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| {{#Wiki_filter:NRC FORM 366 U.S.NUCLEAR REGULAT COMMISSION ie-1999)LICENSEE EVENT REPORT (LER)(See reverse for required number of digits/characters, for each block)APPROV MB NO.3150-0104 EXPIRES 06/30/2001 Estimated burden per response to comply with this mandatory information cogection request: 50 hrs.Reported lessons learned are incorporated into the licensing process and fed back to industry.Forward comments regarding burden estimate to the Records Management Branch (T4 F33), U.S.Nuclear Regutatory Commission. | | {{#Wiki_filter:NRC FORM 366 U.S. NUCLEAR REGULAT COMMISSION APPROV MB NO. 3150-0104 EXPIRES 06/30/2001 ie-1999) |
| Washington, DC 205554&1, and to the PapenNork Reduction Project (3t504104), Oirice of Management and Budget.Washington. | | Estimated burden per response to comply with this mandatory information LICENSEE EVENT REPORT (LER) cogection request: 50 hrs. Reported lessons learned are incorporated into the licensing process and fed back to industry. Forward comments regarding burden estimate to the Records Management Branch (T4 F33), U.S. Nuclear Regutatory (See reverse for required number of Commission. Washington, DC 205554&1, and to the PapenNork Reduction digits/characters, for each block) Project (3t504104), Oirice of Management and Budget. Washington. DC 20503. |
| DC 20503.lf an information collection does not display a currently valrd OMB control number.the NRC may not conduct or sponsor, and a person is not required to respond to.the information collection. | | lf an information collection does not display a currently valrd OMB control number. the NRC may not conduct or sponsor, and a person is not required to respond to. the information collection. |
| FACILITY NAME (1)Cook Nuclear Plant Unit 1 DOCKET NUMBER I2)05000-315 PAGE I3)1OF4 TITLE t4)Inadequate Technical Specification Surveillance Testing of Essential Service Water Pump Engineered Safety Feature Response Time EVENT DATE{5)LER NUMBER{6)REPORT DATE (7)OTHER FACILITIES INVOLVED (8)MONTH DAY YEAR YEAR 1999 SEQUENTIAL REVISION NUMBER NUMBER MONTH DAY YEAR FACILITY NAME FACIUIY NAME DOCKET NUMBER DOCKET NUMBER OPERATING MODE (9)POWER LEVEL (10)20.2201 (b)20.2203(a)
| | FACILITYNAME (1) DOCKET NUMBER I2) PAGE I3) |
| (1)20.2203(a)(2)(i)20.2203(a) | | Cook Nuclear Plant Unit 1 05000-315 1OF4 TITLE t4) |
| (2)(v)20.2203(a)(3)(I) 20.2203(a)(3)(ii) 50.73(a)(2)(i) 50.73(a)(2)(ii) 50.73(a)(2)(iii) | | Inadequate Technical Specification Surveillance Testing of Essential Service Water Pump Engineered Safety Feature Response Time EVENT DATE {5) LER NUMBER {6) REPORT DATE (7) OTHER FACILITIES INVOLVED (8) |
| D PURSUANT TO THE REQUIREMENTS OF 10 CFR 5: (Check THIS REPORT IS SUBMITTE one or moro){11)50.73(a)(2){viii)50.73(a)(2)(x)73.71 20.2203(a)(2)(ii) 20.2203(a)(2)(iii) 20.2203(a)
| | MONTH DAY YEAR YEAR SEQUENTIAL REVISION MONTH DAY YEAR FACILITY NAME DOCKET NUMBER NUMBER NUMBER 1999 FACIUIYNAME DOCKET NUMBER OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR 5: (Check one or moro) {11) |
| {2)(iv)20.2203(a)
| | MODE (9) 20.2201 (b) 20.2203(a) (2)(v) 50.73(a)(2)(i) 50.73(a) (2) {viii) |
| (4)50.36(c){1)50.36(c)(2) | | POWER 20.2203(a) (1) 20.2203(a)(3)(I) 50.73(a)(2)(ii) 50.73(a) (2)(x) |
| LICENSEE CONTACT FOR THIS LER{12)50.73(a)(2)(iv) 50.73(a)(2)(v)50.73(a)(2)(vii)OTHER Specify in Abstract below or In NRC Form 366A NAM E Mary Beth Depuydt, Regulatory Compliance TELEPHONE NUMBER IInrSude Area Code)(616)465-5901 X1 589 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)cAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE To EPIX CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE To EPIX SUPPLEMENTAL REPORT EXPECTED 14 YEs{If yes, complete EXPECTED SUBMISSION DATE).NO EXPECTED MONTH DAY YEAR ABSTRACT (Limit to 1400 spaces, i.o., approximately 15 single-spaced typewritten linos)(16)On June 24, 1999, during a review of the findings of Expanded System Readiness Review (ESRR)team by Performance Assurance (PA), it was discovered that no testing program could be identified which verifies the capability of the Essential Service Water (ESW)pumps to meet the Engineered Safety Feature (ESF)response time speciTied in the Technical Specifications (TS)or the Updated Safety Analysis Report.Subsequent investigation confirmed that in-place TS surveillance testing measured the ESF response time for the ESW pumps as the elapsed time from actuation of the channel sensor until pump breaker closure, but did not include the time until a specified pump discharge pressure is reached or until the ESW pump discharge valve is open;as required by the definition of Engineered Safety Feature Response Time.Since existing surveillance testing did not satisfy the TS definition of ESF response time, the identified condition constitutes a missed surveillance test.This is an operation or condition prohibited by TS and was determined to be reportable pursuant to the requirements of 10CFR50.73(a)(2)(i)(B) on September 7, 1999.The apparent cause of this event was the inadequate understanding of the plant design basis.Surveillance tests will be revised and implemented to include the time to achieve prescribed pump discharge pressure/flow and/or discharge valve position as part of the overall ESF response time testing for the ESW system prior to restart of each respective unit.The ESW ESF response times in UFSAR Table 7.2-7 will be evaluated and revised, if necessary, prior to restart of each respective unit.ESW system performance records and surveillance test results provide reasonable assurance that the system has remained capable of performing its interlded functions.
| | LEVEL (10) 20.2203(a)(2) (i) 20.2203(a)(3)(ii) 50.73(a)(2)(iii) 73.71 20.2203(a)(2)(ii) 20.2203(a) (4) 50.73(a)(2)(iv) OTHER 20.2203(a)(2)(iii) 50.36(c) {1) 50.73(a)(2) (v) Specify in Abstract below or In NRC Form 366A 20.2203(a) {2)(iv) 50.36(c)(2) 50.73(a) (2)(vii) |
| Therefore, there were minimal safety implications to the health and safety of the public as a result of this event.'I)9i0i30i94 99i007 PDR ADGCI{l 050003i5 S PDR | | LICENSEE CONTACT FOR THIS LER {12) |
| | NAME TELEPHONE NUMBER IInrSude Area Code) |
| | Mary Beth Depuydt, Regulatory Compliance (616) 465-5901 X1 589 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13) cAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE To EPIX To EPIX SUPPLEMENTAL REPORT EXPECTED 14 EXPECTED MONTH DAY YEAR YEs NO |
| | {Ifyes, complete EXPECTED SUBMISSION DATE). |
| | ABSTRACT (Limit to 1400 spaces, i.o., approximately 15 single-spaced typewritten linos) (16) |
| | On June 24, 1999, during a review of the findings of Expanded System Readiness Review (ESRR) team by Performance Assurance (PA), it was discovered that no testing program could be identified which verifies the capability of the Essential Service Water (ESW) pumps to meet the Engineered Safety Feature (ESF) response time speciTied in the Technical Specifications (TS) or the Updated Safety Analysis Report. Subsequent investigation confirmed that in-place TS surveillance testing measured the ESF response time for the ESW pumps as the elapsed time from actuation of the channel sensor until pump breaker closure, but did not include the time until a specified pump discharge pressure is reached or until the ESW pump discharge valve is open; as required by the definition of Engineered Safety Feature Response Time. |
| | Since existing surveillance testing did not satisfy the TS definition of ESF response time, the identified condition constitutes a missed surveillance test. This is an operation or condition prohibited by TS and was determined to be reportable pursuant to the requirements of 10CFR50.73(a)(2)(i)(B) on September 7, 1999. The apparent cause of this event was the inadequate understanding of the plant design basis. Surveillance tests will be revised and implemented to include the time to achieve prescribed pump discharge pressure/flow and/or discharge valve position as part of the overall ESF response time testing for the ESW system prior to restart of each respective unit. The ESW ESF response times in UFSAR Table 7.2-7 will be evaluated and revised, if necessary, prior to restart of each respective unit. ESW system performance records and surveillance test results provide reasonable assurance that the system has remained capable of performing its interlded functions. Therefore, there were minimal safety implications to the health and safety of the public as a result of this event. |
