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| {{#Wiki_filter:K) K) July 26, 1996NRC INFORMATION NOTICE 96-41: EFFECTS OF A DECREASE IN FEEDWATER TEMPERATUREON NUCLEAR INSTRUMENTATION | | {{#Wiki_filter:K) K)UNITED STATESNUCLEAR REGULATORY COMMISSIONOFFICE OF NUCLEAR REACTOR REGULATIONWASHINGTON, D.C. 20555-0001July 26, 1996NRC INFORMATION NOTICE 96-41: EFFECTS OF A DECREASE IN FEEDWATER TEMPERATUREON NUCLEAR INSTRUMENTATION |
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| ==Description of Circumstances== | | ==Description of Circumstances== |
| On February 14, 1996, the licensee for the Comanche Peak Steam ElectricStation was operating Unit 2 at 95 percent rated thermal power near end-of-core life when a significant reduction in feedwater temperature occurredbecause of the loss of feedwater heaters. This reduction, in turn, caused areduction in the reactor coolant system cold-leg temperatures. The colderreactor coolant temperature, with a large negative moderator temperaturecoefficient, caused reactor power to increase to approximately 102 percentaccording to ex-core nuclear instrumentation. The nitrogen-16 (N-16)detection system reached the overpower turbine runback setpoint (109 percent)and initiated a turbine runback. The N-16 detection system measures N-16activity in the primary coolant as a measure of the total power generation.This system is a substitute for the resistance temperature detector over-temperature and over-power reactor trip functions used at other WestinghousePWRs. The plant stabil zed at an indicated power of approximately 97 percentaccording to the ex-core nuclear instrumentation.After approximately 90 minutes, a second similar turbine runback occurredwhile restoring balance-of-plant equipment. Following this runback, reactorpower was stabilized at approximately 100 percent according to nuclearinstrumentation. During the next 30 minutes, the reactor was operated atapproximately 100 percent power as indicated by nuclear instrumentation, withreactor coolant temperatures below normal. The licensee noted that the N-169607220l60ujo i 7 9,oi4(R ~IE ctG IN 96-41July 26, 1996 detection system indicated approximately 106 percent power and the computer-based plant calorimetric system indicated approximately 102 percent power.Subsequently, the reactor power was reduced to less than 100 percent by allindications.DiscussionThere are three aspects of this event which have generic implications. First,with a loss of secondary plant efficiency, programmed T e can no longerreliably represent core thermal power. Second, the venturi-based input intothe computer-based calorimetric system may not be accurate with coldfeedwater. And third, the final safety analysis report had not analyzed thistransient accurately.Following the second runback, operators noted that reactor power indicated<100 percent according to nuclear instrumentation. Although the operatorsknew that cold feedwater could cause an increase in the amount of neutronattenuation, they believed that the nuclear instrumentation indicatedconservatively (i.e., higher than actual) because they were maintaining TA"eapproximately 1.7 eC [3 OF] above TRef. The licensee could not use thecomputer-based calorimetric until some time after the second turbine runbackdue to maintenance activities. Te , based on the main turbine impulsepressure, is programmed as a functlon of turbine load and, for normalefficiency, is a good representation of thermal power. When the unit lost thefeedwater heaters, the plant efficiency decreased. Because the main turbineelectro-hydraulic control system maintained generator output, core thermalpower increased to account for the loss of efficiency, and thus, TRef nolonger accurately represented the core thermal power.The cold-leg temperature is a more appropriate indicator of the accuracy ofthe nuclear instrumentation than programmed TY.e. As the cold-leg temperaturedecreased, the amount of neutron attenuation in the downcomer area surroundingthe core increased and hence affected the amount of neutrons reaching thedetectors. The licensee analysis showed that for every 0.6 C (1 OF] of cold-leg temperature change, the nuclear instrumentation was affected by 0.6 to 0.8percent power. A review of the second transient showed that the cold-legtemperature was approximately 2.5 °C [4.5 OF] lower than when the detectorswere last calibrated. This corresponded to a 3 to 4 percent error, whichcorresponded to the difference in the actual versus the indicated power (104percent actual versus 100 percent indicated).During the review, the licensee noted that the computer-based calorimetric was4 percent lower than the actual thermal power (N-16 power monitor). Thecalorimetric was based on feedwater flow measured by venturis. Although thecalorimetric calculation used feedwater temperature as an input, temperaturessignificantly different than the normal 227 OC [440 OF] introduced errors intothe calculation.Finally, the actual events involved temperature and power levels that exceededthose in the analysis of the Decrease in Feedwater Temperature" eventpresented in Chapter 15 of the licensee final safety analysis report. In that IN 96-41July 26, 1996 analysis, the inadvertent opening of the low-pressure heater bypass valve,coupled with the trip of the heater drain pumps, resulted in a feedwatertemperature drop of less than 19 'C (35 OF], and a corresponding powerincrease of less than 10 percent. In the actual event, the feedwatertemperature dropped by approximately 111 °C (200 OF], and the licenseecalculated that reactor power would have increased by approximately 35 percentwithout operator or protective actions. The licensee determined that althoughthe initiating events were the same, the Chapter 15 analysis did not accountfor the loss of extraction steam to the high-pressure heaters, which was thecause of the temperature difference. During the event, a level imbalanceoccurred between the two heater drain tanks, which resulted in the isolationof extraction steam.The NRC staff review of analyses of feedwater temperature events at similarfacilities revealed that most of these analyses assumed similar initiatingevents as the Comanche Peak analysis and had similar conclusions concerningthe amount of feedwater temperature drop. The licensee has reanalyzed theevent to include a 119 OC [246 OF] feedwater temperature drop and concludedthat all accident analysis parameters remained within requirements.This information notice requires no specific action or written response. Ifyou have any questions about the information in this notice, please contactone of the technical contacts listed below or the appropriate Office ofNuclear Reactor Regulation project manager.Brian K. Grimes, Acting DirectorDivision of Reactor Program ManagementOffice of Nuclear Reactor RegulationTechnical contacts: Harry A. Freeman, RIV(817) 897-1500E-mail: haf~nrc.govChu-Yu Liang, NRR(301) 415-2878E-mail: cylenrc.gov | | On February 14, 1996, the licensee for the Comanche Peak Steam ElectricStation was operating Unit 2 at 95 percent rated thermal power near end-of-core life when a significant reduction in feedwater temperature occurredbecause of the loss of feedwater heaters. This reduction, in turn, caused areduction in the reactor coolant system cold-leg temperatures. The colderreactor coolant temperature, with a large negative moderator temperaturecoefficient, caused reactor power to increase to approximately 102 percentaccording to ex-core nuclear instrumentation. The nitrogen-16 (N-16)detection system reached the overpower turbine runback setpoint (109 percent)and initiated a turbine runback. The N-16 detection system measures N-16activity in the primary coolant as a measure of the total power generation.This system is a substitute for the resistance temperature detector over-temperature and over-power reactor trip functions used at other WestinghousePWRs. The plant stabil zed at an indicated power of approximately 97 percentaccording to the ex-core nuclear instrumentation.After approximately 90 minutes, a second similar turbine runback occurredwhile restoring balance-of-plant equipment. Following this runback, reactorpower was stabilized at approximately 100 percent according to nuclearinstrumentation. During the next 30 minutes, the reactor was operated atapproximately 100 percent power as indicated by nuclear instrumentation, withreactor coolant temperatures below normal. The licensee noted that the N-169607220l60ujo i 7 9,oi4(R ~IE ctG |
