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| | issue date = 07/22/1999 | | | issue date = 07/22/1999 |
| | title = 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 | | | title = 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 |
| | author name = DEPUYDT M B | | | author name = Depuydt M |
| | author affiliation = INDIANA MICHIGAN POWER CO. | | | author affiliation = INDIANA MICHIGAN POWER CO. |
| | addressee name = | | | addressee name = |
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| {{#Wiki_filter:NRC Form 366 U.S.NUCLEAR REGULATORY COMMISSION (6-1 996)LICENSEE EVENT REPORT (LER)(See reverse for required number of digits/characters for each block)APPROVED 8Y OM8 NO.3160%104 EXPIRES 06/30/2001 ESTBIATED BURDEN PER RESPONSE TO COMPLY WITH THIS MANDA'tORY INFORMATION COLLECTION REQUEST: 500 HRS.REPORTED LESSONS LEARNED ARE INCORPORATED INTO THE UCENSUIQ PROCESS AND FED BACK TO INDUStRY.FORWARD COLTJENTS REQARINNQ BURDEN ESTUJATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH rt+F33), U.S.NUCLEAR REGULATORY COLBBSSION, WASHINGTON. | | {{#Wiki_filter:NRC Form 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED 8Y OM8 NO. 3160%104 EXPIRES 06/30/2001 (6-1 996) |
| DC 205554001. | | ESTBIATED BURDEN PER RESPONSE TO COMPLY WITH THIS MANDA'tORY INFORMATION COLLECTION REQUEST: 500 HRS. REPORTED LESSONS LEARNED ARE INCORPORATED INTO THE UCENSUIQ PROCESS AND FED BACK TO INDUStRY. |
| AND TO THE PAPERWORK REDUCTION PROJECT I3150oterx OFFICE OF MANAGEMENT AND BUDGEt.WASHINGTON. | | LICENSEE EVENT REPORT (LER) FORWARD COLTJENTS REQARINNQ BURDEN ESTUJATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH rt+ F33), U.S. NUCLEAR REGULATORY COLBBSSION, WASHINGTON. DC 205554001. AND TO THE PAPERWORK REDUCTION PROJECT I3150oterx OFFICE OF MANAGEMENT AND BUDGEt. WASHINGTON. DC (See reverse for required number of 20503 digits/characters for each block) |
| DC 20503 FACIUTY NAME II)TITLE I4)Cook Nuclear Plant Unit 1 DOCKET NUMBER I2)05000-315 PAGE I3)1 of3"Response to High-High Containment Pressure" Procedure Not Consistent with Analysis of Record EVENT DATE (6)LER NUMBER (6)REPORT DATE (7)OTHER FACILITIES INVOLVED (6)MONTH DAY YEAR YEAR SEQUENTIAL NUMBER REVISION NUMBER MONTH DAY FACIUTY NAM DC Cook-Unit 2 DOCKET NUMBER 05000-316 03 10 98 1998-014-03 07 22 1999 ACIU NAM DOCKET NUMB OPERATING MODE (9)POWER LEVEL (10)00 20.2201 (b)20.2203(a)(1) 20.2203(a)(2)(l) 20.2203(a)(2)(v) 20.2203(a)(3)(i) 20.2203(a)(3)(ii) 50.73(a)(2)(i) 50.73(a)(2)(li) 50.73(a)(2)(iii) 50.73(a)(2)(viii) 50.73(a)(2)(x) 73.71 THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR li: (Check one or more)(19)20.2203(a)(2)(ii) 20.2203(a)(2)(iii) 20.2203(a)(2)(iv) 20.2203(a)(4) 50.36(c)(1) 50.36(c)(2) | | FACIUTY NAME II) DOCKET NUMBER I2) PAGE I3) |
| LICENSEE CONTACT FOR THIS LER (12)50.73(a)(2)(lv) 50.73(a)(2)(v) 50.73(a)(2)(vii)
| | Cook Nuclear Plant Unit 1 05000-315 1 of3 TITLE I4) |
| OTHER Specfy in Abstract below or n NRC Form 388A Ms.M.B.Depuydt, Compliance Engineer TELEPHONE NUMBER trnc/ude 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 MANUFACTURER REPORTABLE TO EPIX SUPPLEMENTAL REPORT EXPECTED 14 YES If Yes, corn lets EXPECTED SUBMISSION DATE X NO EXPECTED SUBMISSION DATE 15 YEAR Abstract (Umit to 1400 spaces, l.e., approximately 15 single-spaced typewritten lines)(16)On March 10, 1998, with Units 1 and 2 in Mode 5, it was determined that both units had operated in an unanalyzed condition due to Functional Restoration Procedure FRZ-1,"Response to High-High Containment Pressure", not being consistent with the containment integrity analysis of record.Had the procedure been implemented, the potential existed for post-accident containment pressure to exceed its design basis limit of 12 psig.In accordance with 10CFR50.72(b)(2)(i), an ENS notification was made.This LER is therefore submitted in accordance with 10CFR50.73(a)(2)(ii)(A), for an unanalyzed condition, and 10CFR50.73(a)(2)(ii)(B), for a condition outside the design basis.The root cause of this condition was inadequate interface with Westinghouse regarding the assumptions used in the safety analysis.The procedure will be revised to direct initiation of RHR spray at the appropriate point to ensure that containment design pressure is not exceeded.A program will be established to identify, document and control key accident analyses assumptions, including those impacting the Emergency Operating Procedures.
