113,potential for Particular Common Mode Failure That Could Affect Opposite Trains of Unit 2 Safeguards Equipment Was Noted.Caused by Lack of Physical Separation.Replaced Subject Circuit Breakers
| 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition | | 05000301/LER-1997-005-01, :on 970806,RHR Pump Was Declared Inoperable Due to Abnormal Seal Leakage from Loop a RHR 2P-10A.Repaired RHR Pump |
- on 970806,RHR Pump Was Declared Inoperable Due to Abnormal Seal Leakage from Loop a RHR 2P-10A.Repaired RHR Pump
| 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | | 05000266/LER-1997-005, :on 970116,1SI-852A Was Not Tested IAW Inservice Test Program Required by Tss.Caused Because Condition Revealed That Valve 1SI-852A Had Not Been Completely Tested.Tests Will Be Reviewed |
- on 970116,1SI-852A Was Not Tested IAW Inservice Test Program Required by Tss.Caused Because Condition Revealed That Valve 1SI-852A Had Not Been Completely Tested.Tests Will Be Reviewed
| 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2) 10 CFR 50.73(a)(2)(1) | | 05000266/LER-1997-006, :on 970120,refueling Cavity Drain Failed During Loca.Caused by Inadequate Evaluation of Original Design.Design of Refueling Cavity Drains Was Revised with Respect Capability to Withstand an Earthquake |
- on 970120,refueling Cavity Drain Failed During Loca.Caused by Inadequate Evaluation of Original Design.Design of Refueling Cavity Drains Was Revised with Respect Capability to Withstand an Earthquake
| 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition | | 05000266/LER-1997-007, :on 970124,determined That Potential Existed for EDG Overload Condition.Caused by Failure to Recognize This Condition When Plants Initially Licensed W/Two Edgs. Implemented Procedure Changes |
- on 970124,determined That Potential Existed for EDG Overload Condition.Caused by Failure to Recognize This Condition When Plants Initially Licensed W/Two Edgs. Implemented Procedure Changes
| 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition | | 05000266/LER-1997-008, :on 970131,non-seismic Ductwork Located Above safety-related Equipment in Containment Occurred.Caused by Incomplete Seismic Evaluation.Mods Will Be Completed During Current Unit 2 Refueling Outage |
- on 970131,non-seismic Ductwork Located Above safety-related Equipment in Containment Occurred.Caused by Incomplete Seismic Evaluation.Mods Will Be Completed During Current Unit 2 Refueling Outage
| 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(x) 10 CFR 50.72(b)(3)(ii), Degraded or Unanalyzed Condition | | 05000266/LER-1997-009, :on 970214,potential for Safety Injection Failure During Filling of Safety Injection Accumulator Discovered.Caused by Situation Not Adequately Covered by Procedures.Procedure OI-100 Revised |
- on 970214,potential for Safety Injection Failure During Filling of Safety Injection Accumulator Discovered.Caused by Situation Not Adequately Covered by Procedures.Procedure OI-100 Revised
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function | | 05000266/LER-1997-010, :on 970219,svc Water & Component Cooling Water TS Action Requirements Were Not Met.Caused Because Licensee Did Not Comply W/Cold Shutdown Requirements of TS 15.3.3.C.2 & 15.3.3.D.2.Evaluations Were Performed |
- on 970219,svc Water & Component Cooling Water TS Action Requirements Were Not Met.Caused Because Licensee Did Not Comply W/Cold Shutdown Requirements of TS 15.3.3.C.2 & 15.3.3.D.2.Evaluations Were Performed
| 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | | 05000266/LER-1997-011, :on 970305,containment Fan Cooler Accident Fans Were Not Tested in Accordance with Tss.Caused by non-conservative Interpretation of Literal Requirements of Tss.Unit 1 & 2 Accident Fans Were Tested |
- on 970305,containment Fan Cooler Accident Fans Were Not Tested in Accordance with Tss.Caused by non-conservative Interpretation of Literal Requirements of Tss.Unit 1 & 2 Accident Fans Were Tested
| 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(1) | | 05000266/LER-1997-012, :on 970304,diesel-drive Fire Pump Day Tank Not Sampled IAW TSs.Non-conservative Interpretation of TS Led to Failure.Day Tank T-30 Sample Was Drawn & Analyzed W/Satisfactory Results |
- on 970304,diesel-drive Fire Pump Day Tank Not Sampled IAW TSs.Non-conservative Interpretation of TS Led to Failure.Day Tank T-30 Sample Was Drawn & Analyzed W/Satisfactory Results