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| | IN 96-41July 26, 1996 detection system indicated approximately 106 percent power and the computer-based plant calorimetric system indicated approximately 102 percent power.Subsequently, the reactor power was reduced to less than 100 percent by allindications.DiscussionThere are three aspects of this event which have generic implications. First,with a loss of secondary plant efficiency, programmed T e can no longerreliably represent core thermal power. Second, the venturi-based input intothe computer-based calorimetric system may not be accurate with coldfeedwater. And third, the final safety analysis report had not analyzed thistransient accurately.Following the second runback, operators noted that reactor power indicated<100 percent according to nuclear instrumentation. Although the operatorsknew that cold feedwater could cause an increase in the amount of neutronattenuation, they believed that the nuclear instrumentation indicatedconservatively (i.e., higher than actual) because they were maintaining TA"eapproximately 1.7 eC [3 OF] above TRef. The licensee could not use thecomputer-based calorimetric until some time after the second turbine runbackdue to maintenance activities. Te , based on the main turbine impulsepressure, is programmed as a functlon of turbine load and, for normalefficiency, is a good representation of thermal power. When the unit lost thefeedwater heaters, the plant efficiency decreased. Because the main turbineelectro-hydraulic control system maintained generator output, core thermalpower increased to account for the loss of efficiency, and thus, TRef nolonger accurately represented the core thermal power.The cold-leg temperature is a more appropriate indicator of the accuracy ofthe nuclear instrumentation than programmed TY.e. As the cold-leg temperaturedecreased, the amount of neutron attenuation in the downcomer area surroundingthe core increased and hence affected the amount of neutrons reaching thedetectors. The licensee analysis showed that for every 0.6 C (1 OF] of cold-leg temperature change, the nuclear instrumentation was affected by 0.6 to 0.8percent power. A review of the second transient showed that the cold-legtemperature was approximately 2.5 °C [4.5 OF] lower than when the detectorswere last calibrated. This corresponded to a 3 to 4 percent error, whichcorresponded to the difference in the actual versus the indicated power (104percent actual versus 100 percent indicated).During the review, the licensee noted that the computer-based calorimetric was4 percent lower than the actual thermal power (N-16 power monitor). Thecalorimetric was based on feedwater flow measured by venturis. Although thecalorimetric calculation used feedwater temperature as an input, temperaturessignificantly different than the normal 227 OC [440 OF] introduced errors intothe calculation.Finally, the actual events involved temperature and power levels that exceededthose in the analysis of the Decrease in Feedwater Temperature" eventpresented in Chapter 15 of the licensee final safety analysis report. In that |
| List Of Recently Issued HRC Information NoticesA1h4 Stir A Je6tQ K> KJAttachmentIN 96-41July 26, 1996 LIST OF RECENTLY ISSUEDNRC INFORMATION NOTICESInformation Date ofNotice No. Subject Issuance Issued to96-4096-09,Supp. 196-3996-38Deficiencies in MaterialDedication and Procure-ment Practices and inAudits of VendorsDamage in Foreign SteamGenerator InternalsEstimates of Decay HeatUsing ANS 5.1 Decay HeatStandard May Vary Signi-ficantlyResults of Steam GeneratorTube ExaminationsInaccurate Reactor WaterLevel Indication and Inad-vertent Draindown DuringShutdownDegradation of CoolingWater Systems Due to IcingFailure of Safety Systemson Self-Shielded Irradia-tors Because of InadequateMaintenance and TrainingHydrogen Gas Ignitionduring Closure Weldingof a VSC-24 Multi-AssemblySealed Basket07/25/9607/10/9607/05/9606/21/9606/18/9606/12/9606/11/9605/31/96All holders of OLs or CPsfor nuclear power reactorsAll holders of OLs or CPsfor pressurized-waterreactorsAll holders of OLs or CPsfor nuclear power reactorsAll holders of OLs or CPsfor pressurized waterreactorsAll pressurized waterreactor facilities holdingan operating license or aconstruction permitAll holders of OLs or CPsfor nuclear power reactorsAll U.S. Nuclear RegulatoryCommission irradiatorlicensees and vendorsAll holders of OLs or CPsfor nuclear power reactors96-3796-3696-3596-34OL -Operating LicenseCP -Construction Permit
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| *~ -K> KIN 96-41July 26, 1996 analysis, the inadvertent opening of the low-pressure heater bypass valve,coupled with the trip of the heater drain pumps, resulted in a feedwatertemperature drop of less than 19 *C [35 OF], and a corresponding powerincrease of less than 10 percent. In the actual event, the feedwatertemperature dropped by approximately 111 *C [200 OF], and the licenseecalculated that reactor power would have increased by approximately 35 percentwithout operator or protective actions. The licensee determined that althoughthe initiating events were the same, the Chapter 15 analysis did not accountfor the loss of extraction steam to the high-pressure heaters, which was thecause of the temperature difference. During the event, a level imbalanceoccurred between the two heater drain tanks, which resulted in the isolationof extraction steam.The NRC staff review of analyses of feedwater temperature events at similarfacilities revealed that most of these analyses assumed similar initiatingevents as the Comanche Peak analysis and had similar conclusions concerningthe amount of feedwater temperature drop. The licensee has reanalyzed theevent to include a 119 *C [246 OF] feedwater temperature drop and concludedthat all accident analysis parameters remained within requirements.This information notice requires no specific action or written response. Ifyou have any questions about the information in this notice,-please contactone of the technical contacts listed below or the appropriate Office ofNuclear Reactor Regulation project manager.Original signed by Brian K. GrimesBrian K. Grimes, Acting DirectorDivision of Reactor Program ManagementOffice of Nuclear Reactor RegulationTechnical contacts: Harry A. Freeman, RIV(817) 897-1500E-mail: haf@nrc.govChu-Yu Liang, NRR(301) 415-2878E-mail: cyl~nrc.gov
| | IN 96-41July 26, 1996 analysis, the inadvertent opening of the low-pressure heater bypass valve,coupled with the trip of the heater drain pumps, resulted in a feedwatertemperature drop of less than 19 'C (35 OF], and a corresponding powerincrease of less than 10 percent. In the actual event, the feedwatertemperature dropped by approximately 111 °C (200 OF], and the licenseecalculated that reactor power would have increased by approximately 35 percentwithout operator or protective actions. The licensee determined that althoughthe initiating events were the same, the Chapter 15 analysis did not accountfor the loss of extraction steam to the high-pressure heaters, which was thecause of the temperature difference. During the event, a level imbalanceoccurred between the two heater drain tanks, which resulted in the isolationof extraction steam.The NRC staff review of analyses of feedwater temperature events at similarfacilities revealed that most of these analyses assumed similar initiatingevents as the Comanche Peak analysis and had similar conclusions concerningthe amount of feedwater temperature drop. The licensee has reanalyzed theevent to include a 119 OC [246 OF] feedwater temperature drop and concludedthat all accident analysis parameters remained within requirements.This information notice requires no specific action or written response. Ifyou have any questions about the information in this notice, please contactone of the technical contacts listed below or the appropriate Office ofNuclear Reactor Regulation project manager.Brian K. Grimes, Acting DirectorDivision of Reactor Program ManagementOffice of Nuclear Reactor RegulationTechnical contacts: Harry A. Freeman, RIV(817) 897-1500E-mail: haf~nrc.govChu-Yu Liang, NRR(301) 415-2878E-mail: cylenrc.govAttachment: List Of Recently Issued HRC Information NoticesA1h4 Stir A Je6tQ |