| | 'I)9i0i30i94 99i007 PDR ADGCI{l 050003i5 S PDR |
|
| |
|
| NRC FORM 366A U.S.NUCLEAR RE TORY COMMISSION l+,16.1998) | | NRC FORM 366A U.S. NUCLEAR RE TORY COMMISSION l+,16.1998) |
| LICENSEE EVENT REPORT (LER)TEXT CONTINUATION FACILITY NAME I1)DocKET I2)LER NUMBER (6)PAGE I3)Cook Nuclear Plant Unit 1 05000-315"EAR 1999 SEQUENTIAL NUMBER 023 REVISION 2 QF 4 NUMBER 00 TEXT llf more spaceis required, use additional copies of NRC Form 366A J)17)CONDITIONS PRIOR TO EVENT Unit 1 was defueled Unit 2 was defueled DESCRIPTION OF THE EVENT On June 24, 1999, during a review of the findings of Expanded System Readiness Review (ESRR)team by Performance Assurance (PA), it was documented that no testing program could be identified which verifies the capability of the Essential Service Water (ESW)pumps to meet the Engineered Safety Feature (ESF)response time specified in the Technical Specifications or the Updated Safety Analysis Report.Subsequent investigation of this condition by Engineering, completed September 1, 1999, confirmed that the acceptance criteria for in-place Technical Specification surveillance testing defined the ESF response time for the ESW pumps as the elapsed time from actuation of the channel sensor until pump breaker closure.Testing did not include the time until a specified pump discharge pressure is reached or the ESW pump discharge valve is open, as required by the definition of Engineered Safety Feature Response Time.The Technical Specification (TS)and UFSAR definition of Engineered Safety Feature Response Time.is that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor to until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.).Since existing surveillance testing did not satisfy the TS definition of ESF response time, the identified condition constitutes a missed surveillance test.CAUSE OF THE EVENT The apparent cause of this event was inadequate understanding of the design basis of the plant.During the development of the ESW ESF response times, the design basis requirements for ESW availability during an accident were inadequately understood. | | LICENSEE EVENT REPORT (LER) |
| This resulted in surveillance procedures for ESW which did not satisfy the UFSAR and Technical Specification definition of ESF response time.ANALYSIS OF THE EVENT The Technical Specification (TS)and UFSAR definition of Engineered'Safety Feature Response Time is that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor to until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.).Since existing surveillance testing did not satisfy the TS definition of ESF response time, the identified condition constitutes a missed surveillance test.This is an operation or condition prohibited by TS and was determined to be reportable pursuant to the requirements of 10CFR50.73(a)(2)(i)(B) on September 7, 1999.Response times for Engineered Safety Features are provided in the UFSAR, Section 7.2, Table 7.2-7.The ESF Response Time Basis Procedure specifies the strategy used at Cook Nuclear Plant to demonstrate the operability of various Engineered Safety Features, systems and sub-systems. | | TEXT CONTINUATION FACILITY NAME I1) DocKET I2) LER NUMBER (6) PAGE I3) |
| This procedure defines"Device Response Time as the time from Safeguards Master Relay closing until the component reaches its ESF position.Additionally,"ESF Response Time" is defined as the time interval from when the monitored parameter exceeds its ESF actuation se'tpoint at the channel sensor until the ESF equipment is capable of performing it's safety function.Technical Specifications Surveillance Requirements for ESF response time in Section 4.3.2.1.3 and Table 3.3-3 specify that each Engineered Safety.Feature Actuation Signal (ESFAS)function will be demonstrated to be within limits at least once per 18 moriths.Review of Emergency Diesel Generator Load Sequencing and ESF Testing revealed the ESF response time for the ESW NRC FORM 366A I6-1998) | | Cook Nuclear Plant Unit 1 05000-315 "EAR SEQUENTIAL NUMBER REVISION 2 QF 4 NUMBER 1999 023 00 TEXT llfmore spaceis required, use additional copies of NRC Form 366A J )17) |
| NRC FORM 366A U.S.NUCLEAR REGULATORY COMMISSION i6-1999)" LICENSEE EVENT REPORT tLER)TEXT CONTINUATION FACIUTY NAME I1)DOCKET{2)LER NUMBER I6)PAGE I3)Cook Nuclear Plant Unit 1 05000-315 YEAR SEQUENTIAL NUMBER REYIsI0N 3 OF 4 NUMBER 1999 023 00 TEXT (If more spaceis required, use additional copies of NRC Form 366AJ{17)pumps is measured from the initiating sensor channel to the pump breaker closure.Testing does not include the time for the pump to reach the required discharge pressure or for the ESW pump discharge valve to open.In early 1975, operational problems identified with the ESW system, including severe water hammer at pump start-up, lead to testing being performed under various operational transients. | | CONDITIONS PRIOR TO EVENT Unit 1 was defueled Unit 2 was defueled DESCRIPTION OF THE EVENT On June 24, 1999, during a review of the findings of Expanded System Readiness Review (ESRR) team by Performance Assurance (PA), it was documented that no testing program could be identified which verifies the capability of the Essential Service Water (ESW) pumps to meet the Engineered Safety Feature (ESF) response time specified in the Technical Specifications or the Updated Safety Analysis Report. Subsequent investigation of this condition by Engineering, completed September 1, 1999, confirmed that the acceptance criteria for in-place Technical Specification surveillance testing defined the ESF response time for the ESW pumps as the elapsed time from actuation of the channel sensor until pump breaker closure. Testing did not include the time until a specified pump discharge pressure is reached or the ESW pump discharge valve is open, as required by the definition of Engineered Safety Feature Response Time. The Technical Specification (TS) and UFSAR definition of Engineered Safety Feature Response Time. is that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor to until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Since existing surveillance testing did not satisfy the TS definition of ESF response time, the identified condition constitutes a missed surveillance test. |
| This testing did not result in a significant water hammer, however, a previous test and operating experience showed that the water hammer did not occur when'an idle pump was started with a throttled discharge valve even though its header had not been pressurized for as long as twelve hours.Determination was made that the water hammers were induced upon the start of an idle ESW pump with a fully open discharge valve even when the header had been depressurized for no more than a few minutes.This determination lead to modification of the design of the ESW pump discharge valves, such that the valves remain closed when the ESW pump is idle, and are interlocked to open on ESW pump start at breaker closure.Response times for sensor actuation to ESW pump breaker closure and ESW pump discharge MOV stroke times are measured under the Surveillance Test Program.However, these times are not combined to provide an overall ESF response time which meets the TS definition and which is compared to an acceptance criteria.The ESF response time test procedure was reviewed to verify that ESF pumps other than ESW are tested from pump start to required system pressure/flow. | | CAUSE OF THE EVENT The apparent cause of this event was inadequate understanding of the design basis of the plant. During the development of the ESW ESF response times, the design basis requirements for ESW availability during an accident were inadequately understood. This resulted in surveillance procedures for ESW which did not satisfy the UFSAR and Technical Specification definition of ESF response time. |
| Each was verified to include requirements to measure the overall response time from sensor actuation until an acceptable discharge pressure or flow prescribed by acceptance criteria.Although the ESW ESF response times are included in UFSAR Table 7.2-7, ESW response times are not explicitly included in the UFSAR Chapter 14.0 accident analysis assumptions. | | ANALYSIS OF THE EVENT The Technical Specification (TS) and UFSAR definition of Engineered'Safety Feature Response Time is that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor to until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Since existing surveillance testing did not satisfy the TS definition of ESF response time, the identified condition constitutes a missed surveillance test. This is an operation or condition prohibited by TS and was determined to be reportable pursuant to the requirements of 10CFR50.73(a)(2)(i)(B) on September 7, 1999. |