| | "Response to High-High Containment Pressure" Procedure Not Consistent with Analysis of Record EVENT DATE (6) LER NUMBER (6) REPORT DATE (7) OTHER FACILITIES INVOLVED (6) |
| Additional actions will be taken to strengthen the communications between Operations and Engineering | | FACIUTY NAM DOCKET NUMBER SEQUENTIAL REVISION DC Cook- Unit 2 05000-316 MONTH DAY YEAR YEAR NUMBER NUMBER MONTH DAY ACIU NAM DOCKET NUMB 03 10 98 1998 014 03 07 22 1999 OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR li: (Check one or more) (19) |
| -Safety Analysis, which maintains oversight of vendors performing safety analyses that might impact actions taken by the operators. | | MODE (9) 20.2201 (b) 20.2203(a)(2)(v) 50.73(a)(2)(i) 50.73(a)(2)(viii) |
| Evaluation of this condition has been performed. | | POWER 20.2203(a)(1) 20.2203(a)(3)(i) 50.73(a)(2)(li) 50.73(a)(2)(x) |
| It has been concluded that containment pressure would have exceeded the design pressure of 12 psig, reaching a calculated peak of 13.85 psig.This value remained below the pre-operational structural integrity test value of 16.1 psig, therefore, it was concluded that the containment would have remained functional. | | LEVEL (10) 00 20.2203(a)(2)(l) 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)(lv) OTHER 20.2203(a)(2)(iii) 50.36(c)(1) 50.73(a)(2)(v) |
| '3)'3)07280073) 990722 PDR ADOCK 05000315 PDR NRC FORM 366A U.S.NUCLEAR REGULATORY COMMISSION (6-1 998)LICENSEE EVENT REPORT (LER)TEXT CONTINUATION FACIUTY NAME (1)Cook Nuclear Plant Unit 1 DOCKET NUMBER(2)05000-315 YEAR 1998 LER NUMBER (6)SEQUENTIAL NUMBER 014 REVISION NUMBER 03 PAGE (3)2 of 3 TEXT (if more space is required, use additional copies of NRC Form (366A)(17)Conditions Prior to Event Unit 1 was in Mode 5, Cold Shutdown Unit 2 was in Mode 5, Cold Shutdown Descri tion of Event On March 10, 1998, while performing a Containment Spray self assessment, it was determined that the actions directed by Functional Restoration Procedure 1,2-4023.OHP.FRZ-1,"Response to High-High Containment Pressure", were not consistent with the assumptions in the containment integrity analysis of record.Residual Heat Removal (RHR)(EIIS:BP)spray is designed to supplement the pressure mitigation function of containment spray during either a Loss of Coolant Accident (LOCA)or Main Steam Line Break (MSLB).In accordance with containment integrity analysis input assumptions, FRZ-1 directs that RHR spray be manually, initiated when containment pressure reaches 8 psig.The safety analysis did not make allowance for the time delay between containment pressure reaching 8 psig and the delivery of RHR spray to containment. | | Specfy in Abstract below 20.2203(a)(2)(iv) 50.36(c)(2) 50.73(a)(2)(vii) or n NRC Form 388A LICENSEE CONTACT FOR THIS LER (12) |
| This time delay results from the summation of the time required for the operator to recognize that containment pressure has reached 8 psig;for the RHR spray valves to open and RHR to Reactor Coolant system isolation valves to close/throttle; and for RHR flow to fill the spray line and spray headers.Had the use of this procedure in its current form been required, containment peak pressure mitigation would have been affected.Cause of Event This condition was the result of an inadequate interface with Westinghouse regarding the assumptions used in the safety analysis and how they were implemented at the plant.Equipment response times and operator action times were not included by Westinghouse when assumptions regarding RHR spray were incorporated into the analysis.Anal sis of Event This condition was determined to be reportable in accordance with 10CFR50.73(a)(2)(ii)(A), for an unanalyzed condition that significantly compromises plant safety, and 10CFR50.73(a)(2)(ii)(B) for a condition outside the design basis.The Emergency Core Cooling System (ECCS)is one of the Engineered Safety Feature systems, which mitigate the consequences of a major breach of the Reactor Coolant system (RCS), or main steam lines inside containment. | | TELEPHONE NUMBER trnc/ude Area Code) |