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| | K> KJAttachmentIN 96-41July 26, 1996 LIST OF RECENTLY ISSUEDNRC INFORMATION NOTICESInformation Date ofNotice No. Subject Issuance Issued to96-4096-09,Supp. 196-3996-38Deficiencies in MaterialDedication and Procure-ment Practices and inAudits of VendorsDamage in Foreign SteamGenerator InternalsEstimates of Decay HeatUsing ANS 5.1 Decay HeatStandard May Vary Signi-ficantlyResults of Steam GeneratorTube ExaminationsInaccurate Reactor WaterLevel Indication and Inad-vertent Draindown DuringShutdownDegradation of CoolingWater Systems Due to IcingFailure of Safety Systemson Self-Shielded Irradia-tors Because of InadequateMaintenance and TrainingHydrogen Gas Ignitionduring Closure Weldingof a VSC-24 Multi-AssemblySealed Basket07/25/9607/10/9607/05/9606/21/9606/18/9606/12/9606/11/9605/31/96All holders of OLs or CPsfor nuclear power reactorsAll holders of OLs or CPsfor pressurized-waterreactorsAll holders of OLs or CPsfor nuclear power reactorsAll holders of OLs or CPsfor pressurized waterreactorsAll pressurized waterreactor facilities holdingan operating license or aconstruction permitAll holders of OLs or CPsfor nuclear power reactorsAll U.S. Nuclear RegulatoryCommission irradiatorlicensees and vendorsAll holders of OLs or CPsfor nuclear power reactors96-3796-3696-3596-34OL -Operating LicenseCP -Construction Permit |
| List of Recently Issued NRC Information NoticesDOCUMENT NAME: G:\SSK2\INFONOT.C PTo receive a copy of this docunent, tndicate in the box CO~opy So attachment/enclosure EsCopy with attachment/enctosure N | | |
| | *~ -K> KIN 96-41July 26, 1996 analysis, the inadvertent opening of the low-pressure heater bypass valve,coupled with the trip of the heater drain pumps, resulted in a feedwatertemperature drop of less than 19 *C [35 OF], and a corresponding powerincrease of less than 10 percent. In the actual event, the feedwatertemperature dropped by approximately 111 *C [200 OF], and the licenseecalculated that reactor power would have increased by approximately 35 percentwithout operator or protective actions. The licensee determined that althoughthe initiating events were the same, the Chapter 15 analysis did not accountfor the loss of extraction steam to the high-pressure heaters, which was thecause of the temperature difference. During the event, a level imbalanceoccurred between the two heater drain tanks, which resulted in the isolationof extraction steam.The NRC staff review of analyses of feedwater temperature events at similarfacilities revealed that most of these analyses assumed similar initiatingevents as the Comanche Peak analysis and had similar conclusions concerningthe amount of feedwater temperature drop. The licensee has reanalyzed theevent to include a 119 *C [246 OF] feedwater temperature drop and concludedthat all accident analysis parameters remained within requirements.This information notice requires no specific action or written response. Ifyou have any questions about the information in this notice,-please contactone of the technical contacts listed below or the appropriate Office ofNuclear Reactor Regulation project manager.Original signed by Brian K. GrimesBrian K. Grimes, Acting DirectorDivision of Reactor Program ManagementOffice of Nuclear Reactor RegulationTechnical contacts: Harry A. Freeman, RIV(817) 897-1500E-mail: haf@nrc.govChu-Yu Liang, NRR(301) 415-2878E-mail: cyl~nrc.govAttachment: List of Recently Issued NRC Information NoticesDOCUMENT NAME: G:\SSK2\INFONOT.C PTo receive a copy of this docunent, tndicate in the box CO~opy So attachment/enclosure EsCopy with attachment/enctosure N |
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| * NocopsOFFICE C BC:SRXB I BC:LPECB lI (A) DW M iNAME CYLiang* RJones* AChaffee*HAFreeman* ____ _DATE 16/ 3/96 16/21/96 17/08/96 17LI/96 IOFFILIAL MLLUM LWUF* See previous concurrence Tech Editor reviewed & concurred on 05/28/96 | | * NocopsOFFICE C BC:SRXB I BC:LPECB lI (A) DW M iNAME CYLiang* RJones* AChaffee*HAFreeman* ____ _DATE 16/ 3/96 16/21/96 17/08/96 17LI/96 IOFFILIAL MLLUM LWUF* See previous concurrence Tech Editor reviewed & concurred on 05/28/96 |
| ~1~1 -,K)IN 96-XXJuly XX, 1996 for the loss of extraction steam to ticause of the temperature difference.occurred between the two heater drainof extraction steam.he high-pressure heaters, which was theDuring the event, a level imbalancetanks, which resulted in the isolationThe NRC staff review of analyses of feedwater temperature events at similarfacilities revealed that most of these analyses assumed similar initiatingevents as the Comanche Peak analysis and had similar conclusions concerningthe amount of feedwater temperature drop. The licensee has reanalyzed theevent to include a 119 'C [246 'F] feedwater temperature drop and concludedthat all accident analysis parameters remained within requirements.This information notice requires no specific action or written response. Ifyou have any questions about the information in this notice, please contactone of the technical contacts listed below or the appropriate Office ofNuclear Reactor Regulation project manager.Brian K. Grimes, Acting DirectorDivision of Reactor Program ManagementOffice of Nuclear Reactor RegulationTechnical contacts:Harry A. Freeman, RIV(817) 897-1500E-mail: haf~nrc.govChu-Yu Liang, NRR(301) 415-2878E-mail: cyl~nrc.gov | | ~1~1 -,K)IN 96-XXJuly XX, 1996 *See previous concurrence |
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| * NoOFFICE l kd BC: SRXB BC:PECB )D:DRNAME CYLiang* RJones* AChaffee* BGrimesHAFreeman*DATE 6/ 3/96 6/21/96 7/08/96 7/ /96OFFICIAL RECORD COPY*See previous concurrence IN 96-XXJuly XX, 1996 for the loss of extraction steam to the high-pressure heaters, which was thecause of the temperature difference. During the event, a level imbalanceoccurred between the two heater drain tanks, which resulted in the isolationof extraction steam.The NRC staff review of analyses of feedwater temperature events at similarfacilities revealed that most of these analyses assumed similar initiatingevents as the licensee analysis and had similar conclusions concerning theamount of feedwater temperature drop. The licensee has reanalyzed the eventpursuant to Section 50.59 of Title 10 of the Code of Federal Regulations toinclude a 119 'c [246 OF] feedwater temperature drop and concluded that allaccident analysis parameters remained within requirements.This information notice requires no specific action or written response. Ifyou have any questions about the information in this notice, please contactone of the technical contacts listed below or the appropriate Office ofNuclear Reactor Regulation project manager.Brian K. Grimes, Acting DirectorDivision of Reactor Program ManagementOffice of Nuclear Reactor RegulationTechnical contacts: Harry A. Freeman, RIV(817) 897-1500E-mail: haftnrc.govChu-Yu Liang, NRR(301) 415-2878E-mail: cyl~nrc.gov
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| | IN 96-XXJuly XX, 1996 for the loss of extraction steam to the high-pressure heaters, which was thecause of the temperature difference. During the event, a level imbalanceoccurred between the two heater drain tanks, which resulted in the isolationof extraction steam.The NRC staff review of analyses of feedwater temperature events at similarfacilities revealed that most of these analyses assumed similar initiatingevents as the licensee analysis and had similar conclusions concerning theamount of feedwater temperature drop. The licensee has reanalyzed the eventpursuant to Section 50.59 of Title 10 of the Code of Federal Regulations toinclude a 119 'c [246 OF] feedwater temperature drop and concluded that allaccident analysis parameters remained within requirements.This information notice requires no specific action or written response. Ifyou have any questions about the information in this notice, please contactone of the technical contacts listed below or the appropriate Office ofNuclear Reactor Regulation project manager.Brian K. Grimes, Acting DirectorDivision of Reactor Program ManagementOffice of Nuclear Reactor RegulationTechnical contacts: Harry A. Freeman, RIV(817) 897-1500E-mail: haftnrc.govChu-Yu Liang, NRR(301) 415-2878E-mail: cyl~nrc.govAttachment: List of Recently Issued NRC Information NoticesDOCUMENT NAME: G:\SSK2\INFONOT.C PTo receive a copy of this document, indicate in the box C-Topy u/o attachment/enclosure E=Copy with attachment/enclosure N NocopyOFFICE CONT:i kd l BC:SRXBLl BC:iPECB lI (A)iD:iDRPM I _NAME CYLiang* RJones* AChaffee* BGrimesl _ HAFreeman*DATE 6/ 3/96 6/21/96 7/08/96 7/ /96* See previous concurrenceOFFICIAL KLLUKV UV X! |
