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| | issue date = 02/09/1998 | | | issue date = 02/09/1998 |
| | title = LER 98-001-00:on 980109,potential Condition Outside Design Basis Related to Instrument Air Sys Leak Causing SG pre- Heater Bypass Isolation Valves to Be Inoperable Was Noted. Caused by Inadequate Design Control.Generated Jco 98-01 | | | title = LER 98-001-00:on 980109,potential Condition Outside Design Basis Related to Instrument Air Sys Leak Causing SG pre- Heater Bypass Isolation Valves to Be Inoperable Was Noted. Caused by Inadequate Design Control.Generated Jco 98-01 |
| | author name = VERRILLI M | | | author name = Verrilli M |
| | author affiliation = CAROLINA POWER & LIGHT CO. | | | author affiliation = CAROLINA POWER & LIGHT CO. |
| | addressee name = | | | addressee name = |
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| {{#Wiki_filter:NRC FORM 366 (4.95)U.s.NUCLEAR REGULATORY COMMISSION | | {{#Wiki_filter:NRC FORM 366 U.s. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB No. 3150-0104 (4.95) |
| 'LXCENSEE EVENT REPORT (LER)(Sae reverse for required number of digits/characters for each block)APPROVED BY OMB No.3150-0104 EXPIRES 04/30/96 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS MANDATORY INFORMATION COLLECTION REDUEST.509 HRS.REPORTED LESSONS LEARNED ARE INCORPORATED INTO THE LICENSING PROCESS AND FEO BACK TO UIOUSTRY.FORWARD COMMENTS REGAROIHG BURDEN ESTIMATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH Irk F33L US.NUCLEAR REGULATORY COMMLSSION, WASHINGTON, OC 20555000l, ANO TO THE PAPERV/ORK REDUCTION PRDIECT ISI50.OIOIL OFFICE OF MANAGEMENT ANO BUDGET, WASHUIGTON, OC 205IL FACIUTY NAME lll Harris Nuclear Plant Unit-1 DOCKET NUMBER I2)50-400 PAGE I3I 1 OF4 TITLE (4I Potential Condition Outside Design Basis related to Instrument Air System Leak causing the S/G Pre-heater Bypass Isolation Valves to be inoperable.
| | EXPIRES 04/30/96 |
| EVENT DATE (5)LER NUMBER{6)MONTH DAY 1 9 SEOUENTIAL REVISION NUMBER NUMBER 98 98-001-00 REPORT DATE{7)MONTH DAY YEAR 2 9 98 FACIUTY NAME FACIUTY NAME OTHER FACILITIES INVOLVED (6)DOCKET NUMBER DOCKET NUMBER 05000 OPERATING MODE (9)POWER LEVEL (10)NAME 100%20.2201{b)20.2203(a) | | ' ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS MANDATORY INFORMATION COLLECTION REDUEST. 509 HRS. REPORTED LESSONS LEARNED ARE LXCENSEE EVENT REPORT (LER) INCORPORATED INTO THE LICENSING PROCESS AND FEO BACK TO UIOUSTRY. |
| (1)20.2203(a) | | FORWARD COMMENTS REGAROIHG BURDEN ESTIMATE TO THE INFORMATION AND (Sae reverse for required number of RECORDS MANAGEMENT BRANCH Irk F33L US. NUCLEAR REGULATORY COMMLSSION, WASHINGTON, OC 20555000l, ANO TO THE PAPERV/ORK REDUCTION PRDIECT ISI50. |
| (2)(i)20.2203(a)
| | digits/characters for each block) OIOIL OFFICE OF MANAGEMENT ANO BUDGET, WASHUIGTON, OC 205IL FACIUTY NAME lll DOCKET NUMBER I2) PAGE I3I Harris Nuclear Plant Unit-1 50-400 1 OF4 TITLE (4I Potential Condition Outside Design Basis related to Instrument Air System Leak causing the S/G Pre-heater Bypass Isolation Valves to be inoperable. |
| (2)(ii)20.2203(a) | | EVENT DATE (5) LER NUMBER {6) REPORT DATE {7) OTHER FACILITIES INVOLVED (6) |
| (2)(iii)20.2203(a) | | FACIUTY NAME DOCKET NUMBER MONTH DAY SEOUENTIAL REVISION MONTH DAY YEAR NUMBER NUMBER FACIUTY NAME 1 9 98 98 001 00 2 9 98 DOCKET NUMBER 05000 OPERATING THIS REPORT IS SUBMITTED PURSUANT To THE REQUIREMENTS OF 10 CFR 5: {Chock one or more) (11) |
| (2)(iv)50.73(a){2)(i)50.73(a)(2){ii)20.2203(a)(2)(v) 20.2203(a)