| 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(1) | | 05000266/LER-1997-013, :on 970304,CCWS Found Not in Accordance W/ Plant Design Basis.Caused by Inoperable Valve Due to Overtorquing in Closed position.Cross-tie Will Be Resolved |
- on 970304,CCWS Found Not in Accordance W/ Plant Design Basis.Caused by Inoperable Valve Due to Overtorquing in Closed position.Cross-tie Will Be Resolved
| 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition | | 05000266/LER-1997-013-01, Forwards Suppl LER 97-013-01,re Component Cooling Water Sys Not IAW Plant Design Basis.Rept Replaces LER 97-013-00 in Its Entirety & Includes Addl Similar Occurrence Not Previously Reported to NRC | Forwards Suppl LER 97-013-01,re Component Cooling Water Sys Not IAW Plant Design Basis.Rept Replaces LER 97-013-00 in Its Entirety & Includes Addl Similar Occurrence Not Previously Reported to NRC | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition | | 05000266/LER-1997-014, :on 970321,auxiliary Feedwater Sys Inoperability Due to Loss of Instrument Air.Design Mods Initiated,Providing Pneumatic Supply to Control Valves |
- on 970321,auxiliary Feedwater Sys Inoperability Due to Loss of Instrument Air.Design Mods Initiated,Providing Pneumatic Supply to Control Valves
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.72(b)(3)(ii), Degraded or Unanalyzed Condition 10 CFR 50.73(a)(2)(viii) | | 05000266/LER-1997-015, :on 970324,control Room Ventilation Sys Declared Inoperable Due to Failures of Backdraft Damper & Vent Duct Access Door.Backdraft Damper,Replaced |
- on 970324,control Room Ventilation Sys Declared Inoperable Due to Failures of Backdraft Damper & Vent Duct Access Door.Backdraft Damper,Replaced
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function | | 05000266/LER-1997-016, :on 970325,SG Level Logic Was Not Tested IAW Ts.Caused by Nonconservative Interpretation of Tss.Ts Amends Proposed to Provide Consistency Between Test Requirements & LCO Associated W/Sg Tests |
- on 970325,SG Level Logic Was Not Tested IAW Ts.Caused by Nonconservative Interpretation of Tss.Ts Amends Proposed to Provide Consistency Between Test Requirements & LCO Associated W/Sg Tests
| 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | | 05000266/LER-1997-017, :on 920508,containment Third Door Was Blocked Open During Refueling Operations.Caused by Interpretation That Movement of Core Components Per TS Definitions Rather than Literal Wording.Routine Maintenance Procedure Revised |
- on 920508,containment Third Door Was Blocked Open During Refueling Operations.Caused by Interpretation That Movement of Core Components Per TS Definitions Rather than Literal Wording.Routine Maintenance Procedure Revised
| 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(x) 10 CFR 50.72(b)(3)(ii), Degraded or Unanalyzed Condition | | 05000266/LER-1997-018, :on 970403,potential for RHR Overpressure During Accidents Was Discovered.Original Design Did Not Provide Overpressure Protection for Isolated Piping Section. Evaluation Was Performed to Determine Stress on Piping |
- on 970403,potential for RHR Overpressure During Accidents Was Discovered.Original Design Did Not Provide Overpressure Protection for Isolated Piping Section. Evaluation Was Performed to Determine Stress on Piping
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function | | 05000266/LER-1997-019, :on 970404,RHR Not Aligned IAW TS Requirements. Caused by non-conservative Decision Making & Not Recognizing When TS Were Not Controlling Plant Operations.Pbnp Mgt Philosophy Re TS Interpretations Changed to Minimize Use |
- on 970404,RHR Not Aligned IAW TS Requirements. Caused by non-conservative Decision Making & Not Recognizing When TS Were Not Controlling Plant Operations.Pbnp Mgt Philosophy Re TS Interpretations Changed to Minimize Use
| 10 CFR 50.73(a)(2)(1) | | 05000266/LER-1997-020-01, Forwards LER 97-020-01,describing Plant Conditions in Which Ability to Achieve & Maintain Safe Shutdown in Event of Postulated Fire May Have Been Adversely Affected | Forwards LER 97-020-01,describing Plant Conditions in Which Ability to Achieve & Maintain Safe Shutdown in Event of Postulated Fire May Have Been Adversely Affected | 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor | | 05000266/LER-1997-021, :on 970430,determined That Spent Fuel Pool Cooling Sys Was Not in Accordance W/Plant Design Basis.Cause Indeterminate.Closed & re-tagged Valves SF-27 & SF-28 & Investigated Basis for Fsar,App a |