| List of Recently Issued NRC Information NoticesDOCUMENT NAME: G:\SSK2\INFONOT.C PTo receive a copy of this document, indicate in the box C-Topy u/o attachment/enclosure E=Copy with attachment/enclosure N NocopyOFFICE CONT:i kd l BC:SRXBLl BC:iPECB lI (A)iD:iDRPM I _NAME CYLiang* RJones* AChaffee* BGrimesl _ HAFreeman*DATE 6/ 3/96 6/21/96 7/08/96 7/ /96* See previous concurrenceOFFICIAL KLLUKV UV X! | | IN 96-XXJuly XX, 1996 for the loss of extraction steam to the high-pressure heaters, which was thecause of the temperature difference. During the event, a level imbalanceoccurred between the two heater drain tanks, which resulted in the isolationof extraction steam.The NRC staff review of analyses of feedwater temperature events at similarfacilities revealed that most of these analyses assumed similar initiatingevents as the licensee analysis and had similar conclusions concerning theamount of feedwater temperature drop. The licensee has reanalyzed the eventpursuant to Section 50.59 of Title 10 of the Code of Federal Regulations toinclude a 119 *C [246 *F] feedwater temperature drop and concluded that allaccident analysis parameters remained within requirements.This information notice requires no specific action or written response. Ifyou have any questions about the information in this notice, please contactone of the technical contacts listed below or the appropriate Office ofNuclear Reactor Regulation project manager.Brian K. Grimes, Acting DirectorDivision of Reactor Program ManagementOffice of Nuclear Reactor RegulationTechnical contacts: Harry A. Freeman, RIV(817) 897-1500Internet:haf@nrc.govChu-Yu Liang, NRR(301) 415-2878Internet:cyl nrc.govAttachment: List of Recently Issued NRC Information NoticesDOCUMENT NAME: G:\SSK2\INFONOT.C PTo receive a copy of this document, Indicate in the box Conopy w/c attachment/enclosure EnCopy with attachment/enclosure N |
| IN 96-XXJuly XX, 1996 for the loss of extraction steam to the high-pressure heaters, which was thecause of the temperature difference. During the event, a level imbalanceoccurred between the two heater drain tanks, which resulted in the isolationof extraction steam.The NRC staff review of analyses of feedwater temperature events at similarfacilities revealed that most of these analyses assumed similar initiatingevents as the licensee analysis and had similar conclusions concerning theamount of feedwater temperature drop. The licensee has reanalyzed the eventpursuant to Section 50.59 of Title 10 of the Code of Federal Regulations toinclude a 119 *C [246 *F] feedwater temperature drop and concluded that allaccident analysis parameters remained within requirements.This information notice requires no specific action or written response. Ifyou have any questions about the information in this notice, please contactone of the technical contacts listed below or the appropriate Office ofNuclear Reactor Regulation project manager.Brian K. Grimes, Acting DirectorDivision of Reactor Program ManagementOffice of Nuclear Reactor RegulationTechnical contacts: Harry A. Freeman, RIV(817) 897-1500Internet:haf@nrc.govChu-Yu Liang, NRR(301) 415-2878Internet:cyl nrc.gov | |
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| | * NoOFFICE CONT: Ekd BC: SLB BC:PECB (A)D:DRPMNAME CYLiang* RJones* ACh)f BGrimesl ~~HAFreeman*tVtDATE 6/ 3/96 6/21/96 7/7/96 7/ /96OFFICIAL RECOR COPY* See previous concurrence |
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| * NoOFFICE CONT: Ekd BC: SLB BC:PECB (A)D:DRPMNAME CYLiang* RJones* ACh)f BGrimesl ~~HAFreeman*tVtDATE 6/ 3/96 6/21/96 7/7/96 7/ /96OFFICIAL RECOR COPY* See previous concurrence K-, /IN 96-XXJune XX, 1996 for the loss of extraction steam to ticause of the temperature difference.occurred between the two heater drainof extraction steam.he high-pressure heaters, which was theDuring the event, a level imbalancetanks, which resulted in the isolationThe NRC staff's review of analyses of feedwater temperature events at similarfacilities revealed that most of these analyses assumed similar initiatingevents as the licensee's analysis and had similar conclusions concerning theamount of feedwater temperature drop. The licensee has reanalyzed the eventpursuant to Section 50.59 of Title 10 of the Code of Federal Regulations toinclude a 119 'C [246 OF] feedwater temperature drop and concluded that allaccident analysis parameters remained within requirements.This information notice requires no specific action or written response. Ifyou have any questions about the information in this notice, please contactone of the technical contacts listed below or the appropriate Office ofNuclear Reactor Regulation project manager.Brian K. Grimes, Acting DirectorDivision of Reactor Program ManagementOffice of Nuclear Reactor RegulationTechnical contacts:Harry A. Freeman, RIV(817) 897-1500Internet:haffnrc.govChu-Yu Liang, NRR(301) 415-2878Internet:cyl@nrc.gov
| | K-, /IN 96-XXJune XX, 1996 Ioith attachment/enclosure 1 |
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| | IN 96-XXJune XX, 1996 detection system. The licensee believed that this system would probably notbe significantly affected by feedwater temperatures because of a differentmass flow rate determination method.Finally, the licensee's final safety analysis report did not accuratelyanalyze this transient. The actual events were similar to the analysis of the'Decrease in Feedwater Temperature event presented in Chapter 15. In thatanalysis, the inadvertent opening of the low-pressure heater bypass valve,coupled with the trip of the heater drain pumps, resulted in a feedwatertemperature drop of less than 35 OF, and a corresponding power increase ofless than 10 percent. In the actual event, the feedwater temperature droppedby approximately 200 OF, and the licensee calculated that reactor power wouldhave increased by approximately 35 percent without operator or protectiveactions. The licensee determined that although the initiating events were thesame, the Chapter 15 analysis did not account for the loss of extraction steamto the high-pressure heaters, which was the cause of the temperaturedifference. During the event, a level imbalance occurred between the twoheater drain tanks, which resulted in the isolation of extraction steam.The NRC staff's review of analyses of feedwater temperature events at similarfacilities revealed that most of these analyses assumed similar initiatingevents as the licensee's analysis and had similar conclusions concerning theamount of feedwater temperature drop.This information notice requires no specific action or written response. Ifyou have any questions about the information in this notice, please contactone of the technical contacts listed below or the appropriate Office ofNuclear Reactor Regulation project manager.Brian K. Grimes, Acting DirectorDivision of Reactor Program ManagementOffice of Nuclear Reactor RegulationTechnical contacts: Harry A. Freeman, RIV(817) 897-1500Internet:haf@nrc.govChu-Yu Liang, NRR(301) 415-2878Internet:cyl nrc.govAttachment: List of Recently Issued NRC Information NoticesDOCUMENT NAME: G:\SSK2\INFONOT.C PTo receive a copy of this docunent, indicate in the box Catopy w/o attachment/enclosure E-C with attachment/enclosure N |
| List of Recently Issued NRC Information NoticesDOCUMENT NAME: G:\SSK2\INFONOT.CPTo receive a copy of this document, indicate in the box Ciropy w/dattachmeft1/enctosure EnCOFFICE CONT:jkd _l BC: SRXB E C:PECB I _ A)D:DRPM INAME CYLiang* RJones AChaffee BGrimesHAFreeman* I- _DATE 6/ 3/96 6/2j /96 6/ /96 6/ /96OFFICIAL RECORD COPYIoith attachment/enclosure 1