| | MODE (9) 20.2201 {b) 20.2203(a)(2)(v) 50.73(a) {2)(i) 50.73(a)(2)(viii) |
| {3){i)50.73(a)(2)(iii) 50.73(a)(2)(iv)20.2203(a)
| | POWER 20.2203(a) (1) 20.2203(a) {3){i) 50.73(a) (2) {ii) 50.73(a) {2)(x) |
| (3)(ii)20.2203(a) | | LEVEL (10) 100% |
| (4)50.36(c){1)50.73(a)(2)(v)50.73(a)(2)(vii)50.36(c)(2)LICENSEE CONTACT FOR THIS LER (12)TELEPHONE NUMBER Ilnalvde Ates Cade)50.73(a)(2)(viii) 50.73(a){2)(x)73.71 OTHER Specify in Abstract below or in NRC Form 36BA THIS REPORT IS SUBMITTED PURSUANT To THE REQUIREMENTS OF 10 CFR 5:{Chock one or more)(11)Michael Verrilli Sr.Analyst-Licensing (919)362-2303 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE TO NPRDS CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE TO NPRDS SUPPLEMENTAL REPORT EXPECTED (14)YES{If yas, complete EXPECTED SUBMISSION DATE).X No EXPECTED SUBMISSION DATE{15)MONTH DAY YEAR ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 singla-spaced typawrittan lines)(16)On January 9, 1998, a condition was identified during operation that results in the plant being potentially outside it'design basis.Specifically, a potential failure mechanism exists where a leak in the non-safety Instrument Air System could result in the inoperabihty of the Steam Generator Pre-heater Bypass Isolation Valves.These valves are safety-related containment isolation valves that are required by plant procedures to automatically shut in 10 seconds or less upon receipt of a Main Feedwater Isolation Signal.These valves are positioned by a pneumatic piston-operated actuator which is supplied by the non-safety related Instrument Air System.They are designed to automatically close if control air supply is lost.However, a"smart" air leak has been postulated in the Instrument Air system that could possibly reduce the air inlet pressure to just low enough to affect proper operation of the actuator's 3-way and 4-way pilot valves and not be detected by Operations personnel. | | 20.2203(a) (2) (i) 20.2203(a) (3) (ii) 50.73(a)(2)(iii) 73.71 20.2203(a) (2)(ii) 20.2203(a) (4) 50.73(a) (2)(iv) OTHER 20.2203(a) (2) (iii) 50.36(c) {1) 50.73(a)(2) (v) Specify in Abstract below or in NRC Form 36BA 20.2203(a) (2)(iv) 50.36(c) (2) 50.73(a) (2) (vii) |
| If this occurred, the pilot valves would shuttle, causing the accumulator pressure to bleed off, which would prevent the valves from closing as required.This potential scenario constitutes operation outside the design basis of the plant and was reported to the NRC via the emergency notification system on January 9, 1998 at 1450 hours.The cause of this condition was inadequate design control during development of a plant modification implemented in August 1984 in response to NRC Information Notice 82-25.The investigation for this event also revealed several other missed opportunities to identify this condition during subsequent plant modifications and/or related evaluations.
| | LICENSEE CONTACT FOR THIS LER (12) |
| Immediate corrective actions included development of a Justification for Continued Operation (JCO), with the required evaluation and compensatory measures to ensure continued operability.
| | NAME TELEPHONE NUMBER Ilnalvde Ates Cade) |
| The JCO evaluation determined that the isolation valves were operable dependent upon once per 12-hour monitoring of the discharge pressure on the inlet regulator for each valve actuator.Additional corrective actions will include a plant modification to resolve the design deficiency, a FSAR revision and training for appropriate Engineering personnel.