| The RCS line break results in a LOCA, during which the ECCS provides a significant volume of makeup to the RCS as well as core cooling and reactivity control.The LOCA has been determined to be the bounding accident scenario for peak containment pressure.In response to a LOCA the ECCS operates in two phases.The initial phase, known as the injection phase, starts at the receipt of a safety injection signal resulting in automatic start of the ECCS pumps.The pumps transfer the borated water contained in the Refueling Water Storage Tank (RWST)to the RCS to provide makeup for lost coolant and core cooling/reactivity control.As the RWST is depleted, the ECCS pump suctions are re-aligned to the containment recirculation sump to commence the recirculation phase, which provides long term reactor core and containment cooling.The ECCS consists of 6 ECCS pumps-2 high head Centrifugal Charging pumps (EIIS:BQ), 2 medium head Safety Injection pumps (EIIS:BQ), and 2 low head RHR pumps (EIIS:BP)-plus heat exchangers, accumulator tanks, and the associated piping valves and instrumentation. | | Ms. M.B. Depuydt, Compliance Engineer 616/465-5901, x1 589 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13) |
| The RHR pumps start on a safety injection signal and inject borated water from the RWST at a high rate of flow into the RCS when the RCS pressure drops below the shutoff head of the RHR pumps, as in the case of a large break LOCA (LB LOCA).NRC FORM 366A (6-1998) | | REPORTABLE REPORTABLE TO |
| NRC FORM 366A U.S.NUCLEAR REGULATORY COMMISSION (6-1 998)LICENSEE EVENT REPORT (LER)TEXT CONTINUATION FACILITY NAME (1)Cook Nuclear Plant Unit 1 DOCKET NUMBER(2)05000-315 YEAR 1998 LER NUMBER (6)SEQUENTIAL NUMBER 014 REVISION NUMBER 03 PAGE (3)3 of 3 TEXT (if more space is required, use addilfonal copies of NRC Form (366A)(17)The injection phase of the ECCS operation is terminated after the level of the RWST level drops to a pre-determined point.The suction for the RHR pumps and Containment Spray pumps (CTS)is then transferred to the containment recirculation sump.If required, a portion of the RHR flow can be diverted to the upper containment RHR spray headers during the recirculation phase to supplement the containment cooling operation of the CTS.This can be initiated should the containment pressure rise after the initial pressure suppression following a LOCA.Under these conditions, if core temperature is satisfactory, the operator may divert one or both trains of RHR from injection to RHR sprays, thereby supplementing CTS spray flow with an additional 1890 gallons per minute per train.Evaluation of the identified delay in commencing RHR spray has been performed, considering not only this particular condition, but other conditions which could have an effect on peak containment pressure.The results of this evaluation, using the licensing basis LOTIC code, indicated the peak containment pressure to be 13.85 psig, which is above the current design basis of 12 psig but below its ultimate capability of 36 psig.While 13.85 psig is above the licensing and technical specification basis of 12 psig, it is less than the 16.1 psig that the units were subjected to in their pre-operational structural integrity testing.Therefore, it was concluded that the containment would have remained functional even if it was potentially subjected to pressures as high as 13.85 psig.Corrective Actions The containment integrity analysis will be used to determine the appropriate point to initiate RHR spray to ensure that the 12 psig containment design pressure, following a postulated accident, is not exceeded.This task will be completed by August 31, 1999.The Function Restoration Procedure FRZ-1,"Response to High-High Containment Pressure" will be revised to be consistent with the new analysis and will allow time for initiation of RHR spray, repositioning of valves, and filling of RHR spray lines.As Unit 2 will be returned to service first, the Unit 2 procedure will be revised and approved by December 1, 1999, with the procedure