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| * No copy* See previous concurrence IN 96-XXJune XX, 1996 detection system. The licensee believed that this system would probably notbe significantly affected by feedwater temperatures because of a differentmass flow rate determination method.Finally, the licensee's final safety analysis report did not accuratelyanalyze this transient. The actual events were similar to the analysis of the'Decrease in Feedwater Temperature event presented in Chapter 15. In thatanalysis, the inadvertent opening of the low-pressure heater bypass valve,coupled with the trip of the heater drain pumps, resulted in a feedwatertemperature drop of less than 35 OF, and a corresponding power increase ofless than 10 percent. In the actual event, the feedwater temperature droppedby approximately 200 OF, and the licensee calculated that reactor power wouldhave increased by approximately 35 percent without operator or protectiveactions. The licensee determined that although the initiating events were thesame, the Chapter 15 analysis did not account for the loss of extraction steamto the high-pressure heaters, which was the cause of the temperaturedifference. During the event, a level imbalance occurred between the twoheater drain tanks, which resulted in the isolation of extraction steam.The NRC staff's review of analyses of feedwater temperature events at similarfacilities revealed that most of these analyses assumed similar initiatingevents as the licensee's analysis and had similar conclusions concerning theamount of feedwater temperature drop.This information notice requires no specific action or written response. Ifyou have any questions about the information in this notice, please contactone of the technical contacts listed below or the appropriate Office ofNuclear Reactor Regulation project manager.Brian K. Grimes, Acting DirectorDivision of Reactor Program ManagementOffice of Nuclear Reactor RegulationTechnical contacts: Harry A. Freeman, RIV(817) 897-1500Internet:haf@nrc.govChu-Yu Liang, NRR(301) 415-2878Internet:cyl nrc.gov
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| List of Recently Issued NRC Information NoticesDOCUMENT NAME: G:\SSK2\INFONOT.C PTo receive a copy of this docunent, indicate in the box Catopy w/o attachment/enclosure E-C with attachment/enclosure N
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| * No copyOFFICE lCONT:kd l BC:SRXB l BC:PECB l (A)D:DRPMNAME CYLiang 9 RJones AChaffee BGrimesHAFreema r _ _DATE /96 /96 6/ /96 6/ /96OFFICIAL RECORD COPY}}
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Effects of a Decrease in Feedwater Temperature on Nuclear InstrumentationML031060009 |
Person / Time |
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Site: |
Beaver Valley, Millstone, Hatch, Monticello, Calvert Cliffs, Dresden, Davis Besse, Peach Bottom, Browns Ferry, Salem, Oconee, Mcguire, Nine Mile Point, Palisades, Palo Verde, Perry, Indian Point, Fermi, Kewaunee, Catawba, Harris, Wolf Creek, Saint Lucie, Point Beach, Oyster Creek, Watts Bar, Hope Creek, Grand Gulf, Cooper, Sequoyah, Byron, Pilgrim, Arkansas Nuclear, Three Mile Island, Braidwood, Susquehanna, Summer, Prairie Island, Columbia, Seabrook, Brunswick, Surry, Limerick, North Anna, Turkey Point, River Bend, Vermont Yankee, Crystal River, Haddam Neck, Ginna, Diablo Canyon, Callaway, Vogtle, Waterford, Duane Arnold, Farley, Robinson, Clinton, South Texas, San Onofre, Cook, Comanche Peak, Yankee Rowe, Maine Yankee, Quad Cities, Humboldt Bay, La Crosse, Big Rock Point, Rancho Seco, Zion, Midland, Bellefonte, Fort Calhoun, FitzPatrick, McGuire, LaSalle, Fort Saint Vrain, Shoreham, Satsop, Trojan, Atlantic Nuclear Power Plant |
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Issue date: |
07/26/1996 |
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From: |
Grimes B K Office of Nuclear Reactor Regulation |
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To: |
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References |
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IN-96-041, NUDOCS 9607220160 |
Download: ML031060009 (10) |
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Similar Documents at Beaver Valley, Millstone, Hatch, Monticello, Calvert Cliffs, Dresden, Davis Besse, Peach Bottom, Browns Ferry, Salem, Oconee, Mcguire, Nine Mile Point, Palisades, Palo Verde, Perry, Indian Point, Fermi, Kewaunee, Catawba, Harris, Wolf Creek, Saint Lucie, Point Beach, Oyster Creek, Watts Bar, Hope Creek, Grand Gulf, Cooper, Sequoyah, Byron, Pilgrim, Arkansas Nuclear, Three Mile Island, Braidwood, Susquehanna, Summer, Prairie Island, Columbia, Seabrook, Brunswick, Surry, Limerick, North Anna, Turkey Point, River Bend, Vermont Yankee, Crystal River, Haddam Neck, Ginna, Diablo Canyon, Callaway, Vogtle, Waterford, Duane Arnold, Farley, Robinson, Clinton, South Texas, San Onofre, Cook, Comanche Peak, Yankee Rowe, Maine Yankee, Quad Cities, Humboldt Bay, La Crosse, Big Rock Point, Rancho Seco, Zion, Midland, Bellefonte, Fort Calhoun, FitzPatrick, McGuire, LaSalle, Fort Saint Vrain, Shoreham, Satsop, Trojan, Atlantic Nuclear Power Plant |
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Category:NRC Information Notice
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Mclaughlin on NRC, Regarding NRC Information Notice 2006-13: Groundwater Contamination2006-07-13013 July 2006 E-mail from M. Mclaughlin on NRC, Regarding NRC Information Notice 2006-13: Groundwater Contamination 2020-09-03 The following query condition could not be considered due to this wiki's restrictions on query size or depth: <code> [[:Beaver Valley]] OR [[:Millstone]] OR [[:Hatch]] OR [[:Monticello]] OR [[:Calvert Cliffs]] OR [[:Dresden]] OR [[:Davis Besse]] OR [[:Peach Bottom]] OR [[:Browns Ferry]] OR [[:Salem]] OR [[:Oconee]] OR [[:Mcguire]] OR [[:Nine Mile Point]] OR [[:Palisades]] OR [[:Palo Verde]] OR [[:Perry]] OR [[:Indian Point]] OR [[:Fermi]] OR [[:Kewaunee]] OR [[:Catawba]] OR [[:Harris]] OR [[:Wolf Creek]] OR [[:Saint Lucie]] OR [[:Point Beach]] OR [[:Oyster Creek]] OR [[:Watts Bar]] OR [[:Hope Creek]] OR [[:Grand Gulf]] OR [[:Cooper]] OR [[:Sequoyah]] OR [[:Byron]] OR [[:Pilgrim]] OR [[:Arkansas Nuclear]] OR [[:Three Mile Island]] OR [[:Braidwood]] OR [[:Susquehanna]] OR [[:Summer]] OR [[:Prairie Island]] OR [[:Columbia]] OR [[:Seabrook]] OR [[:Brunswick]] OR [[:Surry]] OR [[:Limerick]] OR [[:North Anna]] OR [[:Turkey Point]] OR [[:River Bend]] OR [[:Vermont Yankee]] OR [[:Crystal River]] OR [[:Haddam Neck]] OR [[:Ginna]] OR [[:Diablo Canyon]] OR [[:Callaway]] OR [[:Vogtle]] OR [[:Waterford]] OR [[:Duane Arnold]] OR [[:Farley]] OR [[:Robinson]] OR [[:Clinton]] OR [[:South Texas]] OR [[:San Onofre]] OR [[:Cook]] OR [[:Comanche Peak]] OR [[:Yankee Rowe]] OR [[:Maine Yankee]] OR [[:Quad Cities]] OR [[:Humboldt Bay]] OR [[:La Crosse]] OR [[:Big Rock Point]] OR [[:Rancho Seco]] OR [[:Zion]] OR [[:Midland]] OR [[:Bellefonte]] OR [[:Fort Calhoun]] OR [[:FitzPatrick]] OR [[:McGuire]] OR [[:LaSalle]] OR [[:Fort Saint Vrain]] OR [[:Shoreham]] OR [[:Satsop]] OR [[:Trojan]] OR [[:Atlantic Nuclear Power Plant]] </code>.
[Table view]The following query condition could not be considered due to this wiki's restrictions on query size or depth: <code> [[:Beaver Valley]] OR [[:Millstone]] OR [[:Hatch]] OR [[:Monticello]] OR [[:Calvert Cliffs]] OR [[:Dresden]] OR [[:Davis Besse]] OR [[:Peach Bottom]] OR [[:Browns Ferry]] OR [[:Salem]] OR [[:Oconee]] OR [[:Mcguire]] OR [[:Nine Mile Point]] OR [[:Palisades]] OR [[:Palo Verde]] OR [[:Perry]] OR [[:Indian Point]] OR [[:Fermi]] OR [[:Kewaunee]] OR [[:Catawba]] OR [[:Harris]] OR [[:Wolf Creek]] OR [[:Saint Lucie]] OR [[:Point Beach]] OR [[:Oyster Creek]] OR [[:Watts Bar]] OR [[:Hope Creek]] OR [[:Grand Gulf]] OR [[:Cooper]] OR [[:Sequoyah]] OR [[:Byron]] OR [[:Pilgrim]] OR [[:Arkansas Nuclear]] OR [[:Three Mile Island]] OR [[:Braidwood]] OR [[:Susquehanna]] OR [[:Summer]] OR [[:Prairie Island]] OR [[:Columbia]] OR [[:Seabrook]] OR [[:Brunswick]] OR [[:Surry]] OR [[:Limerick]] OR [[:North Anna]] OR [[:Turkey Point]] OR [[:River Bend]] OR [[:Vermont Yankee]] OR [[:Crystal River]] OR [[:Haddam Neck]] OR [[:Ginna]] OR [[:Diablo Canyon]] OR [[:Callaway]] OR [[:Vogtle]] OR [[:Waterford]] OR [[:Duane Arnold]] OR [[:Farley]] OR [[:Robinson]] OR [[:Clinton]] OR [[:South Texas]] OR [[:San Onofre]] OR [[:Cook]] OR [[:Comanche Peak]] OR [[:Yankee Rowe]] OR [[:Maine Yankee]] OR [[:Quad Cities]] OR [[:Humboldt Bay]] OR [[:La Crosse]] OR [[:Big Rock Point]] OR [[:Rancho Seco]] OR [[:Zion]] OR [[:Midland]] OR [[:Bellefonte]] OR [[:Fort Calhoun]] OR [[:FitzPatrick]] OR [[:McGuire]] OR [[:LaSalle]] OR [[:Fort Saint Vrain]] OR [[:Shoreham]] OR [[:Satsop]] OR [[:Trojan]] OR [[:Atlantic Nuclear Power Plant]] </code>. |
K) K)UNITED STATESNUCLEAR REGULATORY COMMISSIONOFFICE OF NUCLEAR REACTOR REGULATIONWASHINGTON, D.C. 20555-0001July 26, 1996NRC INFORMATION NOTICE 96-41: EFFECTS OF A DECREASE IN FEEDWATER TEMPERATUREON NUCLEAR INSTRUMENTATION
Addressees
All holders of operating licenses or construction permits for pressurizedwater reactors (PWRs).