| | Michael Verrilli Sr. Analyst - Licensing (919) 362-2303 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13) |
| 9802170061 980209 PDR ADCICK 05000400 S PDR NRC FORM 366A M-SS)LICENSEE EVENT REPORT (LER)TEXT CONTINUATlON US.NUCLEAR REGULATORY COMMISSION FACIUTY NAME (I)Shearon Harris NUclear Plant-Unit¹1 TEXT Pr more spscers nod.vso oddMmsl copies of ABC Povv 3IFQI (11)DOCKET 50400 LER NUMBER (6)YEAR SEOUENTIAL REVISION NUMBER NUMBER 98-001-00 PAGE (3)2 OF 4 EVENT DESCRIPTION:
| | CAUSE SYSTEM COMPONENT REPORTABLE REPORTABLE MANUFACTURER CAUSE SYSTEM COMPONENT MANUFACTURER TO NPRDS TO NPRDS SUPPLEMENTAL REPORT EXPECTED (14) MONTH DAY YEAR EXPECTED YES SUBMISSION |
| On January 9, 1998, with the plant operating in mode 1 at 100%power, a condition was identified that results in the plant being potentially outside the design basis.Specifically, a potential failure mechanism exists where a leak in the non-safety Instrument Air System (EIIS Code: LD)could result in the inoperability of the Steam Generator (S/G)Pre-heater Bypass Isolation Valves (1AF-64, 1AF-102, and 1AF-81, EIIS Code: SJ-V).These valves are safety-related containment isolation valves that are required by plant procedure PLP-106 (Technical Specification Equipment List and Core Operating Limit Report)to automatically shut in 10 seconds or less upon receipt of a Main Feedwater Isolation Signal (MFIS).These valves are positioned by a pneumatic piston-operated actuator and are opened and closed by high pressure air (-150 psig)from the actuator's accumulator.
| | {Ifyas, complete EXPECTED SUBMISSION DATE). X No DATE {15) |
| The accumulator is maintained at a higher air pressure by an air intensifier pump.The intensifier is a double piston compressor that is driven by control air.The air intensifier is needed to boost inlet control air pressure to approximately 150 psig, since normal control air pressure (approx.75 psig)will not close the valve alone.With the use of pneumatic pilot valves located on the actuator, the intensified air in the accumulator is used to position the actuator piston, which is connected to the valve disc by a common shaft.The pneumatic pilot valves guide the intensified air from the accumulator to either face of the piston to position the operator.Directing the high pressure air below the piston will open the S/G pre-heater bypass valves;and directing the high pressure air to the top of the piston will cause the valves to close.The S/G pre-heater bypass valves are designed to automatically close if control air supply is lost.This is accomplished by two in-series solenoid valves that are energized open to supply the control air from the Instrument Air System.If a Main Feedwater Isolation signal is generated the solenoid valves will de-energize, securing the control air supply, which causes the S/G pre-heater bypass isolation valves to close.The solenoid valves will also de-energize if Instrument Air System pressure drops to 66 psig.However, a"smart" air leak has been postulated in the Instrument Air System piping that could possibly reduce the air inlet pressure to just low enough to affect proper operation of the actuator's 3-way and 4-way pilot valves and not be detected by the pressure switches in the main header of the Instrument Air System that would de-energize the solenoid valves at 66 psig.If this occurred, the pilot valves would shuttle, causing the accumulator pressure to bleed off, which would prevent the valves from closing as required.Operations personnel would have no indication of accumulator low pressure other than local observations made by an auxiliary operator and possibly dual valve indication in the main control room due to the valves cycling slightly.This potential scenario constitutes potential operation outside the design basis of the plant and was immediately reported to the NRC via the emergency notification system on January 9, 1998 at 1450 hours.CAUSE: The cause of this condition was inadequate design control during development of a plant modification implemented in August 1984, which was prior to issuance of the Harris Plant Operating License.Specifically, NRC Information Notice 82-25,"Failure of Hiller Actuators Upon Gradual Loss of Air Pressure" stated that on a gradual loss of control air, the pneumatic control valves may assume some intermediate position and cause the stored air in the accumulator to vent to atmosphere and prevent the actuator from performing it's safety function of closing.To resolve this concern Field Change Request (FCR-I-992)was developed to install two pressure switches in the Instrument Air System header that would de-energize the solenoid valves (described above)at 66 psig.However, ANSI Standard N18.2.a"Nuclear Safety Criteria for Design of Stationary PWR Plants" requires that a barrier be installed between safety class interfaces, such as the air control circuit and the instrument air supply.This design requirement was not properly applied during the development of FCR-I-992; therefore, the safety related system did not meet the single failure design criteria.The investigation for this event also revealed additional missed opportunities to identify this design deficiency in the Instrument Air System.These included: (1)HNP's response development for NRC Generic Letter 88-14 in 1989, which specifically included an evaluation (Plant Change Request, PCR-4151)of the Instrument Air System and the Pre-heater Bypass Valve actuator design, (2)Adverse Condition Report¹91-314 initiated in June 1991, which identified a leak in the supply air regulator for 1AF-81 that significantly lowered accumulator pressure, and (3)development of PCR-6158 and it's associated evaluation (PCR-6066) in 1992, which implemented a modification to the actuator air circuitry to enhance the air intensifier pump and upgrade portions of the control air piping to safety-related. | | ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 singla-spaced typawrittan lines) (16) |
| In each of these cases, the"smart" leak scenario described in this LER was not identified nor considered credible due to the incorrect assumption that the system design met the single failure criteria.N M (4.)