| ESW is not immediately required to support the containment spray system (CTS)and Emergency Diesel Generator during a design basis Loss of Coolant Accident (LOCA).ESW system performance records and surveillance test results provide reasonable assurance that the system has remained capable of performing its intended functions. | | Response times for Engineered Safety Features are provided in the UFSAR, Section 7.2, Table 7.2-7. The ESF Response Time Basis Procedure specifies the strategy used at Cook Nuclear Plant to demonstrate the operability of various Engineered Safety Features, systems and sub-systems. This procedure defines "Device Response Time as the time from Safeguards Master Relay closing until the component reaches its ESF position. Additionally, "ESF Response Time" is defined as the time interval from when the monitored parameter exceeds its ESF actuation se'tpoint at the channel sensor until the ESF equipment is capable of performing it's safety function. |
| Based upon the above information, there were minimal safety implications to the health and safety of the public as a result of this event.CORRECTIVE ACTIONS Surveillance tests will be revised and implemented to include the time to achieve prescribed pump discharge pressure/flow and/or discharge valve position as part of the overall ESF response time testing for the ESW system prior to restart of each respective unit.The ESW ESF response times in UFSAR Table 7.2-7 will be evaluated and revised, if necessary, prior to restart of each respective unit.As discussed in letter AEP:NRC:1260GH,"Donald C.Cook Nuclear Power Plant, Units 1 and 2, Enforcement Actions 98-150, 98-151, 98-152 and 98-156, Reply To Notice Of Violation Dated October 13, 1998," dated March 19, 1999, a surveillance program owner and manager position has been established, reporting to the Work Control Director.A Leadership Plan has been developed which includes the creation of a detailed surveillance data base to align surveillance requirements to specific implementing procedures and a comprehensive adequacy review of surveillance testing procedures. | | Technical Specifications Surveillance Requirements for ESF response time in Section 4.3.2.1.3 and Table 3.3-3 specify that each Engineered Safety. Feature Actuation Signal (ESFAS) function will be demonstrated to be within limits at least once per 18 moriths. |
| As previously discussed in LER 315/99-021-00 and as part of Restart Action Plan 0 0001 for the Programmatic Breakdown in Surveillance Testing, the adequacy of the TS surveillance program will be evaluated. | | Review of Emergency Diesel Generator Load Sequencing and ESF Testing revealed the ESF response time for the ESW NRC FORM 366A I6-1998) |
| This evaluation includes verification that TS surveillance requirements for all modes of plant operation are incorporated into TS surveillance test procedures. | | |
| Also, as part of the Restart effort, System and programmatic assessments in the Expanded System Readiness Reviews and Licensing Basis Reviews are reestablishing and documenting the plant's Design and Licensing Basis.NRC FORM 366A i6-1999) | | NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION i6-1999) |
| | " |
| | LICENSEE EVENT REPORT tLER) |
| | TEXT CONTINUATION FACIUTY NAME I1) DOCKET {2) LER NUMBER I6) PAGE I3) |
| | Cook Nuclear Plant Unit 1 05000-315 YEAR SEQUENTIAL NUMBER REYIsI0N 3 OF 4 NUMBER 1999 023 00 TEXT (Ifmore spaceis required, use additional copies of NRC Form 366AJ {17) pumps is measured from the initiating sensor channel to the pump breaker closure. Testing does not include the time for the pump to reach the required discharge pressure or for the ESW pump discharge valve to open. |
| | In early 1975, operational problems identified with the ESW system, including severe water hammer at pump start-up, lead to testing being performed under various operational transients. This testing did not result in a significant water hammer, however, a previous test and operating experience showed that the water hammer did not occur when'an idle pump was started with a throttled discharge valve even though its header had not been pressurized for as long as twelve hours. Determination was made that the water hammers were induced upon the start of an idle ESW pump with a fully open discharge valve even when the header had been depressurized for no more than a few minutes. This determination lead to modification of the design of the ESW pump discharge valves, such that the valves remain closed when the ESW pump is idle, and are interlocked to open on ESW pump start at breaker closure. |
| | Response times for sensor actuation to ESW pump breaker closure and ESW pump discharge MOV stroke times are measured under the Surveillance Test Program. However, these times are not combined to provide an overall ESF response time which meets the TS definition and which is compared to an acceptance criteria. The ESF response time test procedure was reviewed to verify that ESF pumps other than ESW are tested from pump start to required system pressure/flow. Each was verified to include requirements to measure the overall response time from sensor actuation until an acceptable discharge pressure or flow prescribed by acceptance criteria. |
| | Although the ESW ESF response times are included in UFSAR Table 7.2-7, ESW response times are not explicitly included in the UFSAR Chapter 14.0 accident analysis assumptions. ESW is not immediately required to support the containment spray system (CTS) and Emergency Diesel Generator during a design basis Loss of Coolant Accident (LOCA). ESW system performance records and surveillance test results provide reasonable assurance that the system has remained capable of performing its intended functions. Based upon the above information, there were minimal safety implications to the health and safety of the public as a result of this event. |
| | CORRECTIVE ACTIONS Surveillance tests will be revised and implemented to include the time to achieve prescribed pump discharge pressure/flow and/or discharge valve position as part of the overall ESF response time testing for the ESW system prior to restart of each respective unit. |
| | The ESW ESF response times in UFSAR Table 7.2-7 will be evaluated and revised, if necessary, prior to restart of each respective unit. |
| | As discussed in letter AEP:NRC:1260GH, "Donald C. Cook Nuclear Power Plant, Units 1 and 2, Enforcement Actions 98-150, 98-151, 98-152 and 98-156, Reply To Notice Of Violation Dated October 13, 1998," dated March 19, 1999, a surveillance program owner and manager position has been established, reporting to the Work Control Director. A Leadership Plan has been developed which includes the creation of a detailed surveillance data base to align surveillance requirements to specific implementing procedures and a comprehensive adequacy review of surveillance testing procedures. |
| | As previously discussed in LER 315/99-021-00 and as part of Restart Action Plan 0 0001 for the Programmatic Breakdown in Surveillance Testing, the adequacy of the TS surveillance program will be evaluated. This evaluation includes verification that TS surveillance requirements for all modes of plant operation are incorporated into TS surveillance test procedures. Also, as part of the Restart effort, System and programmatic assessments in the Expanded System Readiness Reviews and Licensing Basis Reviews are reestablishing and documenting the plant's Design and Licensing Basis. |
| | NRC FORM 366A i6-1999) |
| | |
| | NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION I6-1998) |
| | LICENSEE EVENT REPORT (LER) |
| | TEXT CONTINUATION FACILITY NAME I1) DOCKET I2) LER NUMBER I6) PAGE I3) |
| | Cook Nuclear'lant Unit 1 05000-315 YEAR SEQUENTIAL NUMBER REVISION 4 OF 4 NUMBER 1999 023 00 TEXT (lfmore spaco is required, uso additional copies of NRC Form 386A/ I 17) |
| | SIMILAR EVENTS 315/99-010-00 315/99-015-00 315/99-016-00 315/99-021-00 NRC FORM 366A I6.1998) |