Purpose
The U.S. Nuclear Regulatory Commission (NRC) is issuing this informationnotice to alert addressees to the potential for operation above licensed poweras a result of a decrease in feedwater temperature event affecting nuclearinstrumentation. It is expected that recipients will review the informationfor applicability to their facilities and consider actions, as appropriate, toavoid similar problems. However, suggestions contained in this informationnotice are not NRC requirements; therefore, no specific action or writtenresponse is required.
Description of Circumstances
On February 14, 1996, the licensee for the Comanche Peak Steam ElectricStation was operating Unit 2 at 95 percent rated thermal power near end-of-core life when a significant reduction in feedwater temperature occurredbecause of the loss of feedwater heaters. This reduction, in turn, caused areduction in the reactor coolant system cold-leg temperatures. The colderreactor coolant temperature, with a large negative moderator temperaturecoefficient, caused reactor power to increase to approximately 102 percentaccording to ex-core nuclear instrumentation. The nitrogen-16 (N-16)detection system reached the overpower turbine runback setpoint (109 percent)and initiated a turbine runback. The N-16 detection system measures N-16activity in the primary coolant as a measure of the total power generation.This system is a substitute for the resistance temperature detector over-temperature and over-power reactor trip functions used at other WestinghousePWRs. The plant stabil zed at an indicated power of approximately 97 percentaccording to the ex-core nuclear instrumentation.After approximately 90 minutes, a second similar turbine runback occurredwhile restoring balance-of-plant equipment. Following this runback, reactorpower was stabilized at approximately 100 percent according to nuclearinstrumentation. During the next 30 minutes, the reactor was operated atapproximately 100 percent power as indicated by nuclear instrumentation, withreactor coolant temperatures below normal. The licensee noted that the N-169607220l60ujo i 7 9,oi4(R ~IE ctG
IN 96-41July 26, 1996 detection system indicated approximately 106 percent power and the computer-based plant calorimetric system indicated approximately 102 percent power.Subsequently, the reactor power was reduced to less than 100 percent by allindications.DiscussionThere are three aspects of this event which have generic implications. First,with a loss of secondary plant efficiency, programmed T e can no longerreliably represent core thermal power. Second, the venturi-based input intothe computer-based calorimetric system may not be accurate with coldfeedwater. And third, the final safety analysis report had not analyzed thistransient accurately.Following the second runback, operators noted that reactor power indicated<100 percent according to nuclear instrumentation. Although the operatorsknew that cold feedwater could cause an increase in the amount of neutronattenuation, they believed that the nuclear instrumentation indicatedconservatively (i.e., higher than actual) because they were maintaining TA"eapproximately 1.7 eC [3 OF] above TRef. The licensee could not use thecomputer-based calorimetric until some time after the second turbine runbackdue to maintenance activities. Te , based on the main turbine impulsepressure, is programmed as a functlon of turbine load and, for normalefficiency, is a good representation of thermal power. When the unit lost thefeedwater heaters, the plant efficiency decreased. Because the main turbineelectro-hydraulic control system maintained generator output, core thermalpower increased to account for the loss of efficiency, and thus, TRef nolonger accurately represented the core thermal power.The cold-leg temperature is a more appropriate indicator of the accuracy ofthe nuclear instrumentation than programmed TY.e. As the cold-leg temperaturedecreased, the amount of neutron attenuation in the downcomer area surroundingthe core increased and hence affected the amount of neutrons reaching thedetectors. The licensee analysis showed that for every 0.6 C (1 OF] of cold-leg temperature change, the nuclear instrumentation was affected by 0.6 to 0.8percent power. A review of the second transient showed that the cold-legtemperature was approximately 2.5 °C [4.5 OF] lower than when the detectorswere last calibrated. This corresponded to a 3 to 4 percent error, whichcorresponded to the difference in the actual versus the indicated power (104percent actual versus 100 percent indicated).During the review, the licensee noted that the computer-based calorimetric was4 percent lower than the actual thermal power (N-16 power monitor). Thecalorimetric was based on feedwater flow measured by venturis. Although thecalorimetric calculation used feedwater temperature as an input, temperaturessignificantly different than the normal 227 OC [440 OF] introduced errors intothe calculation.Finally, the actual events involved temperature and power levels that exceededthose in the analysis of the Decrease in Feedwater Temperature" eventpresented in Chapter 15 of the licensee final safety analysis report. In that
IN 96-41July 26, 1996 analysis, the inadvertent opening of the low-pressure heater bypass valve,coupled with the trip of the heater drain pumps, resulted in a feedwatertemperature drop of less than 19 'C (35 OF], and a corresponding powerincrease of less than 10 percent. In the actual event, the feedwatertemperature dropped by approximately 111 °C (200 OF], and the licenseecalculated that reactor power would have increased by approximately 35 percentwithout operator or protective actions. The licensee determined that althoughthe initiating events were the same, the Chapter 15 analysis did not accountfor the loss of extraction steam to the high-pressure heaters, which was thecause of the temperature difference. During the event, a level imbalanceoccurred between the two heater drain tanks, which resulted in the isolationof extraction steam.The NRC staff review of analyses of feedwater temperature events at similarfacilities revealed that most of these analyses assumed similar initiatingevents as the Comanche Peak analysis and had similar conclusions concerningthe amount of feedwater temperature drop. The licensee has reanalyzed theevent to include a 119 OC [246 OF] feedwater temperature drop and concludedthat all accident analysis parameters remained within requirements.This information notice requires no specific action or written response. Ifyou have any questions about the information in this notice, please contactone of the technical contacts listed below or the appropriate Office ofNuclear Reactor Regulation project manager.Brian K. Grimes, Acting DirectorDivision of Reactor Program ManagementOffice of Nuclear Reactor RegulationTechnical contacts: Harry A. Freeman, RIV(817) 897-1500E-mail: haf~nrc.govChu-Yu Liang, NRR(301) 415-2878E-mail: cylenrc.govAttachment: List Of Recently Issued HRC Information NoticesA1h4 Stir A Je6tQ
K> KJAttachmentIN 96-41July 26, 1996 LIST OF RECENTLY ISSUEDNRC INFORMATION NOTICESInformation Date ofNotice No. Subject Issuance Issued to96-4096-09,Supp. 196-3996-38Deficiencies in MaterialDedication and Procure-ment Practices and inAudits of VendorsDamage in Foreign SteamGenerator InternalsEstimates of Decay HeatUsing ANS 5.1 Decay HeatStandard May Vary Signi-ficantlyResults of Steam GeneratorTube ExaminationsInaccurate Reactor WaterLevel Indication and Inad-vertent Draindown DuringShutdownDegradation of CoolingWater Systems Due to IcingFailure of Safety Systemson Self-Shielded Irradia-tors Because of InadequateMaintenance and TrainingHydrogen Gas Ignitionduring Closure Weldingof a VSC-24 Multi-AssemblySealed Basket07/25/9607/10/9607/05/9606/21/9606/18/9606/12/9606/11/9605/31/96All holders of OLs or CPsfor nuclear power reactorsAll holders of OLs or CPsfor pressurized-waterreactorsAll holders of OLs or CPsfor nuclear power reactorsAll holders of OLs or CPsfor pressurized waterreactorsAll pressurized waterreactor facilities holdingan operating license or aconstruction permitAll holders of OLs or CPsfor nuclear power reactorsAll U.S. Nuclear RegulatoryCommission irradiatorlicensees and vendorsAll holders of OLs or CPsfor nuclear power reactors96-3796-3696-3596-34OL -Operating LicenseCP -Construction Permit