for Unit 1 scheduled for revision and approval by January 30, 2000.To alleviate the interface problem with Westinghouse, a program is being developed and implemented to identify, document and control the key accident analyses assumptions used in the safety analyses, including those that can be impacted by operator action in the EOPs, the key events involving operator action duration that can impact the safety analyses and are part of the EOPs, and the setpoints that will be subject to engineering control that are part of the EOPs.Implementation of the program will be complete by August 31, 1999.Previous Similar Events None NRC FORM 366A (6-1998)}} | | " CAUSE SYSTEM COMPONENT MANUFACTURER TO EPIX CAUSE MANUFACTURER EPIX SUPPLEMENTAL REPORT EXPECTED 14 EXPECTED YEAR YES X SUBMISSION IfYes, corn lets EXPECTED SUBMISSION DATE NO DATE 15 Abstract (Umit to 1400 spaces, l.e., approximately 15 single-spaced typewritten lines) (16) |
| | On March 10, 1998, with Units 1 and 2 in Mode 5, it was determined that both units had operated in an unanalyzed condition due to Functional Restoration Procedure FRZ-1, "Response to High-High Containment Pressure", not being consistent with the containment integrity analysis of record. Had the procedure been implemented, the potential existed for post-accident containment pressure to exceed its design basis limit of 12 psig. In accordance with 10CFR50.72(b)(2)(i), an ENS notification was made. This LER is therefore submitted in accordance with 10CFR50.73(a)(2)(ii)(A), for an unanalyzed condition, and 10CFR50.73(a)(2)(ii)(B), for a condition outside the design basis. |
| | The root cause of this condition was inadequate interface with Westinghouse regarding the assumptions used in the safety analysis. The procedure will be revised to direct initiation of RHR spray at the appropriate point to ensure that containment design pressure is not exceeded. A program will be established to identify, document and control key accident analyses assumptions, including those impacting the Emergency Operating Procedures. Additional actions will be taken to strengthen the communications between Operations and Engineering - Safety Analysis, which maintains oversight of vendors performing safety analyses that might impact actions taken by the operators. |
| | Evaluation of this condition has been performed. It has been concluded that containment pressure would have exceeded the design pressure of 12 psig, reaching a calculated peak of 13.85 psig. This value remained below the pre-operational structural integrity test value of 16.1 psig, therefore, it was concluded that the containment would have remained functional. |
| | '3)'3)07280073) 990722 PDR ADOCK 05000315 PDR |
| | |
| | NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (6-1 998) |
| | LICENSEE EVENT REPORT (LER) |
| | TEXT CONTINUATION FACIUTYNAME (1) DOCKET NUMBER(2) LER NUMBER (6) PAGE (3) |
| | YEAR SEQUENTIAL REVISION Cook Nuclear Plant Unit 1 05000-315 NUMBER NUMBER 2 of 3 1998 014 03 TEXT (ifmore space is required, use additional copies of NRC Form (366A) (17) |
| | Conditions Prior to Event Unit 1 was in Mode 5, Cold Shutdown Unit 2 was in Mode 5, Cold Shutdown Descri tion of Event On March 10, 1998, while performing a Containment Spray self assessment, it was determined that the actions directed by Functional Restoration Procedure 1,2-4023.OHP.FRZ-1, "Response to High-High Containment Pressure", were not consistent with the assumptions in the containment integrity analysis of record. |
| | Residual Heat Removal (RHR) (EIIS:BP) spray is designed to supplement the pressure mitigation function of containment spray during either a Loss of Coolant Accident (LOCA) or Main Steam Line Break (MSLB). In accordance with containment integrity analysis input assumptions, FRZ-1 directs that RHR spray be manually, initiated when containment pressure reaches 8 psig. The safety analysis did not make allowance for the time delay between containment pressure reaching 8 psig and the delivery of RHR spray to containment. This time delay results from the summation of the time required for the operator to recognize that containment pressure has reached 8 psig; for the RHR spray valves to open and RHR to Reactor Coolant system isolation valves to close/throttle; and for RHR flow to fill the spray line and spray headers. |