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| NRC FORM 366A U.S.NUCLEAR REGULATORY COMMISSION I6-1998)LICENSEE EVENT REPORT (LER)TEXT CONTINUATION FACILITY NAME I1)Cook Nuclear'lant Unit 1 DOCKET I2)05000-315 YEAR LER NUMBER I6)SEQUENTIAL REVISION NUMBER NUMBER PAGE I3)4 OF 4 1999 023 00 TEXT (lf more spaco is required, uso additional copies of NRC Form 386A/I 17)SIMILAR EVENTS 315/99-010-00 315/99-015-00 315/99-016-00 315/99-021-00 NRC FORM 366A I6.1998)
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| ~'J}} | | ~'J}} |
LER 99-023-00:on 990907,inadequate TS Surveillance Testing of ESW Pump ESF Response Time Noted.Caused by Inadequate Understanding of Plant Design Basis.Surveillance Tests Will Be Revised & Implemented| ML17335A553 |
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| Site: |
Cook  |
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| Issue date: |
10/07/1999 |
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| From: |
Depuydt M
INDIANA MICHIGAN POWER CO. |
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| To: |
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| Shared Package |
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| ML17335A552 |
List: |
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| References |
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| LER-99-023, NUDOCS 9910130194 |
| Download: ML17335A553 (7) |
|
Similar Documents at Cook |
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Category:LICENSEE EVENT REPORT (SEE ALSO AO
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LER 99-S02-00:on 990428,vulnerability in Safeguard Sys That Could Allow Unauthorized Access to Protected Area Was Noted. Caused by Inadequate Original Plant Design.Mods Were Made to Wall Opening to Eliminate Nonconforming Conditions
ML17265A8231999-05-24024 May 1999
LER 98-037-01:on 990422,determined That Ice Condenser Bypass Leakage Exceeds Design Basis Limit.Caused by Pressure Seal Required by Revised W Design Not Incorporated Into Aep Design.Numerous Matl Condition Walkdowns & Assessments Made
ML17325B6001999-05-20020 May 1999
LER 99-012-00:on 990420,concluded That Auxiliary Bldg ESF Ventilation Sys Not Capable of Maintaining ESF Room Temps post-accident.Caused by Inadequate Control of Sys Design Inputs.Comprehensive Action Plan Being Developed
ML17325B5861999-05-10010 May 1999
LER 99-002-00:on 990415,discovered That TS 4.0.5 Requirements Were Not Met Due to Improperly Performed Test. Caused by Incorrect Interpretation of ASME Code.App J Testing Will Be Completed & Procedures Will Be Revised
ML17325B5811999-05-0404 May 1999
LER 99-011-00:on 990407,air Sys for EDG Will Not Support Long Operability.Caused by Original Design Error.Temporary Mod to Supply Makeup Air Capability in Modes 5 & 6 Was Prepared
ML17325B5771999-05-0303 May 1999
LER 99-010-00:on 990401,RCS Leak Detection Sys Sensitivity Not in Accordance with Design Requirements Occurred.Caused by Inadequate Original Design of Containment Sump Level. Evaluation Will Be Performed to Clearly Define Design
ML17325B5581999-04-16016 April 1999
LER 99-006-00:on 990115,personnel Identified Discrepancy Between TS 3.9.7 Impact Energy Limit & Procedure 12 Ohp 4030.STP.046.Caused by Lack of Design Basis Control.Placed Procedure 12 Ohp 4030.STP.046 on Administrative Hold
ML17325B5471999-04-12012 April 1999
LER 99-009-00:on 990304,as-found RHR Safety Relief Valve Lift Setpoint Greater than TS Limit Occurred.Cause Investigation for Condition Has Not Been Completed.Update to LER Will Be Submitted,Upon Completion of Investigation
ML17325B5321999-04-0707 April 1999
LER 99-S01-00:on 990308,discovered That Lock for Vital Gate Leading to Plant 4KV Switchgear Area Was Nonconforming & Vulnerable to Unauthorized Access.Caused by Inadequate Gate Design & Inadequate Procedures.Mods Are Being Made to Gate
ML17325B5161999-04-0101 April 1999
LER 99-007-00:on 981020,calculations Showed That Divider Barrier Between Upper & Lower Containment Vols Were Overstressed.Engineers Are Currently Working on Analyses of Loads & Stress on Enclosures
ML17325B5221999-03-29029 March 1999
LER 99-001-00:on 960610,degraded Component Cooling Water Flow to Containment Main Steam Line Penetrations,Identified on 990226.Caused by Inadequate Understanding of Design Basis.Additional Investigations Ongoing
ML17325B4801999-03-18018 March 1999
LER 99-004-00:on 971030,failure to Perform TS Surveillance Analyses of Rc Chemistry with Fuel Removed Was Noted.Cause of Event Is Under Investigation.Corrected Written Job Order Activities Used to Control SD Chemistry Sampling
ML17325B4741999-03-18018 March 1999
LER 99-005-00:on 940512,determined That Rt Breaker Manual Actuations During Rod Drop Testing Were Not Previously Reported.Caused by Lack of Training.Addl Corrective Actions,Including Preventative Actions May Be Developed
ML17325B4571999-02-24024 February 1999
LER 99-003-00:on 990107,CR Pressurization Sys Surveillance Test Did Not Test Sys in Normal Operating Condition.Caused by Failure to Recognize Door 12DR-AUX415 as Part of CR Pressure Boundary.Performed Walkdown of Other Doors
ML17335A5171999-02-11011 February 1999
LER 99-002-00:on 990112,determined That RCS Pressurizer PORVs Had Not Been Tested,Per Ts.Caused by Inadequate Scheduling Controls Allowing Personnel Error.Surveillance Procedure Was Completed & Updated LER Will Be Submitted
ML17335A5141999-02-10010 February 1999
LER 99-001-00:on 990106,noted That GE Hfa Relays Installed in EDGs May Not Meet Seismic Qualification.Caused by Operating Experience Info Incorrectly Dispositioned in 1985. Updated LER Will Be Submitted by 990405
ML17335A5011999-02-0101 February 1999
LER 98-060-00:on 981231,identified That Rt Sys Response Time Testing Did Not Comply with TS Definition.Caused by Inadequate Procedures.Corrective Actions Will Be Developed & Update to LER Will Be Submitted by 990415.With 990201 Ltr
ML17335A4951999-01-29029 January 1999
LER 98-059-00:on 981230,interim LER -single Failure in Containment Spray Sys Could Result in Containment Spray Ph Outside Design Occurred.Investigation Into Condition Continuing.Update Will Be Submitted by 990514 Ltr
ML17335A4961999-01-27027 January 1999
LER 98-057-00:on 981228,discovered That AFW Valves Were Not Tested IAW Inservice Testing Program.Caused by Failure to Recognize Design Bases Features Re Afws by Personnel. Updated LER Will Be Submittted by 990415.With 990127 Ltr
ML17335A4921999-01-19019 January 1999
LER 98-052-01:on 981128,no Analysis for NSR Sc Manual Loader for Tdafwp Could Be Found in Original Design.Cause Due to All Failure Modes Not Considered When Compressed Air Sys Originally Designed.Performed Review of Components
ML17335A4721999-01-0606 January 1999
LER 98-055-00:on 981207,potential for Condition Outside of Design Bases for Rod Control Sys Was Noted.Caused by Calibration Error Coupled with Single Rod Failure.Condition Rept Investigation Is Ongoing
ML17335A4691999-01-0606 January 1999
LER 98-056-00:on 981211,hot Leg Nozzle Gaps Resulted in Plant Being in Unanalyzed Condition.Analyses Are Being Performed by W to Resolve Problem.Updated LER Will Be Submitted by 990211.With 990106 Ltr
ML17335A4661999-01-0505 January 1999
LER 98-049-00:on 981020,emergency Boron Injection Flow Path Was Inoperable.Caused by Original Design Deficiency. Engineering Evaluation of Event Is Continuing
ML17335A4631999-01-0404 January 1999
LER 98-054-00:on 981202,discovered That at Least One MSSV Had Not Been Reset as Required by Ts.Engineering Is Continuing Review of Extent of Condition for Event.Updated LER Will Be Submitted by 990129.With 990104 Ltr
ML17335A4481998-12-30030 December 1998
LER 98-053-00:on 981130,discovered Use of Inoperable Substitute Subcooling Margin Monitor.Caused by Condition Existing Since Installation of Plant Process Computer in 1992.Updated LER Will Be Submitted.With 981230 Ltr
ML17335A4581998-12-28028 December 1998
LER 98-052-00:on 981128,turbine Driven AFW Pump Speed Controller Failure Mode Occurred.Caused Because Not All Failure Modes Were Considered When Compressed Air Sys Was Originally Designed.Verified Current Design Change Process
ML17335A4281998-12-22022 December 1998
LER 98-051-00:on 981122,reactor Trip Signal from Manual Safety Injection Not Verified as Required by TS Surveillance,Was Discovered.Maintenance Currently Evaluating Significance & Cause of Event
ML17335A4111998-12-17017 December 1998
LER 98-047-00:on 981117,potential for Increase Leakage from Reactor Coolant Pump Seals Was Identified.Util Is Working with W to Resolve Issue.Current Expectations Are to Submit Update to LER by 990215.With 981217 Ltr
ML17335A4141998-12-16016 December 1998