- ~ -K> KIN 96-41July 26, 1996 analysis, the inadvertent opening of the low-pressure heater bypass valve,coupled with the trip of the heater drain pumps, resulted in a feedwatertemperature drop of less than 19 *C [35 OF], and a corresponding powerincrease of less than 10 percent. In the actual event, the feedwatertemperature dropped by approximately 111 *C [200 OF], and the licenseecalculated that reactor power would have increased by approximately 35 percentwithout operator or protective actions. The licensee determined that althoughthe initiating events were the same, the Chapter 15 analysis did not accountfor the loss of extraction steam to the high-pressure heaters, which was thecause of the temperature difference. During the event, a level imbalanceoccurred between the two heater drain tanks, which resulted in the isolationof extraction steam.The NRC staff review of analyses of feedwater temperature events at similarfacilities revealed that most of these analyses assumed similar initiatingevents as the Comanche Peak analysis and had similar conclusions concerningthe amount of feedwater temperature drop. The licensee has reanalyzed theevent to include a 119 *C [246 OF] feedwater temperature drop and concludedthat all accident analysis parameters remained within requirements.This information notice requires no specific action or written response. Ifyou have any questions about the information in this notice,-please contactone of the technical contacts listed below or the appropriate Office ofNuclear Reactor Regulation project manager.Original signed by Brian K. GrimesBrian K. Grimes, Acting DirectorDivision of Reactor Program ManagementOffice of Nuclear Reactor RegulationTechnical contacts: Harry A. Freeman, RIV(817) 897-1500E-mail: haf@nrc.govChu-Yu Liang, NRR(301) 415-2878E-mail: cyl~nrc.govAttachment: List of Recently Issued NRC Information NoticesDOCUMENT NAME: G:\SSK2\INFONOT.C PTo receive a copy of this docunent, tndicate in the box CO~opy So attachment/enclosure EsCopy with attachment/enctosure N
- NocopsOFFICE C BC:SRXB I BC:LPECB lI (A) DW M iNAME CYLiang* RJones* AChaffee*HAFreeman* ____ _DATE 16/ 3/96 16/21/96 17/08/96 17LI/96 IOFFILIAL MLLUM LWUF* See previous concurrence Tech Editor reviewed & concurred on 05/28/96
~1~1 -,K)IN 96-XXJuly XX, 1996 *See previous concurrence
IN 96-XXJuly XX, 1996 for the loss of extraction steam to the high-pressure heaters, which was thecause of the temperature difference. During the event, a level imbalanceoccurred between the two heater drain tanks, which resulted in the isolationof extraction steam.The NRC staff review of analyses of feedwater temperature events at similarfacilities revealed that most of these analyses assumed similar initiatingevents as the licensee analysis and had similar conclusions concerning theamount of feedwater temperature drop. The licensee has reanalyzed the eventpursuant to Section 50.59 of Title 10 of the Code of Federal Regulations toinclude a 119 'c [246 OF] feedwater temperature drop and concluded that allaccident analysis parameters remained within requirements.This information notice requires no specific action or written response. Ifyou have any questions about the information in this notice, please contactone of the technical contacts listed below or the appropriate Office ofNuclear Reactor Regulation project manager.Brian K. Grimes, Acting DirectorDivision of Reactor Program ManagementOffice of Nuclear Reactor RegulationTechnical contacts: Harry A. Freeman, RIV(817) 897-1500E-mail: haftnrc.govChu-Yu Liang, NRR(301) 415-2878E-mail: cyl~nrc.govAttachment: List of Recently Issued NRC Information NoticesDOCUMENT NAME: G:\SSK2\INFONOT.C PTo receive a copy of this document, indicate in the box C-Topy u/o attachment/enclosure E=Copy with attachment/enclosure N NocopyOFFICE CONT:i kd l BC:SRXBLl BC:iPECB lI (A)iD:iDRPM I _NAME CYLiang* RJones* AChaffee* BGrimesl _ HAFreeman*DATE 6/ 3/96 6/21/96 7/08/96 7/ /96* See previous concurrenceOFFICIAL KLLUKV UV X!
IN 96-XXJuly XX, 1996 for the loss of extraction steam to the high-pressure heaters, which was thecause of the temperature difference. During the event, a level imbalanceoccurred between the two heater drain tanks, which resulted in the isolationof extraction steam.The NRC staff review of analyses of feedwater temperature events at similarfacilities revealed that most of these analyses assumed similar initiatingevents as the licensee analysis and had similar conclusions concerning theamount of feedwater temperature drop. The licensee has reanalyzed the eventpursuant to Section 50.59 of Title 10 of the Code of Federal Regulations toinclude a 119 *C [246 *F] feedwater temperature drop and concluded that allaccident analysis parameters remained within requirements.This information notice requires no specific action or written response. Ifyou have any questions about the information in this notice, please contactone of the technical contacts listed below or the appropriate Office ofNuclear Reactor Regulation project manager.Brian K. Grimes, Acting DirectorDivision of Reactor Program ManagementOffice of Nuclear Reactor RegulationTechnical contacts: Harry A. Freeman, RIV(817) 897-1500Internet:haf@nrc.govChu-Yu Liang, NRR(301) 415-2878Internet:cyl nrc.govAttachment: List of Recently Issued NRC Information NoticesDOCUMENT NAME: G:\SSK2\INFONOT.C PTo receive a copy of this document, Indicate in the box Conopy w/c attachment/enclosure EnCopy with attachment/enclosure N
- NoOFFICE CONT: Ekd BC: SLB BC:PECB (A)D:DRPMNAME CYLiang* RJones* ACh)f BGrimesl ~~HAFreeman*tVtDATE 6/ 3/96 6/21/96 7/7/96 7/ /96OFFICIAL RECOR COPY* See previous concurrence
K-, /IN 96-XXJune XX, 1996 Ioith attachment/enclosure 1
- No copy* See previous concurrence
IN 96-XXJune XX, 1996 detection system. The licensee believed that this system would probably notbe significantly affected by feedwater temperatures because of a differentmass flow rate determination method.Finally, the licensee's final safety analysis report did not accuratelyanalyze this transient. The actual events were similar to the analysis of the'Decrease in Feedwater Temperature event presented in Chapter 15. In thatanalysis, the inadvertent opening of the low-pressure heater bypass valve,coupled with the trip of the heater drain pumps, resulted in a feedwatertemperature drop of less than 35 OF, and a corresponding power increase ofless than 10 percent. In the actual event, the feedwater temperature droppedby approximately 200 OF, and the licensee calculated that reactor power wouldhave increased by approximately 35 percent without operator or protectiveactions. The licensee determined that although the initiating events were thesame, the Chapter 15 analysis did not account for the loss of extraction steamto the high-pressure heaters, which was the cause of the temperaturedifference. During the event, a level imbalance occurred between the twoheater drain tanks, which resulted in the isolation of extraction steam.The NRC staff's review of analyses of feedwater temperature events at similarfacilities revealed that most of these analyses assumed similar initiatingevents as the licensee's analysis and had similar conclusions concerning theamount of feedwater temperature drop.This information notice requires no specific action or written response. Ifyou have any questions about the information in this notice, please contactone of the technical contacts listed below or the appropriate Office ofNuclear Reactor Regulation project manager.Brian K. Grimes, Acting DirectorDivision of Reactor Program ManagementOffice of Nuclear Reactor RegulationTechnical contacts: Harry A. Freeman, RIV(817) 897-1500Internet:haf@nrc.govChu-Yu Liang, NRR(301) 415-2878Internet:cyl nrc.govAttachment: List of Recently Issued NRC Information NoticesDOCUMENT NAME: G:\SSK2\INFONOT.C PTo receive a copy of this docunent, indicate in the box Catopy w/o attachment/enclosure E-C with attachment/enclosure N