| | Had the use of this procedure in its current form been required, containment peak pressure mitigation would have been affected. |
| | Cause of Event This condition was the result of an inadequate interface with Westinghouse regarding the assumptions used in the safety analysis and how they were implemented at the plant. Equipment response times and operator action times were not included by Westinghouse when assumptions regarding RHR spray were incorporated into the analysis. |
| | Anal sis of Event This condition was determined to be reportable in accordance with 10CFR50.73(a)(2)(ii)(A), for an unanalyzed condition that significantly compromises plant safety, and 10CFR50.73(a)(2)(ii)(B) for a condition outside the design basis. |
| | The Emergency Core Cooling System (ECCS) is one of the Engineered Safety Feature systems, which mitigate the consequences of a major breach of the Reactor Coolant system (RCS), or main steam lines inside containment. The RCS line break results in a LOCA, during which the ECCS provides a significant volume of makeup to the RCS as well as core cooling and reactivity control. The LOCA has been determined to be the bounding accident scenario for peak containment pressure. |
| | In response to a LOCA the ECCS operates in two phases. The initial phase, known as the injection phase, starts at the receipt of a safety injection signal resulting in automatic start of the ECCS pumps. The pumps transfer the borated water contained in the Refueling Water Storage Tank (RWST) to the RCS to provide makeup for lost coolant and core cooling/reactivity control. As the RWST is depleted, the ECCS pump suctions are re-aligned to the containment recirculation sump to commence the recirculation phase, which provides long term reactor core and containment cooling. |
| | The ECCS consists of 6 ECCS pumps -2 high head Centrifugal Charging pumps (EIIS:BQ), 2 medium head Safety Injection pumps (EIIS:BQ), and 2 low head RHR pumps (EIIS:BP) - plus heat exchangers, accumulator tanks, and the associated piping valves and instrumentation. |
| | The RHR pumps start on a safety injection signal and inject borated water from the RWST at a high rate of flow into the RCS when the RCS pressure drops below the shutoff head of the RHR pumps, as in the case of a large break LOCA (LBLOCA). |
| | NRC FORM 366A (6-1998) |
| | |
| | NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (6-1 998) |
| | LICENSEE EVENT REPORT (LER) |
| | TEXT CONTINUATION FACILITYNAME (1) DOCKET NUMBER(2) LER NUMBER (6) PAGE (3) |
| | YEAR SEQUENTIAL REVISION Cook Nuclear Plant Unit 1 05000-315 NUMBER NUMBER 3 of 3 1998 014 03 TEXT (ifmore space is required, use addilfonal copies of NRC Form (366A) (17) |
| | The injection phase of the ECCS operation is terminated after the level of the RWST level drops to a pre-determined point. |
| | The suction for the RHR pumps and Containment Spray pumps (CTS) is then transferred to the containment recirculation sump. If required, a portion of the RHR flow can be diverted to the upper containment RHR spray headers during the recirculation phase to supplement the containment cooling operation of the CTS. This can be initiated should the containment pressure rise after the initial pressure suppression following a LOCA. Under these conditions, if core temperature is satisfactory, the operator may divert one or both trains of RHR from injection to RHR sprays, thereby supplementing CTS spray flow with an additional 1890 gallons per minute per train. |
| | Evaluation of the identified delay in commencing RHR spray has been performed, considering not only this particular condition, but other conditions which could have an effect on peak containment pressure. The results of this evaluation, using the licensing basis LOTIC code, indicated the peak containment pressure to be 13.85 psig, which is above the current design basis of 12 psig but below its ultimate capability of 36 psig. While 13.85 psig is above the licensing and technical specification basis of 12 psig, it is less than the 16.1 psig that the units were subjected to in their pre-operational structural integrity testing. Therefore, it was concluded that the containment would have remained functional even if it was potentially subjected to pressures as high as 13.85 psig. |