| | On January 9, 1998, a condition was identified during operation that results in the plant being potentially outside it' design basis. Specifically, a potential failure mechanism exists where a leak in the non-safety Instrument Air System could result in the inoperabihty of the Steam Generator Pre-heater Bypass Isolation Valves. These valves are safety-related containment isolation valves that are required by plant procedures to automatically shut in 10 seconds or less upon receipt of a Main Feedwater Isolation Signal. These valves are positioned by a pneumatic piston-operated actuator which is supplied by the non-safety related Instrument Air System. They are designed to automatically close if control air supply is lost. However, a "smart" air leak has been postulated in the Instrument Air system that could possibly reduce the air inlet pressure to just low enough to affect proper operation of the actuator's 3-way and 4-way pilot valves and not be detected by Operations personnel. If this occurred, the pilot valves would shuttle, causing the accumulator pressure to bleed off, which would prevent the valves from closing as required. This potential scenario constitutes operation outside the design basis of the plant and was reported to the NRC via the emergency notification system on January 9, 1998 at 1450 hours. |
| NRC FORM 366A F)a6)LlCENSEE EVENT REPORT (LER)TEXT CONTINUATION US.NUCLEAR REGULATORY COMMISSION FACILITY NAME (I)Shearen Harris Nuclear Plant~Unit//1 TEXT P secre s~seir cere'rerf.
| | The cause of this condition was inadequate design control during development of a plant modification implemented in August 1984 in response to NRC Information Notice 82-25. The investigation for this event also revealed several other missed opportunities to identify this condition during subsequent plant modifications and/or related evaluations. |
| ese errrFse)re copes el h'RC&m NQl)17)DOCKET 50400 LER NUMBER IS)SEQUENTIAL REVISION NUMBER NUMBER 98-Ool.-00 PAGE)3)3 OF 4 SAFETY SIGNIFICANCE:
| | Immediate corrective actions included development of a Justification for Continued Operation (JCO), with the required evaluation and compensatory measures to ensure continued operability. The JCO evaluation determined that the isolation valves were operable dependent upon once per 12-hour monitoring of the discharge pressure on the inlet regulator for each valve actuator. Additional corrective actions will include a plant modification to resolve the design deficiency, a FSAR revision and training for appropriate Engineering personnel. |
| There were no actual consequences associated with this potential failure scenario.JCO 98-01 established the basis for continued operability of the pre-heater bypass valves.These valves are listed as containment isolation valves in PLP-106 (TS Equipment List and Core Operating Limits Report)and are required to automatically shut following receipt of a Main Feedwater Isolation Signal (MFIS).A MFIS is generated from the safety injection actuation logic as well as on high-high S/G level in any steam generator.
| | 9802170061 980209 PDR ADCICK 05000400 S PDR |
| As such, any accident analyzed in the FSAR that results in a safety injection or high-high S/G level may be impacted by degradation of the non-safety related air supply to these valves.The list of applicable accidents includes: Main Steamline Break (15.1.5), Feedwater Line Break (15.2.8), Steam Generator Tube Rupture (15.6.3), and Loss of Coolant (15.6.5).Two different failure scenarios must be considered.
| | |
| The first is the failure of the air supply to the valve actuators coincident with initiation of the accident.This is necessary because a non-safety related component cannot be credited in the mitigation of an accident, nor can the loss of a non-safety related component result in failure of a safety related component.
| | NRC FORM 366A US. NUCLEAR REGULATORY COMMISSION M-SS) |
| The second condition to be considered is the undetected loss of air to the actuators during normal plant operation such that a pre-existing, degraded con'dition could be created.Loss of Air Su I Coincident with Accident Initiation Of the events listed above, only the analysis of the Main Steamline Break (MSLB)actually takes credit for isolation of main feedwater flow in mitigating the accident.However, due to the rapid event sequence of this accident scenario, the pre-heater bypass valves would remain capable of performing their safety function as analyzed in the limiting MSLB events.Loss of Air Su I as a Pre-existin Condition Over the course of the plant life, degradation of the air supply line to the pre-heater bypass valves that could result in a worst case leak (i.e., a smart leak)must be considered as a possibility.