LER 98-058-00:on 981216,postulated High Line Break Could Result in Condition Outside Design Bases for AF Occurred. Caused by Deficiencies Associated with Administration of HELB Program.Analysis of AF Will Be Completed by 990122
ML17335A4181998-12-16016 December 1998
LER 98-050-00:on 980814,ancillary Equipment Installed in Ice Condenser Was Not Designed to Withstand Design Basis Accident/Earthquake Loads.Caused by Lack of Established Design Criteria.Developed Design Criteria
ML17335A3871998-12-11011 December 1998
LER 98-031-01:on 980610,potential Common Mode Failure of RHR Pumps Were Noted.Caused by Inaccurate Values.Accurate Miniflow Numbers Have Been Determined by Flow Testing
ML17335A3821998-12-0808 December 1998
LER 98-039-01 Re EOP Step Conflicts with Small Break LOCA Analysis.Ler 98-039-00 Has Been Canceled.With 981208 Ltr
ML17335A3781998-12-0707 December 1998
LER 98-007-00:on 981106,high Energy Line Break Effects in Auxiliary FW Sys Was Noted.Cause of Event Is Under Investigation & Will Be Completed by 990220.Updated LER Will Be Submitted by 990310.With 981207 Ltr
ML17335A3771998-12-0303 December 1998
LER 98-046-00:on 981103,determined That Afs Was Unable to Meet Design Flow Requirements During Special Test.Caused by Failure to Consider All Aspects of Sys Operation in Design of Suction Basket Strainers.Sys Will Be Redesigned
ML17335A3741998-12-0202 December 1998
LER 97-011-02:on 970822,operation Was Noted Outside Design Bases for ECCS & CSP for Switchover to Recirculation Sump Suction.Caused by Ineffective Change Mgt.Revised Procedure for Switchover 01(02) Ohp 4023.ES-1.3
1999-09-17
[Table view]
Category:RO)
MONTHYEARML17335A5641999-10-18018 October 1999
LER 99-024-00:on 990708,literal TS Requirements Were Not Met by Accumlator Valve Surveillance.Caused by Misjudgement Made in Conversion from Initial DC Cook TS to W Std Ts.Submitted License Amend Request.With 991018 Ltr
ML17335A5531999-10-0707 October 1999
LER 99-023-00:on 990907,inadequate TS Surveillance Testing of ESW Pump ESF Response Time Noted.Caused by Inadequate Understanding of Plant Design Basis.Surveillance Tests Will Be Revised & Implemented
ML17326A1291999-09-17017 September 1999
LER 99-022-00:on 990609,electrical Bus Degraded Voltage Setpoints Too Low for Safety Related Loads,Was Discovered. Caused by Lack of Understanding of Design of Plant.No Immediate Corrective Actions Necessary
ML17326A1121999-08-27027 August 1999
LER 99-021-00:on 990728,determined That GL 96-01 Test Requirements Were Not Met in Surveillance Tests.Caused by Failure to Understand Full Extent of GL Requirements. Surveillance Procedures Will Be Revised or Developed
ML17326A1011999-08-26026 August 1999
LER 99-020-00:on 990727,EDGs Were Declared Inoperable.Caused by Inadequate Protection of Air Intake,Exhaust & Room Ventilation Structures from Tornado Missile Hazards. Implemented Compensatory Measures in Form of ACs
ML17326A0911999-08-16016 August 1999
LER 99-019-00:on 990716,noted Victoreen Containment Hrrms Not Environmentally Qualified to Withstand post-LOCA Conditions.Caused by Inadequate Design Control.Reviewing Options to Support Hrrms Operability in Modes 1-4
ML17326A0771999-08-0404 August 1999
LER 98-029-01:on 980422,noted That Fuel Handling Area Ventilation Sys Was Inoperable.Caused by Original Design Deficiency.Radiological Analysis for Spent Fuel Handling Accidents in Auxiliary Bldg Will Be Redone by 990830
ML17326A0741999-07-29029 July 1999
LER 99-018-00:on 990629,determined That Valve Yokes May Yield Under Combined Stress of Seismic Event & Static,Valve Closed,Stem Thrust.Caused by Inadequate Design of Associated Movs.Operability Determinations Were Performed for Valves
ML17326A0661999-07-26026 July 1999
LER 99-017-00:on 990625,noted That Improperly Installed Fuel Oil Return Relief Valve Rendered EDG Inoperable.Caused by Personnel Error.Fuel Oil Return Valve Was Replaced with Valve in Correct Orientation.With 990722 Ltr
ML17326A0651999-07-22022 July 1999
LER 98-014-03:on 980310,noted That Response to high-high Containment Pressure Procedure Was Not Consistent with Analysis of Record.Caused by Inadequate Interface with W. FRZ-1 Will Be Revised to Be Consistent with New Analysis
ML17326A0491999-07-13013 July 1999
LER 99-016-00:on 990615,TS Requirements for Source Range Neutron Flux Monitors Not Met.Caused by Failure to Understand Design Basis of Plant.Procedures Revised.With 990713 Ltr
ML17326A0331999-07-0101 July 1999
LER 99-004-01:on 971030,failure to Perform TS Surveillance Analyses of Reactor Coolant Chemistry with Fuel Removed Was Noted.Caused by Ineffective Mgt of Tss.Chemistry Personnel Have Been Instructed on Requirement to Follow TS as Written
ML17326A0151999-06-18018 June 1999
LER 99-014-00:on 990521,determined That Boron Injection Tank Manway Bolts Were Not Included in ISI Program,Creating Missed Exam for Previous ISI Interval.Caused by Programmatic Weakness.Isi Program & Associated ISI Database Modified
ML17325B6311999-06-0101 June 1999
LER 99-S03-00:on 990430,vital Area Barrier Degradation Was Noted.Caused by Inadequate Insp & Maint of Vital Area Barrier.Repairs & Mods Were Made to Barriers to Eliminate Degraded & Nonconforming Conditions
ML17325B6421999-06-0101 June 1999
LER 99-013-00:on 990327,safety Injection & Centrifugal Charging Throttle Valve Cavitation During LOCA Could Have Led to ECCS Pump Failure.Caused by Inadequate Original Design Application of Si.Throttle Valves Will Be Developed
ML17325B6351999-05-28028 May 1999
LER 99-S02-00:on 990428,vulnerability in Safeguard Sys That Could Allow Unauthorized Access to Protected Area Was Noted. Caused by Inadequate Original Plant Design.Mods Were Made to Wall Opening to Eliminate Nonconforming Conditions
ML17265A8231999-05-24024 May 1999
LER 98-037-01:on 990422,determined That Ice Condenser Bypass Leakage Exceeds Design Basis Limit.Caused by Pressure Seal Required by Revised W Design Not Incorporated Into Aep Design.Numerous Matl Condition Walkdowns & Assessments Made
ML17325B6001999-05-20020 May 1999
LER 99-012-00:on 990420,concluded That Auxiliary Bldg ESF Ventilation Sys Not Capable of Maintaining ESF Room Temps post-accident.Caused by Inadequate Control of Sys Design Inputs.Comprehensive Action Plan Being Developed
ML17325B5861999-05-10010 May 1999
LER 99-002-00:on 990415,discovered That TS 4.0.5 Requirements Were Not Met Due to Improperly Performed Test. Caused by Incorrect Interpretation of ASME Code.App J Testing Will Be Completed & Procedures Will Be Revised
ML17325B5811999-05-0404 May 1999
LER 99-011-00:on 990407,air Sys for EDG Will Not Support Long Operability.Caused by Original Design Error.Temporary Mod to Supply Makeup Air Capability in Modes 5 & 6 Was Prepared
ML17325B5771999-05-0303 May 1999
LER 99-010-00:on 990401,RCS Leak Detection Sys Sensitivity Not in Accordance with Design Requirements Occurred.Caused by Inadequate Original Design of Containment Sump Level. Evaluation Will Be Performed to Clearly Define Design
ML17325B5581999-04-16016 April 1999
LER 99-006-00:on 990115,personnel Identified Discrepancy Between TS 3.9.7 Impact Energy Limit & Procedure 12 Ohp 4030.STP.046.Caused by Lack of Design Basis Control.Placed Procedure 12 Ohp 4030.STP.046 on Administrative Hold
ML17325B5471999-04-12012 April 1999
LER 99-009-00:on 990304,as-found RHR Safety Relief Valve Lift Setpoint Greater than TS Limit Occurred.Cause Investigation for Condition Has Not Been Completed.Update to LER Will Be Submitted,Upon Completion of Investigation
ML17325B5321999-04-0707 April 1999
LER 99-S01-00:on 990308,discovered That Lock for Vital Gate Leading to Plant 4KV Switchgear Area Was Nonconforming & Vulnerable to Unauthorized Access.Caused by Inadequate Gate Design & Inadequate Procedures.Mods Are Being Made to Gate
ML17325B5161999-04-0101 April 1999
LER 99-007-00:on 981020,calculations Showed That Divider Barrier Between Upper & Lower Containment Vols Were Overstressed.Engineers Are Currently Working on Analyses of Loads & Stress on Enclosures
ML17325B5221999-03-29029 March 1999
LER 99-001-00:on 960610,degraded Component Cooling Water Flow to Containment Main Steam Line Penetrations,Identified on 990226.Caused by Inadequate Understanding of Design Basis.Additional Investigations Ongoing
ML17325B4801999-03-18018 March 1999
LER 99-004-00:on 971030,failure to Perform TS Surveillance Analyses of Rc Chemistry with Fuel Removed Was Noted.Cause of Event Is Under Investigation.Corrected Written Job Order Activities Used to Control SD Chemistry Sampling
ML17325B4741999-03-18018 March 1999