| | Corrective Actions The containment integrity analysis will be used to determine the appropriate point to initiate RHR spray to ensure that the 12 psig containment design pressure, following a postulated accident, is not exceeded. This task will be completed by August 31, 1999. The Function Restoration Procedure FRZ-1, "Response to High-High Containment Pressure" will be revised to be consistent with the new analysis and will allow time for initiation of RHR spray, repositioning of valves, and filling of RHR spray lines. As Unit 2 will be returned to service first, the Unit 2 procedure will be revised and approved by December 1, 1999, with the procedure for Unit 1 scheduled for revision and approval by January 30, 2000. |
| | To alleviate the interface problem with Westinghouse, a program is being developed and implemented to identify, document and control the key accident analyses assumptions used in the safety analyses, including those that can be impacted by operator action in the EOPs, the key events involving operator action duration that can impact the safety analyses and are part of the EOPs, and the setpoints that will be subject to engineering control that are part of the EOPs. |
| | Implementation of the program will be complete by August 31, 1999. |
| | Previous Similar Events None NRC FORM 366A (6-1998)}} |
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| ML17326A065 |
| Person / Time |
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| Site: |
Cook  |
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| Issue date: |
07/22/1999 |
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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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| ML17326A064 |
List: |
|---|
| References |
|---|
| LER-98-014, NUDOCS 9907280079 |
| Download: ML17326A065 (3) |
|
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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
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NRC Form 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED 8Y OM8 NO. 3160%104 EXPIRES 06/30/2001 (6-1 996)
ESTBIATED BURDEN PER RESPONSE TO COMPLY WITH THIS MANDA'tORY INFORMATION COLLECTION REQUEST: 500 HRS. REPORTED LESSONS LEARNED ARE INCORPORATED INTO THE UCENSUIQ PROCESS AND FED BACK TO INDUStRY.
LICENSEE EVENT REPORT (LER) FORWARD COLTJENTS REQARINNQ BURDEN ESTUJATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH rt+ F33), U.S. NUCLEAR REGULATORY COLBBSSION, WASHINGTON. DC 205554001. AND TO THE PAPERWORK REDUCTION PROJECT I3150oterx OFFICE OF MANAGEMENT AND BUDGEt. WASHINGTON. DC (See reverse for required number of 20503 digits/characters for each block)
FACIUTY NAME II) DOCKET NUMBER I2) PAGE I3)
Cook Nuclear Plant Unit 1 05000-315 1 of3 TITLE I4)
"Response to High-High Containment Pressure" Procedure Not Consistent with Analysis of Record EVENT DATE (6) LER NUMBER (6) REPORT DATE (7) OTHER FACILITIES INVOLVED (6)
FACIUTY NAM DOCKET NUMBER SEQUENTIAL REVISION DC Cook- Unit 2 05000-316 MONTH DAY YEAR YEAR NUMBER NUMBER MONTH DAY ACIU NAM DOCKET NUMB 03 10 98 1998 014 03 07 22 1999 OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR li: (Check one or more) (19)
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)(li) 50.73(a)(2)(x)
LEVEL (10) 00 20.2203(a)(2)(l) 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)(lv) OTHER 20.2203(a)(2)(iii) 50.36(c)(1) 50.73(a)(2)(v)
Specfy in Abstract below 20.2203(a)(2)(iv) 50.36(c)(2) 50.73(a)(2)(vii) or n NRC Form 388A LICENSEE CONTACT FOR THIS LER (12)
TELEPHONE NUMBER trnc/ude Area Code)
Ms. M.B. Depuydt, Compliance Engineer 616/465-5901, x1 589 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)
REPORTABLE REPORTABLE TO
" CAUSE SYSTEM COMPONENT MANUFACTURER TO EPIX CAUSE MANUFACTURER EPIX SUPPLEMENTAL REPORT EXPECTED 14 EXPECTED YEAR YES X SUBMISSION IfYes, corn lets EXPECTED SUBMISSION DATE NO DATE 15 Abstract (Umit to 1400 spaces, l.e., approximately 15 single-spaced typewritten lines) (16)