- on 970430,determined That Spent Fuel Pool Cooling Sys Was Not in Accordance W/Plant Design Basis.Cause Indeterminate.Closed & re-tagged Valves SF-27 & SF-28 & Investigated Basis for Fsar,App a
| 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition | | 05000266/LER-1997-022, :on 970507,discovered That Postulated Control Room Fire May Cause Electrical Hot Short That Disables Limit or Torque Switches for Certain Movs.Mods Initiated to Remedy Condition |
- on 970507,discovered That Postulated Control Room Fire May Cause Electrical Hot Short That Disables Limit or Torque Switches for Certain Movs.Mods Initiated to Remedy Condition
| 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor | | 05000266/LER-1997-023, :on 970508,discovered Noncompliant Emergency Lighting for Postulated App R Fires.Caused by Alternative Provisions Made in Original Safe Shutdown Analysis.Emergency Lights Will Be Installed |
- on 970508,discovered Noncompliant Emergency Lighting for Postulated App R Fires.Caused by Alternative Provisions Made in Original Safe Shutdown Analysis.Emergency Lights Will Be Installed
| 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition | | 05000266/LER-1997-024, :on 970501,determined Post Accident Sampling Sys Degradation.Caused by Inadequate Design Review.Will Upgrade Containment Atmosphere Sample Sys & Will Perform Mod to Reduce Dose within GDC 19 Dose Limits |
- on 970501,determined Post Accident Sampling Sys Degradation.Caused by Inadequate Design Review.Will Upgrade Containment Atmosphere Sample Sys & Will Perform Mod to Reduce Dose within GDC 19 Dose Limits
| | | 05000266/LER-1997-025, :on 970520,pressurizer Level Was Controlled Higher than Assumed in Accident Analysis.Caused by Inappropriately Changing Procedures W/O Adequate Consideration.Listed Affected Procedures Will Be Revised |
- on 970520,pressurizer Level Was Controlled Higher than Assumed in Accident Analysis.Caused by Inappropriately Changing Procedures W/O Adequate Consideration.Listed Affected Procedures Will Be Revised
| 10 CFR 50.72(b)(3)(ii), Degraded or Unanalyzed Condition | | 05000266/LER-1997-026, :on 970521,discovered TS Violation of Operability Requirement of MSL Isolation.Caused by Inadequate Consideration for Operability of All Required Functions.Verified Low RCS Sys Average Temp |
- on 970521,discovered TS Violation of Operability Requirement of MSL Isolation.Caused by Inadequate Consideration for Operability of All Required Functions.Verified Low RCS Sys Average Temp
| 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | | 05000266/LER-1997-027, :on 970521,non-environmentally Qualified Matl Existed in Containment Hatch Applications.Caused by Inadequate Design Review.Mods Will Be Performed to Remove Existing Teflon Material |
- on 970521,non-environmentally Qualified Matl Existed in Containment Hatch Applications.Caused by Inadequate Design Review.Mods Will Be Performed to Remove Existing Teflon Material
| | | 05000266/LER-1997-031, :on 970619,discovered That Auxiliary Feedwater (AFW) Pump Low Suction Pressure Trip Setpoints May Not Ensure Adequate Suction Pressure Protection for AFW Pumps Following Tornado Event.Caused by Inadequate Design |
- on 970619,discovered That Auxiliary Feedwater (AFW) Pump Low Suction Pressure Trip Setpoints May Not Ensure Adequate Suction Pressure Protection for AFW Pumps Following Tornado Event.Caused by Inadequate Design
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function | | 05000266/LER-1997-032, :on 970630,discovered Inadequately Rated Electrical Buses Could Disable Switchgear & Cause Secondary Fires.Caused by Characteristic of Original Design. Established twice-per-shift Fire Watches |
- on 970630,discovered Inadequately Rated Electrical Buses Could Disable Switchgear & Cause Secondary Fires.Caused by Characteristic of Original Design. Established twice-per-shift Fire Watches
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function | | 05000266/LER-1997-034, :on 970707,discovered Unplanned Loss of Voltage on Train B Safeguards Buses.Caused by Inadequate Design & Design Review for Installation of New Train B Edgs.Incorrect Wiring Reworked |
- on 970707,discovered Unplanned Loss of Voltage on Train B Safeguards Buses.Caused by Inadequate Design & Design Review for Installation of New Train B Edgs.Incorrect Wiring Reworked