LER 99-005-00:on 940512,determined That Rt Breaker Manual Actuations During Rod Drop Testing Were Not Previously Reported.Caused by Lack of Training.Addl Corrective Actions,Including Preventative Actions May Be Developed
ML17325B4571999-02-24024 February 1999
LER 99-003-00:on 990107,CR Pressurization Sys Surveillance Test Did Not Test Sys in Normal Operating Condition.Caused by Failure to Recognize Door 12DR-AUX415 as Part of CR Pressure Boundary.Performed Walkdown of Other Doors
ML17335A5171999-02-11011 February 1999
LER 99-002-00:on 990112,determined That RCS Pressurizer PORVs Had Not Been Tested,Per Ts.Caused by Inadequate Scheduling Controls Allowing Personnel Error.Surveillance Procedure Was Completed & Updated LER Will Be Submitted
ML17335A5141999-02-10010 February 1999
LER 99-001-00:on 990106,noted That GE Hfa Relays Installed in EDGs May Not Meet Seismic Qualification.Caused by Operating Experience Info Incorrectly Dispositioned in 1985. Updated LER Will Be Submitted by 990405
ML17335A5011999-02-0101 February 1999
LER 98-060-00:on 981231,identified That Rt Sys Response Time Testing Did Not Comply with TS Definition.Caused by Inadequate Procedures.Corrective Actions Will Be Developed & Update to LER Will Be Submitted by 990415.With 990201 Ltr
ML17335A4951999-01-29029 January 1999
LER 98-059-00:on 981230,interim LER -single Failure in Containment Spray Sys Could Result in Containment Spray Ph Outside Design Occurred.Investigation Into Condition Continuing.Update Will Be Submitted by 990514 Ltr
ML17335A4961999-01-27027 January 1999
LER 98-057-00:on 981228,discovered That AFW Valves Were Not Tested IAW Inservice Testing Program.Caused by Failure to Recognize Design Bases Features Re Afws by Personnel. Updated LER Will Be Submittted by 990415.With 990127 Ltr
ML17335A4921999-01-19019 January 1999
LER 98-052-01:on 981128,no Analysis for NSR Sc Manual Loader for Tdafwp Could Be Found in Original Design.Cause Due to All Failure Modes Not Considered When Compressed Air Sys Originally Designed.Performed Review of Components
ML17335A4721999-01-0606 January 1999
LER 98-055-00:on 981207,potential for Condition Outside of Design Bases for Rod Control Sys Was Noted.Caused by Calibration Error Coupled with Single Rod Failure.Condition Rept Investigation Is Ongoing
ML17335A4691999-01-0606 January 1999
LER 98-056-00:on 981211,hot Leg Nozzle Gaps Resulted in Plant Being in Unanalyzed Condition.Analyses Are Being Performed by W to Resolve Problem.Updated LER Will Be Submitted by 990211.With 990106 Ltr
ML17335A4661999-01-0505 January 1999
LER 98-049-00:on 981020,emergency Boron Injection Flow Path Was Inoperable.Caused by Original Design Deficiency. Engineering Evaluation of Event Is Continuing
ML17335A4631999-01-0404 January 1999
LER 98-054-00:on 981202,discovered That at Least One MSSV Had Not Been Reset as Required by Ts.Engineering Is Continuing Review of Extent of Condition for Event.Updated LER Will Be Submitted by 990129.With 990104 Ltr
ML17335A4481998-12-30030 December 1998
LER 98-053-00:on 981130,discovered Use of Inoperable Substitute Subcooling Margin Monitor.Caused by Condition Existing Since Installation of Plant Process Computer in 1992.Updated LER Will Be Submitted.With 981230 Ltr
ML17335A4581998-12-28028 December 1998
LER 98-052-00:on 981128,turbine Driven AFW Pump Speed Controller Failure Mode Occurred.Caused Because Not All Failure Modes Were Considered When Compressed Air Sys Was Originally Designed.Verified Current Design Change Process
ML17335A4281998-12-22022 December 1998
LER 98-051-00:on 981122,reactor Trip Signal from Manual Safety Injection Not Verified as Required by TS Surveillance,Was Discovered.Maintenance Currently Evaluating Significance & Cause of Event
ML17335A4111998-12-17017 December 1998
LER 98-047-00:on 981117,potential for Increase Leakage from Reactor Coolant Pump Seals Was Identified.Util Is Working with W to Resolve Issue.Current Expectations Are to Submit Update to LER by 990215.With 981217 Ltr
ML17335A4141998-12-16016 December 1998
LER 98-058-00:on 981216,postulated High Line Break Could Result in Condition Outside Design Bases for AF Occurred. Caused by Deficiencies Associated with Administration of HELB Program.Analysis of AF Will Be Completed by 990122
ML17335A4181998-12-16016 December 1998
LER 98-050-00:on 980814,ancillary Equipment Installed in Ice Condenser Was Not Designed to Withstand Design Basis Accident/Earthquake Loads.Caused by Lack of Established Design Criteria.Developed Design Criteria
ML17335A3871998-12-11011 December 1998
LER 98-031-01:on 980610,potential Common Mode Failure of RHR Pumps Were Noted.Caused by Inaccurate Values.Accurate Miniflow Numbers Have Been Determined by Flow Testing
ML17335A3821998-12-0808 December 1998
LER 98-039-01 Re EOP Step Conflicts with Small Break LOCA Analysis.Ler 98-039-00 Has Been Canceled.With 981208 Ltr
ML17335A3781998-12-0707 December 1998
LER 98-007-00:on 981106,high Energy Line Break Effects in Auxiliary FW Sys Was Noted.Cause of Event Is Under Investigation & Will Be Completed by 990220.Updated LER Will Be Submitted by 990310.With 981207 Ltr
ML17335A3771998-12-0303 December 1998
LER 98-046-00:on 981103,determined That Afs Was Unable to Meet Design Flow Requirements During Special Test.Caused by Failure to Consider All Aspects of Sys Operation in Design of Suction Basket Strainers.Sys Will Be Redesigned
ML17335A3741998-12-0202 December 1998
LER 97-011-02:on 970822,operation Was Noted Outside Design Bases for ECCS & CSP for Switchover to Recirculation Sump Suction.Caused by Ineffective Change Mgt.Revised Procedure for Switchover 01(02) Ohp 4023.ES-1.3
1999-09-17
[Table view]
Category:TEXT-SAFETY REPORT
MONTHYEARML17335A5641999-10-18018 October 1999
LER 99-024-00:on 990708,literal TS Requirements Were Not Met by Accumlator Valve Surveillance.Caused by Misjudgement Made in Conversion from Initial DC Cook TS to W Std Ts.Submitted License Amend Request.With 991018 Ltr
ML17335A5531999-10-0707 October 1999
LER 99-023-00:on 990907,inadequate TS Surveillance Testing of ESW Pump ESF Response Time Noted.Caused by Inadequate Understanding of Plant Design Basis.Surveillance Tests Will Be Revised & Implemented
ML17335A5631999-09-30030 September 1999
Monthly Operating Rept for Sept 1999 for DC Cook Nuclear Plant,Unit 1.With 991012 Ltr
ML17335A5621999-09-30030 September 1999
Monthly Operating Rept for Sept 1999 for DC Cook Nuclear Plant,Unit 2.With 991012 Ltr
ML17335A5481999-09-30030 September 1999
Non-proprietary DC Cook Nuclear Plant Units 1 & 2 Mods to Containment Sys W SE (Secl 99-076,Rev 3).
ML17335A5451999-09-28028 September 1999
Rev 1 to Containment Sump Level Design Condition & Failure Effects Analysis for Potential Draindown Scenarios.
ML17326A1291999-09-17017 September 1999
LER 99-022-00:on 990609,electrical Bus Degraded Voltage Setpoints Too Low for Safety Related Loads,Was Discovered. Caused by Lack of Understanding of Design of Plant.No Immediate Corrective Actions Necessary
ML17326A1481999-09-17017 September 1999
Independent Review of Control Rod Insertion Following Cold Leg Lbloca,Dc Cook,Units 1 & 2.
ML17326A1211999-08-31031 August 1999
Monthly Operating Rept for Aug 1999 for Cook Nuclear Plant, Unit 2.With 990915 Ltr
ML17326A1201999-08-31031 August 1999
Monthly Operating Rept for Aug 1999 for Cook Nuclear Plant, Unit 1.With 990915 Ltr
ML17326A1121999-08-27027 August 1999
LER 99-021-00:on 990728,determined That GL 96-01 Test Requirements Were Not Met in Surveillance Tests.Caused by Failure to Understand Full Extent of GL Requirements. Surveillance Procedures Will Be Revised or Developed
ML17326A1011999-08-26026 August 1999
LER 99-020-00:on 990727,EDGs Were Declared Inoperable.Caused by Inadequate Protection of Air Intake,Exhaust & Room Ventilation Structures from Tornado Missile Hazards. Implemented Compensatory Measures in Form of ACs
ML17326A0911999-08-16016 August 1999
LER 99-019-00:on 990716,noted Victoreen Containment Hrrms Not Environmentally Qualified to Withstand post-LOCA Conditions.Caused by Inadequate Design Control.Reviewing Options to Support Hrrms Operability in Modes 1-4
ML17326A0771999-08-0404 August 1999
LER 98-029-01:on 980422,noted That Fuel Handling Area Ventilation Sys Was Inoperable.Caused by Original Design Deficiency.Radiological Analysis for Spent Fuel Handling Accidents in Auxiliary Bldg Will Be Redone by 990830
ML17335A5461999-08-0202 August 1999
Rev 0 to Evaluation of Cook Recirculation Sump Level for Reduced Pump Flow Rates.