On March 10, 1998, with Units 1 and 2 in Mode 5, it was determined that both units had operated in an unanalyzed condition due to Functional Restoration Procedure FRZ-1, "Response to High-High Containment Pressure", not being consistent with the containment integrity analysis of record. Had the procedure been implemented, the potential existed for post-accident containment pressure to exceed its design basis limit of 12 psig. In accordance with 10CFR50.72(b)(2)(i), an ENS notification was made. This LER is therefore submitted in accordance with 10CFR50.73(a)(2)(ii)(A), for an unanalyzed condition, and 10CFR50.73(a)(2)(ii)(B), for a condition outside the design basis.
The root cause of this condition was inadequate interface with Westinghouse regarding the assumptions used in the safety analysis. The procedure will be revised to direct initiation of RHR spray at the appropriate point to ensure that containment design pressure is not exceeded. A program will be established to identify, document and control key accident analyses assumptions, including those impacting the Emergency Operating Procedures. Additional actions will be taken to strengthen the communications between Operations and Engineering - Safety Analysis, which maintains oversight of vendors performing safety analyses that might impact actions taken by the operators.
Evaluation of this condition has been performed. It has been concluded that containment pressure would have exceeded the design pressure of 12 psig, reaching a calculated peak of 13.85 psig. This value remained below the pre-operational structural integrity test value of 16.1 psig, therefore, it was concluded that the containment would have remained functional.
'3)'3)07280073) 990722 PDR ADOCK 05000315 PDR
NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (6-1 998)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACIUTYNAME (1) DOCKET NUMBER(2) LER NUMBER (6) PAGE (3)
YEAR SEQUENTIAL REVISION Cook Nuclear Plant Unit 1 05000-315 NUMBER NUMBER 2 of 3 1998 014 03 TEXT (ifmore space is required, use additional copies of NRC Form (366A) (17)
Conditions Prior to Event Unit 1 was in Mode 5, Cold Shutdown Unit 2 was in Mode 5, Cold Shutdown Descri tion of Event On March 10, 1998, while performing a Containment Spray self assessment, it was determined that the actions directed by Functional Restoration Procedure 1,2-4023.OHP.FRZ-1, "Response to High-High Containment Pressure", were not consistent with the assumptions in the containment integrity analysis of record.
Residual Heat Removal (RHR) (EIIS:BP) spray is designed to supplement the pressure mitigation function of containment spray during either a Loss of Coolant Accident (LOCA) or Main Steam Line Break (MSLB). In accordance with containment integrity analysis input assumptions, FRZ-1 directs that RHR spray be manually, initiated when containment pressure reaches 8 psig. The safety analysis did not make allowance for the time delay between containment pressure reaching 8 psig and the delivery of RHR spray to containment. This time delay results from the summation of the time required for the operator to recognize that containment pressure has reached 8 psig; for the RHR spray valves to open and RHR to Reactor Coolant system isolation valves to close/throttle; and for RHR flow to fill the spray line and spray headers.
Had the use of this procedure in its current form been required, containment peak pressure mitigation would have been affected.
Cause of Event This condition was the result of an inadequate interface with Westinghouse regarding the assumptions used in the safety analysis and how they were implemented at the plant. Equipment response times and operator action times were not included by Westinghouse when assumptions regarding RHR spray were incorporated into the analysis.
Anal sis of Event This condition was determined to be reportable in accordance with 10CFR50.73(a)(2)(ii)(A), for an unanalyzed condition that significantly compromises plant safety, and 10CFR50.73(a)(2)(ii)(B) for a condition outside the design basis.