- No copyOFFICE lCONT:kd l BC:SRXB l BC:PECB l (A)D:DRPMNAME CYLiang 9 RJones AChaffee BGrimesHAFreema r _ _DATE /96 /96 6/ /96 6/ /96OFFICIAL RECORD COPY
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list | - Information Notice 1996-01, Potential For High Post-Accident Closed-Cycle Cooling Water Temperatures to Disable Equipment Important to Safety (3 January 1996)
- Information Notice 1996-01, Potential for High Post-Accident Closed-Cycle Cooling Water Temperatures to Disable Equipment Important to Safety (3 January 1996)
- Information Notice 1996-02, Inoperability of Power-Operated Relief Valves Masked by Downstream Indications During Testing (5 January 1996, Topic: Stroke time)
- Information Notice 1996-03, Main Steam Safety Valve Setpoint Variation as a Result of Thermal Effects (5 January 1996)
- Information Notice 1996-03, Main Steam Safety Valve Setpoint Variation As a Result of Thermal Effects (5 January 1996)
- Information Notice 1996-04, Incident Reporting Requirements for Radiography Licensees (10 January 1996, Topic: Brachytherapy, Overexposure, Depleted uranium)
- Information Notice 1996-05, Partial Bypass of Shutdown Cooling Flow from Reactor Vessel (18 January 1996, Topic: Reactor Vessel Water Level)
- Information Notice 1996-06, Design & Testing Deficiencies of Tornado Dampers at Nuclear Power Plants (25 January 1996)
- Information Notice 1996-07, Slow Five Percent Scram Insertion Times Caused by Viton Diaphragms in Scram Solenoid Pilot Valves (26 January 1996)
- Information Notice 1996-08, Thermally Induced Pressure Locking of a High Pressure Coolant Injection Gate Valve (5 February 1996, Topic: Anchor Darling, Cold shutdown justification)
- Information Notice 1996-09, Damage in Foreign Steam Generator Internals (12 February 1996, Topic: Eddy Current Testing, Earthquake)
- Information Notice 1996-10, Potential Blockage by Debris of Safety System Piping Which Is Not Used During Normal Operation or Tested During Surveillances (13 February 1996)
- Information Notice 1996-10, Potential Blockage by Debris of Safety System Piping Which is Not Used During Normal Operation or Tested During Surveillances (13 February 1996)
- Information Notice 1996-11, Ingress of Demineralizer Resins Increases Potential For Stress Corrosion Cracking of Control Rod Drive Mechanism Penetrations (14 February 1996, Topic: Intergranular Stress Corrosion Cracking, Stress corrosion cracking)
- Information Notice 1996-11, Ingress of Demineralizer Resins Increases Potential for Stress Corrosion Cracking of Control Rod Drive Mechanism Penetrations (14 February 1996, Topic: Intergranular Stress Corrosion Cracking, Stress corrosion cracking)
- Information Notice 1996-12, Control Rod Insertion Problems (15 February 1996, Topic: Stress corrosion cracking)
- Information Notice 1996-13, Potential Containment Leak Paths Through Hydrogen Analysis (26 February 1996, Topic: Stress corrosion cracking, Integrated leak rate test)
- Information Notice 1996-14, Degradation of Radwaste Facility Equipment at Millstone Nuclear Power Station, Unit 1 (1 March 1996, Topic: Stress corrosion cracking)
- Information Notice 1996-15, Unexpected Plant Performance During Performance of New Surveillance (8 March 1996, Topic: Stress corrosion cracking)
- Information Notice 1996-16, BWR Operation with Indicated Flow Less than Natural Circulation (14 March 1996, Topic: Stress corrosion cracking)
- Information Notice 1996-17, Reactor Operation Inconsistent with the Updated Final Safety Analysis Report (18 March 1996)
- Information Notice 1996-18, Compliance with 10 CFR Part 20 for Airborne Thorium (25 March 1996, Topic: Stress corrosion cracking, Brachytherapy)
- Information Notice 1996-19, Failure of Tone Alert Radios to Activate When Receiving a Shortened Activation Signal (2 April 1996, Topic: Tone Alert Radio, Siren)
- Information Notice 1996-20, Demonstration of Associated Equipment Compliance with 10 CFR 34.20 (4 April 1996, Topic: Brachytherapy)
- Information Notice 1996-21, Safety Concerns Related to the Design of the Door Interlock Circuit on Nucletron High-Dose Rate and Pulsed Dose Rate Remote Afterloading Brachytherapy Devices (10 April 1996, Topic: Brachytherapy)
- Information Notice 1996-22, Improper Equipment Settings Due to Use of Nontemperature-Compensated Test Equipment (11 April 1996, Topic: Brachytherapy)
- Information Notice 1996-23, Fires in Emergency Diesel Generator Exciters During Operation Following Undetected Fuse Blowing (22 April 1996, Topic: Brachytherapy, Overspeed trip, Overspeed)
- Information Notice 1996-24, Preconditioning of Molded-Case Circuit Breakers Before Surveillance Testing (25 April 1996, Topic: Brachytherapy)
- Information Notice 1996-25, Traversing In-Core Probe Overwithdrawn at Lasalle County Station, Unit 1 (30 April 1996, Topic: Brachytherapy)
- Information Notice 1996-26, Recent Problems with Overhead Cranes (30 April 1996, Topic: Brachytherapy)
- Information Notice 1996-26, Recent Problems With Overhead Cranes (30 April 1996)
- Information Notice 1996-27, Potential Clogging of High Pressure Safety Injection Throttle Valves During Recirculation (1 May 1996, Topic: Brachytherapy)
- Information Notice 1996-28, Suggested Guidance Relating to Development and Implementation of Corrective Action (1 May 1996, Topic: Brachytherapy)
- Information Notice 1996-29, Requirements in 10 CFR Part 21 for Reporting and Evaluating Software Errors (20 May 1996, Topic: Brachytherapy)
- Information Notice 1996-30, Inaccuracy of Diagnostic Equipment for Motor-Operated Butterfly Valves (21 May 1996)
- Information Notice 1996-31, Cross-Tied Safety Injection Accumulators (22 May 1996)
- Information Notice 1996-32, Implementation of 10 CFR 50.55a(g) (6) (II) (A), Augmented Examination of Reactor Vessel (5 June 1996, Topic: Non-Destructive Examination)
- Information Notice 1996-32, Implementation of 10 CFR 50.55a(g) (6) (ii) (A), Augmented Examination of Reactor Vessel (5 June 1996, Topic: Non-Destructive Examination)
- Information Notice 1996-33, Erroneous Data From Defective Thermocouple Results in a Fire (24 May 1996, Topic: Reverse polarity)
- Information Notice 1996-33, Erroneous Data from Defective Thermocouple Results in a Fire (24 May 1996, Topic: Reverse polarity)
- Information Notice 1996-34, Hydrogen Gas Ignition During Closure Welding of a VSC-24 Multi-Assembly Sealed Basket (31 May 1996)
- Information Notice 1996-35, Failure of Safety Systems on Self-Shielded Irradiators Because of Inadequate Maintenance and Training (11 June 1996)
- Information Notice 1996-36, Degradation of Cooling Water Systems Due to Icing (12 June 1996, Topic: High winds, Ultimate heat sink, Frazil ice)
- Information Notice 1996-37, Inaccurate Reactor Water Level Indication and Inadvertent Draindown During Shutdown (18 June 1996, Topic: Reactor Vessel Water Level)
- Information Notice 1996-38, Results of Steam Generator Tube Examinations (21 June 1996, Topic: Stress corrosion cracking)
- Information Notice 1996-39, Estimates of Decay Heat Using ANS 5.1 Decay Heat Standard May Vary Significantly (5 July 1996)
- Information Notice 1996-40, Defciencies in Material Dedication and Procurement Practices and in Audits of Vendors (7 October 1996, Topic: Coatings, Commercial Grade, Troxler)
- Information Notice 1996-41, Effects of a Decrease in Feedwater Temperature on Nuclear Instrumentation (26 July 1996, Topic: Feedwater Heater)
- Information Notice 1996-42, Unexpected Opening of Multiple Safety Relief Valves (5 August 1996, Topic: Reactor Vessel Water Level)
- Information Notice 1996-43, Failures of General Electric Magne-Blast Circuit Breakers (2 August 1996, Topic: Hardened grease)
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