ML17326A0871999-07-31031 July 1999
Monthly Operating Rept for July 1999 for DC Cook Nuclear Plant,Unit 1.With 990812 Ltr
ML17326A0861999-07-31031 July 1999
Monthly Operating Rept for July 1999 for DC Cook Nuclear Plant,Units 2.With 990812 Ltr
ML17326A0741999-07-29029 July 1999
LER 99-018-00:on 990629,determined That Valve Yokes May Yield Under Combined Stress of Seismic Event & Static,Valve Closed,Stem Thrust.Caused by Inadequate Design of Associated Movs.Operability Determinations Were Performed for Valves
ML17326A0661999-07-26026 July 1999
LER 99-017-00:on 990625,noted That Improperly Installed Fuel Oil Return Relief Valve Rendered EDG Inoperable.Caused by Personnel Error.Fuel Oil Return Valve Was Replaced with Valve in Correct Orientation.With 990722 Ltr
ML17326A0651999-07-22022 July 1999
LER 98-014-03:on 980310,noted That Response to high-high Containment Pressure Procedure Was Not Consistent with Analysis of Record.Caused by Inadequate Interface with W. FRZ-1 Will Be Revised to Be Consistent with New Analysis
ML17326A0491999-07-13013 July 1999
LER 99-016-00:on 990615,TS Requirements for Source Range Neutron Flux Monitors Not Met.Caused by Failure to Understand Design Basis of Plant.Procedures Revised.With 990713 Ltr
ML17326A0331999-07-0101 July 1999
LER 99-004-01:on 971030,failure to Perform TS Surveillance Analyses of Reactor Coolant Chemistry with Fuel Removed Was Noted.Caused by Ineffective Mgt of Tss.Chemistry Personnel Have Been Instructed on Requirement to Follow TS as Written
ML17326A0511999-06-30030 June 1999
Monthly Operating Rept for June 1999 for DC Cook Nuclear Plant,Unit 2.With 990709 Ltr
ML17326A0501999-06-30030 June 1999
Monthly Operating Rept for June 1999 for DC Cook Nuclear Plant,Unit 1.With 990709 Ltr
ML17326A0151999-06-18018 June 1999
LER 99-014-00:on 990521,determined That Boron Injection Tank Manway Bolts Were Not Included in ISI Program,Creating Missed Exam for Previous ISI Interval.Caused by Programmatic Weakness.Isi Program & Associated ISI Database Modified
ML17325B6421999-06-0101 June 1999
LER 99-013-00:on 990327,safety Injection & Centrifugal Charging Throttle Valve Cavitation During LOCA Could Have Led to ECCS Pump Failure.Caused by Inadequate Original Design Application of Si.Throttle Valves Will Be Developed
ML17325B6311999-06-0101 June 1999
LER 99-S03-00:on 990430,vital Area Barrier Degradation Was Noted.Caused by Inadequate Insp & Maint of Vital Area Barrier.Repairs & Mods Were Made to Barriers to Eliminate Degraded & Nonconforming Conditions
ML17326A0061999-05-31031 May 1999
Monthly Operating Rept for May 1999 for Dcp.With 990609 Ltr
ML17326A0071999-05-31031 May 1999
Monthly Operating Rept for May 1999 for DC Cook Nuclear Plant,Unit 2.With 990609 Ltr
ML17325B6351999-05-28028 May 1999
LER 99-S02-00:on 990428,vulnerability in Safeguard Sys That Could Allow Unauthorized Access to Protected Area Was Noted. Caused by Inadequate Original Plant Design.Mods Were Made to Wall Opening to Eliminate Nonconforming Conditions
ML17265A8231999-05-24024 May 1999
LER 98-037-01:on 990422,determined That Ice Condenser Bypass Leakage Exceeds Design Basis Limit.Caused by Pressure Seal Required by Revised W Design Not Incorporated Into Aep Design.Numerous Matl Condition Walkdowns & Assessments Made
ML17325B6001999-05-20020 May 1999
LER 99-012-00:on 990420,concluded That Auxiliary Bldg ESF Ventilation Sys Not Capable of Maintaining ESF Room Temps post-accident.Caused by Inadequate Control of Sys Design Inputs.Comprehensive Action Plan Being Developed
ML17325B5861999-05-10010 May 1999
LER 99-002-00:on 990415,discovered That TS 4.0.5 Requirements Were Not Met Due to Improperly Performed Test. Caused by Incorrect Interpretation of ASME Code.App J Testing Will Be Completed & Procedures Will Be Revised
ML17325B5811999-05-0404 May 1999
LER 99-011-00:on 990407,air Sys for EDG Will Not Support Long Operability.Caused by Original Design Error.Temporary Mod to Supply Makeup Air Capability in Modes 5 & 6 Was Prepared
ML17325B5771999-05-0303 May 1999
LER 99-010-00:on 990401,RCS Leak Detection Sys Sensitivity Not in Accordance with Design Requirements Occurred.Caused by Inadequate Original Design of Containment Sump Level. Evaluation Will Be Performed to Clearly Define Design
ML17335A5301999-04-30030 April 1999
Monthly Operating Rept for Apr 1999 for DC Cook Nuclear Plant,Unit 1.With 990508 Ltr
ML17335A5291999-04-30030 April 1999
Monthly Operating Rept for Apr 1999 for DC Cook Nuclear Plant,Unit 2.With 990508 Ltr
ML17325B5581999-04-16016 April 1999
LER 99-006-00:on 990115,personnel Identified Discrepancy Between TS 3.9.7 Impact Energy Limit & Procedure 12 Ohp 4030.STP.046.Caused by Lack of Design Basis Control.Placed Procedure 12 Ohp 4030.STP.046 on Administrative Hold
ML17325B5471999-04-12012 April 1999
LER 99-009-00:on 990304,as-found RHR Safety Relief Valve Lift Setpoint Greater than TS Limit Occurred.Cause Investigation for Condition Has Not Been Completed.Update to LER Will Be Submitted,Upon Completion of Investigation
ML17325B5321999-04-0707 April 1999
LER 99-S01-00:on 990308,discovered That Lock for Vital Gate Leading to Plant 4KV Switchgear Area Was Nonconforming & Vulnerable to Unauthorized Access.Caused by Inadequate Gate Design & Inadequate Procedures.Mods Are Being Made to Gate
ML17325B5161999-04-0101 April 1999
LER 99-007-00:on 981020,calculations Showed That Divider Barrier Between Upper & Lower Containment Vols Were Overstressed.Engineers Are Currently Working on Analyses of Loads & Stress on Enclosures
ML17325B5491999-03-31031 March 1999
Monthly Operating Rept for Mar 1999 for DC Cook Nuclear Plant Unit 2.With 990408 Ltr
ML17325B5441999-03-31031 March 1999
Monthly Operating Rept for Mar 1999 for DC Cook Nuclear Plant,Unit 1.With 990408 Ltr
ML17325B5221999-03-29029 March 1999
LER 99-001-00:on 960610,degraded Component Cooling Water Flow to Containment Main Steam Line Penetrations,Identified on 990226.Caused by Inadequate Understanding of Design Basis.Additional Investigations Ongoing
ML17325B4801999-03-18018 March 1999
LER 99-004-00:on 971030,failure to Perform TS Surveillance Analyses of Rc Chemistry with Fuel Removed Was Noted.Cause of Event Is Under Investigation.Corrected Written Job Order Activities Used to Control SD Chemistry Sampling
ML17325B4741999-03-18018 March 1999
LER 99-005-00:on 940512,determined That Rt Breaker Manual Actuations During Rod Drop Testing Were Not Previously Reported.Caused by Lack of Training.Addl Corrective Actions,Including Preventative Actions May Be Developed
ML17325B5671999-03-0202 March 1999
Summary of Unit 1 Steam Generator Layup Chemistry from 980101 to 990218.
ML17325B4631999-02-28028 February 1999
Monthly Operating Rept for Feb 1999 for DC Cook Nuclear Power Station,Unit 2.With 990308 Ltr
ML17325B4621999-02-28028 February 1999
Monthly Operating Rept for Feb 1999 for DC Cook Nuclear Plant,Unit 1.With 990308 Ltr
ML17325B4571999-02-24024 February 1999
LER 99-003-00:on 990107,CR Pressurization Sys Surveillance Test Did Not Test Sys in Normal Operating Condition.Caused by Failure to Recognize Door 12DR-AUX415 as Part of CR Pressure Boundary.Performed Walkdown of Other Doors
1999-09-30
[Table view] |
Text
NRC FORM 366 U.S. NUCLEAR REGULAT COMMISSION APPROV MB NO. 3150-0104 EXPIRES 06/30/2001 ie-1999)
Estimated burden per response to comply with this mandatory information LICENSEE EVENT REPORT (LER) cogection request: 50 hrs. Reported lessons learned are incorporated into the licensing process and fed back to industry. Forward comments regarding burden estimate to the Records Management Branch (T4 F33), U.S. Nuclear Regutatory (See reverse for required number of Commission. Washington, DC 205554&1, and to the PapenNork Reduction digits/characters, for each block) Project (3t504104), Oirice of Management and Budget. Washington. DC 20503.
lf an information collection does not display a currently valrd OMB control number. the NRC may not conduct or sponsor, and a person is not required to respond to. the information collection.
FACILITYNAME (1) DOCKET NUMBER I2) PAGE I3)
Cook Nuclear Plant Unit 1 05000-315 1OF4 TITLE t4)
Inadequate Technical Specification Surveillance Testing of Essential Service Water Pump Engineered Safety Feature Response Time EVENT DATE {5) LER NUMBER {6) REPORT DATE (7) OTHER FACILITIES INVOLVED (8)
MONTH DAY YEAR YEAR SEQUENTIAL REVISION MONTH DAY YEAR FACILITY NAME DOCKET NUMBER NUMBER NUMBER 1999 FACIUIYNAME DOCKET NUMBER OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR 5: (Check one or moro) {11)
MODE (9) 20.2201 (b) 20.2203(a) (2)(v) 50.73(a)(2)(i) 50.73(a) (2) {viii)
POWER 20.2203(a) (1) 20.2203(a)(3)(I) 50.73(a)(2)(ii) 50.73(a) (2)(x)
LEVEL (10) 20.2203(a)(2) (i) 20.2203(a)(3)(ii) 50.73(a)(2)(iii) 73.71 20.2203(a)(2)(ii) 20.2203(a) (4) 50.73(a)(2)(iv) OTHER 20.2203(a)(2)(iii) 50.36(c) {1) 50.73(a)(2) (v) Specify in Abstract below or In NRC Form 366A 20.2203(a) {2)(iv) 50.36(c)(2) 50.73(a) (2)(vii)
LICENSEE CONTACT FOR THIS LER {12)
NAME TELEPHONE NUMBER IInrSude Area Code)
Mary Beth Depuydt, Regulatory Compliance (616) 465-5901 X1 589 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13) cAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE To EPIX To EPIX SUPPLEMENTAL REPORT EXPECTED 14 EXPECTED MONTH DAY YEAR YEs NO
{Ifyes, complete EXPECTED SUBMISSION DATE).