The Emergency Core Cooling System (ECCS) is one of the Engineered Safety Feature systems, which mitigate the consequences of a major breach of the Reactor Coolant system (RCS), or main steam lines inside containment. The RCS line break results in a LOCA, during which the ECCS provides a significant volume of makeup to the RCS as well as core cooling and reactivity control. The LOCA has been determined to be the bounding accident scenario for peak containment pressure.
In response to a LOCA the ECCS operates in two phases. The initial phase, known as the injection phase, starts at the receipt of a safety injection signal resulting in automatic start of the ECCS pumps. The pumps transfer the borated water contained in the Refueling Water Storage Tank (RWST) to the RCS to provide makeup for lost coolant and core cooling/reactivity control. As the RWST is depleted, the ECCS pump suctions are re-aligned to the containment recirculation sump to commence the recirculation phase, which provides long term reactor core and containment cooling.
The ECCS consists of 6 ECCS pumps -2 high head Centrifugal Charging pumps (EIIS:BQ), 2 medium head Safety Injection pumps (EIIS:BQ), and 2 low head RHR pumps (EIIS:BP) - plus heat exchangers, accumulator tanks, and the associated piping valves and instrumentation.
The RHR pumps start on a safety injection signal and inject borated water from the RWST at a high rate of flow into the RCS when the RCS pressure drops below the shutoff head of the RHR pumps, as in the case of a large break LOCA (LBLOCA).
NRC FORM 366A (6-1998)
NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (6-1 998)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITYNAME (1) DOCKET NUMBER(2) LER NUMBER (6) PAGE (3)
YEAR SEQUENTIAL REVISION Cook Nuclear Plant Unit 1 05000-315 NUMBER NUMBER 3 of 3 1998 014 03 TEXT (ifmore space is required, use addilfonal copies of NRC Form (366A) (17)
The injection phase of the ECCS operation is terminated after the level of the RWST level drops to a pre-determined point.
The suction for the RHR pumps and Containment Spray pumps (CTS) is then transferred to the containment recirculation sump. If required, a portion of the RHR flow can be diverted to the upper containment RHR spray headers during the recirculation phase to supplement the containment cooling operation of the CTS. This can be initiated should the containment pressure rise after the initial pressure suppression following a LOCA. Under these conditions, if core temperature is satisfactory, the operator may divert one or both trains of RHR from injection to RHR sprays, thereby supplementing CTS spray flow with an additional 1890 gallons per minute per train.
Evaluation of the identified delay in commencing RHR spray has been performed, considering not only this particular condition, but other conditions which could have an effect on peak containment pressure. The results of this evaluation, using the licensing basis LOTIC code, indicated the peak containment pressure to be 13.85 psig, which is above the current design basis of 12 psig but below its ultimate capability of 36 psig. While 13.85 psig is above the licensing and technical specification basis of 12 psig, it is less than the 16.1 psig that the units were subjected to in their pre-operational structural integrity testing. Therefore, it was concluded that the containment would have remained functional even if it was potentially subjected to pressures as high as 13.85 psig.
Corrective Actions The containment integrity analysis will be used to determine the appropriate point to initiate RHR spray to ensure that the 12 psig containment design pressure, following a postulated accident, is not exceeded. This task will be completed by August 31, 1999. The Function Restoration Procedure FRZ-1, "Response to High-High Containment Pressure" will be revised to be consistent with the new analysis and will allow time for initiation of RHR spray, repositioning of valves, and filling of RHR spray lines. As Unit 2 will be returned to service first, the Unit 2 procedure will be revised and approved by December 1, 1999, with the procedure for Unit 1 scheduled for revision and approval by January 30, 2000.
To alleviate the interface problem with Westinghouse, a program is being developed and implemented to identify, document and control the key accident analyses assumptions used in the safety analyses, including those that can be impacted by operator action in the EOPs, the key events involving operator action duration that can impact the safety analyses and are part of the EOPs, and the setpoints that will be subject to engineering control that are part of the EOPs.
Implementation of the program will be complete by August 31, 1999.
Previous Similar Events None NRC FORM 366A (6-1998)