| 10 CFR 50.73(a)(2)(iv), System Actuation | | 05000266/LER-1997-035, :on 970516,discovered Inadequate Seismic Support for Reactor Coolant Pump Rotor Stand.Caused by Rotor Stand Being Stored Since Initial Plant Construction.Moved Rotor Stand & Verified as Seismically Adequate |
- on 970516,discovered Inadequate Seismic Support for Reactor Coolant Pump Rotor Stand.Caused by Rotor Stand Being Stored Since Initial Plant Construction.Moved Rotor Stand & Verified as Seismically Adequate
| 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat | | 05000266/LER-1997-036, :on 970826,potential Common Mode Failure in DC Power Supply Which Could Disable AFW Sys Was Noted.Caused by Inadequate Design & Design Review.Plant Mods Were Performed to Eliminate Potential Common Mode Failure |
- on 970826,potential Common Mode Failure in DC Power Supply Which Could Disable AFW Sys Was Noted.Caused by Inadequate Design & Design Review.Plant Mods Were Performed to Eliminate Potential Common Mode Failure
| | | 05000266/LER-1997-037, :on 970903,potential Failure of EDG Load Sequence Occurred.Caused by Inadequate Design of EDG Load Sequencing Logic.Mod Restored Operability of EDG During Load Sequencing |
- on 970903,potential Failure of EDG Load Sequence Occurred.Caused by Inadequate Design of EDG Load Sequencing Logic.Mod Restored Operability of EDG During Load Sequencing
| | | 05000266/LER-1997-038, :on 970926,determined That Inoperability of Standby Emergency Power Placed Unit 2 in 7-day Lco.Caused by Failure That Occurred When EDG G-03 Was Shutdown.Repaired Governor & Returned EDG G-03 to Service |
- on 970926,determined That Inoperability of Standby Emergency Power Placed Unit 2 in 7-day Lco.Caused by Failure That Occurred When EDG G-03 Was Shutdown.Repaired Governor & Returned EDG G-03 to Service
| 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | | 05000266/LER-1997-039-01, :on 970615,RHR Loop Inoperable.Caused by Removal of CCW Pump from Svc.Ccw Pump Restored |
- on 970615,RHR Loop Inoperable.Caused by Removal of CCW Pump from Svc.Ccw Pump Restored
| 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | | 05000266/LER-1997-039, Forwards LER 97-039-00 Re RHR Loop Inoperable,Due to Inoperable CCW Pump.New Commitments within Rept Indicated in Italics | Forwards LER 97-039-00 Re RHR Loop Inoperable,Due to Inoperable CCW Pump.New Commitments within Rept Indicated in Italics | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | | 05000266/LER-1997-040-01, Forwards LER 97-040-01 Which Documents Event That Occurred at Point Beach Nuclear Plant,Unit 1.Commitments Made within Ltr,Encl | Forwards LER 97-040-01 Which Documents Event That Occurred at Point Beach Nuclear Plant,Unit 1.Commitments Made within Ltr,Encl | 10 CFR 50.72(b)(3)(ii), Degraded or Unanalyzed Condition | | 05000266/LER-1997-041, :on 971023,potential Common Mode Failure in Afws Control Circuits Was Noted.Caused by AFW Control Circuits Installed by Plant Mods.Temporary Mods Will Restore Physical Separation for Cables |
- on 971023,potential Common Mode Failure in Afws Control Circuits Was Noted.Caused by AFW Control Circuits Installed by Plant Mods.Temporary Mods Will Restore Physical Separation for Cables
| | | 05000266/LER-1997-042, :on 971030,discovered That Upper Containment Personnel Air Interlock Had Been Inoperable.Caused by Removal of Remote Operating Gear.Reinstalled Remote Operating Connector Gear |
- on 971030,discovered That Upper Containment Personnel Air Interlock Had Been Inoperable.Caused by Removal of Remote Operating Gear.Reinstalled Remote Operating Connector Gear
| 10 CFR 50.73(a)(2)(1) | | 05000266/LER-1997-043-01, Forwards LER 97-043-01,re Discovery That TS Surveillance of Reactor Trip Sys Interlocks Were Not Adequate. Supplemental Info Is Provided at End of Rept.Previous Commitments Made within Rept Also Encl | Forwards LER 97-043-01,re Discovery That TS Surveillance of Reactor Trip Sys Interlocks Were Not Adequate. Supplemental Info Is Provided at End of Rept.Previous Commitments Made within Rept Also Encl | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown | | 05000266/LER-1997-044, :on 971216,use of Dedicated Operators During IST of Containment Spray Sys Constituted Operation Prohibited by Ts.Caused by Improper Consideration for Use of Dedicated Operators.Revised Procedures |
- on 971216,use of Dedicated Operators During IST of Containment Spray Sys Constituted Operation Prohibited by Ts.Caused by Improper Consideration for Use of Dedicated Operators.Revised Procedures
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