ABSTRACT (Limit to 1400 spaces, i.o., approximately 15 single-spaced typewritten linos) (16)
On June 24, 1999, during a review of the findings of Expanded System Readiness Review (ESRR) team by Performance Assurance (PA), it was discovered that no testing program could be identified which verifies the capability of the Essential Service Water (ESW) pumps to meet the Engineered Safety Feature (ESF) response time speciTied in the Technical Specifications (TS) or the Updated Safety Analysis Report. Subsequent investigation confirmed that in-place TS surveillance testing measured the ESF response time for the ESW pumps as the elapsed time from actuation of the channel sensor until pump breaker closure, but did not include the time until a specified pump discharge pressure is reached or until the ESW pump discharge valve is open; as required by the definition of Engineered Safety Feature Response Time.
Since existing surveillance testing did not satisfy the TS definition of ESF response time, the identified condition constitutes a missed surveillance test. This is an operation or condition prohibited by TS and was determined to be reportable pursuant to the requirements of 10CFR50.73(a)(2)(i)(B) on September 7, 1999. The apparent cause of this event was the inadequate understanding of the plant design basis. Surveillance tests will be revised and implemented to include the time to achieve prescribed pump discharge pressure/flow and/or discharge valve position as part of the overall ESF response time testing for the ESW system prior to restart of each respective unit. The ESW ESF response times in UFSAR Table 7.2-7 will be evaluated and revised, if necessary, prior to restart of each respective unit. ESW system performance records and surveillance test results provide reasonable assurance that the system has remained capable of performing its interlded functions. Therefore, there were minimal safety implications to the health and safety of the public as a result of this event.
'I)9i0i30i94 99i007 PDR ADGCI{l 050003i5 S PDR
NRC FORM 366A U.S. NUCLEAR RE TORY COMMISSION l+,16.1998)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME I1) DocKET I2) LER NUMBER (6) PAGE I3)
Cook Nuclear Plant Unit 1 05000-315 "EAR SEQUENTIAL NUMBER REVISION 2 QF 4 NUMBER 1999 023 00 TEXT llfmore spaceis required, use additional copies of NRC Form 366A J )17)
CONDITIONS PRIOR TO EVENT Unit 1 was defueled Unit 2 was defueled DESCRIPTION OF THE EVENT On June 24, 1999, during a review of the findings of Expanded System Readiness Review (ESRR) team by Performance Assurance (PA), it was documented that no testing program could be identified which verifies the capability of the Essential Service Water (ESW) pumps to meet the Engineered Safety Feature (ESF) response time specified in the Technical Specifications or the Updated Safety Analysis Report. Subsequent investigation of this condition by Engineering, completed September 1, 1999, confirmed that the acceptance criteria for in-place Technical Specification surveillance testing defined the ESF response time for the ESW pumps as the elapsed time from actuation of the channel sensor until pump breaker closure. Testing did not include the time until a specified pump discharge pressure is reached or the ESW pump discharge valve is open, as required by the definition of Engineered Safety Feature Response Time. The Technical Specification (TS) and UFSAR definition of Engineered Safety Feature Response Time. is that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor to until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Since existing surveillance testing did not satisfy the TS definition of ESF response time, the identified condition constitutes a missed surveillance test.
CAUSE OF THE EVENT The apparent cause of this event was inadequate understanding of the design basis of the plant. During the development of the ESW ESF response times, the design basis requirements for ESW availability during an accident were inadequately understood. This resulted in surveillance procedures for ESW which did not satisfy the UFSAR and Technical Specification definition of ESF response time.
ANALYSIS OF THE EVENT The Technical Specification (TS) and UFSAR definition of Engineered'Safety Feature Response Time is that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor to until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Since existing surveillance testing did not satisfy the TS definition of ESF response time, the identified condition constitutes a missed surveillance test. This is an operation or condition prohibited by TS and was determined to be reportable pursuant to the requirements of 10CFR50.73(a)(2)(i)(B) on September 7, 1999.
Response times for Engineered Safety Features are provided in the UFSAR, Section 7.2, Table 7.2-7. The ESF Response Time Basis Procedure specifies the strategy used at Cook Nuclear Plant to demonstrate the operability of various Engineered Safety Features, systems and sub-systems. This procedure defines "Device Response Time as the time from Safeguards Master Relay closing until the component reaches its ESF position. Additionally, "ESF Response Time" is defined as the time interval from when the monitored parameter exceeds its ESF actuation se'tpoint at the channel sensor until the ESF equipment is capable of performing it's safety function.
Technical Specifications Surveillance Requirements for ESF response time in Section 4.3.2.1.3 and Table 3.3-3 specify that each Engineered Safety. Feature Actuation Signal (ESFAS) function will be demonstrated to be within limits at least once per 18 moriths.
Review of Emergency Diesel Generator Load Sequencing and ESF Testing revealed the ESF response time for the ESW NRC FORM 366A I6-1998)
NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION i6-1999)
"
LICENSEE EVENT REPORT tLER)
TEXT CONTINUATION FACIUTY NAME I1) DOCKET {2) LER NUMBER I6) PAGE I3)
Cook Nuclear Plant Unit 1 05000-315 YEAR SEQUENTIAL NUMBER REYIsI0N 3 OF 4 NUMBER 1999 023 00 TEXT (Ifmore spaceis required, use additional copies of NRC Form 366AJ {17) pumps is measured from the initiating sensor channel to the pump breaker closure. Testing does not include the time for the pump to reach the required discharge pressure or for the ESW pump discharge valve to open.
In early 1975, operational problems identified with the ESW system, including severe water hammer at pump start-up, lead to testing being performed under various operational transients. This testing did not result in a significant water hammer, however, a previous test and operating experience showed that the water hammer did not occur when'an idle pump was started with a throttled discharge valve even though its header had not been pressurized for as long as twelve hours. Determination was made that the water hammers were induced upon the start of an idle ESW pump with a fully open discharge valve even when the header had been depressurized for no more than a few minutes. This determination lead to modification of the design of the ESW pump discharge valves, such that the valves remain closed when the ESW pump is idle, and are interlocked to open on ESW pump start at breaker closure.
Response times for sensor actuation to ESW pump breaker closure and ESW pump discharge MOV stroke times are measured under the Surveillance Test Program. However, these times are not combined to provide an overall ESF response time which meets the TS definition and which is compared to an acceptance criteria. The ESF response time test procedure was reviewed to verify that ESF pumps other than ESW are tested from pump start to required system pressure/flow. Each was verified to include requirements to measure the overall response time from sensor actuation until an acceptable discharge pressure or flow prescribed by acceptance criteria.
Although the ESW ESF response times are included in UFSAR Table 7.2-7, ESW response times are not explicitly included in the UFSAR Chapter 14.0 accident analysis assumptions. ESW is not immediately required to support the containment spray system (CTS) and Emergency Diesel Generator during a design basis Loss of Coolant Accident (LOCA). ESW system performance records and surveillance test results provide reasonable assurance that the system has remained capable of performing its intended functions. Based upon the above information, there were minimal safety implications to the health and safety of the public as a result of this event.
CORRECTIVE ACTIONS Surveillance tests will be revised and implemented to include the time to achieve prescribed pump discharge pressure/flow and/or discharge valve position as part of the overall ESF response time testing for the ESW system prior to restart of each respective unit.
The ESW ESF response times in UFSAR Table 7.2-7 will be evaluated and revised, if necessary, prior to restart of each respective unit.
As discussed in letter AEP:NRC:1260GH, "Donald C. Cook Nuclear Power Plant, Units 1 and 2, Enforcement Actions98-150, 98-151,98-152 and 98-156, Reply To Notice Of Violation Dated October 13, 1998," dated March 19, 1999, a surveillance program owner and manager position has been established, reporting to the Work Control Director. A Leadership Plan has been developed which includes the creation of a detailed surveillance data base to align surveillance requirements to specific implementing procedures and a comprehensive adequacy review of surveillance testing procedures.
As previously discussed in LER 315/99-021-00 and as part of Restart Action Plan 0 0001 for the Programmatic Breakdown in Surveillance Testing, the adequacy of the TS surveillance program will be evaluated. This evaluation includes verification that TS surveillance requirements for all modes of plant operation are incorporated into TS surveillance test procedures. Also, as part of the Restart effort, System and programmatic assessments in the Expanded System Readiness Reviews and Licensing Basis Reviews are reestablishing and documenting the plant's Design and Licensing Basis.
NRC FORM 366A i6-1999)
NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION I6-1998)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME I1) DOCKET I2) LER NUMBER I6) PAGE I3)
Cook Nuclear'lant Unit 1 05000-315 YEAR SEQUENTIAL NUMBER REVISION 4 OF 4 NUMBER 1999 023 00 TEXT (lfmore spaco is required, uso additional copies of NRC Form 386A/ I 17)
SIMILAR EVENTS 315/99-010-00 315/99-015-00 315/99-016-00 315/99-021-00 NRC FORM 366A I6.1998)
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