| | LICENSEE EVENT REPORT (LER) |
| This presents a concern in that if the valves were to become disabled due to a small drop in the Instrument Air System header pressure, the inoperable condition of the pre-heater bypass valves may not be detected in the main control room.A MSLB or LOCA that occurred with this pre-existing condition would not progress within the constraints of the current analysis as described above.Therefore, the compensatory actions identified in the corrective actions section below are necessary to ensure that the air supply to the pre-heater bypass valves remains intact.This action is not intended to be a manual response credited in the mitigation of an accident in place of a normal automatic function.Rather, the increased level of surveillance improves the confidence in the integrity of the Instrument Air System header and ensures high reliability of the pre-heater bypass valves.PREVIOUS SIMILAR EVENTS: There have been no previous events reported related to a newly identified failure scenario that would potentially render the S/G Pre-heater Bypass Valves inoperable.
| | TEXT CONTINUATlON FACIUTY NAME (I) DOCKET LER NUMBER (6) PAGE (3) |
| NRC FORM 36BA)4.9S)LlCENSEE EVENT REPORT (LEB)TEXT CONTINUATION US.NUCLEAR RECULATORY COMMISSION FACILITY NAME II)Shearon Harris Nuclear Plant~Unit//I TEXT Rl psdsd spssd ss ssqvi ssE ssss ddt')psl copes ol NRC ppfps 3BQI (17)OOCXET 50400 LER NUMBER IB)YEAR SEOUENTIAL REVISION NUMBER NUMBER 98-00'I-00 PACE I31 4 OF 4 CORRECTIVE ACTIONS COMPLETED:
| | SEOUENTIAL REVISION YEAR NUMBER NUMBER Shearon Harris NUclear Plant - Unit ¹1 50400 2 OF 4 98 - 001 - 00 TEXT Pr more spscers nod. vso oddMmsl copies of ABC Povv 3IFQI (11) |
| 1.Justification for Continued Operation (JCO¹98-01)was generated along with the associated engineering evaluation (ESR¹9800014)..This JCO/evaluation determined that the pre-heater bypass valves remained operable based upon a once per shift (12-hour)monitoring of the discharge pressure on the inlet regulator for each valve actuator to ensure the integrity of the non-safety related instrument air piping.2.The Daily Operations Surveillance Test procedures (OST-1021 and OST-1022)were revised to direct 12-hour monitoring of the discharge pressure on the inlet regulator for each valve actuator.CORRECTIVE ACTIONS PLANNED: 1.A plant modification will be developed and implemented to resolve the smart leak in the non-safety related Instrument Air System header.This will be completed prior to entering Mode<following completion of refueling outage¹8, which is currently scheduled to begin in October 1998.2.The FSAR will be revised to include a description of the modified S/G Pre-heater Bypass Isolation Valves.This will be completed by December 31, 1998.3.Training will be provided to appropriate Engineering personnel, which addresses the design control aspects of this event.This will be completed by July 15, 1998.}}
| | EVENT DESCRIPTION: |
| | On January 9, 1998, with the plant operating in mode 1 at 100% power, a condition was identified that results in the plant being potentially outside the design basis. Specifically, a potential failure mechanism exists where a leak in the non-safety Instrument Air System (EIIS Code: LD) could result in the inoperability of the Steam Generator (S/G) Pre-heater Bypass Isolation Valves (1AF-64, 1AF-102, and 1AF-81, EIIS Code: SJ-V). These valves are safety-related containment isolation valves that are required by plant procedure PLP-106 (Technical Specification Equipment List and Core Operating Limit Report) to automatically shut in 10 seconds or less upon receipt of a Main Feedwater Isolation Signal (MFIS). These valves are positioned by a pneumatic piston-operated actuator and are opened and closed by high pressure air (-150 psig) from the actuator's accumulator. The accumulator is maintained at a higher air pressure by an air intensifier pump. The intensifier is a double piston compressor that is driven by control air. The air intensifier is needed to boost inlet control air pressure to approximately 150 psig, since normal control air pressure (approx. 75 psig) will not close the valve alone. With the use of pneumatic pilot valves located on the actuator, the intensified air in the accumulator is used to position the actuator piston, which is connected to the valve disc by a common shaft. The pneumatic pilot valves guide the intensified air from the accumulator to either face of the piston to position the operator. Directing the high pressure air below the piston will open the S/G pre-heater bypass valves; and directing the high pressure air to the top of the piston will cause the valves to close. |
| | The S/G pre-heater bypass valves are designed to automatically close if control air supply is lost. This is accomplished by two in-series solenoid valves that are energized open to supply the control air from the Instrument Air System. Ifa Main Feedwater Isolation signal is generated the solenoid valves will de-energize, securing the control air supply, which causes the S/G pre-heater bypass isolation valves to close. The solenoid valves will also de-energize if Instrument Air System pressure drops to 66 psig. |
| | However, a "smart" air leak has been postulated in the Instrument Air System piping that could possibly reduce the air inlet pressure to just low enough to affect proper operation of the actuator's 3-way and 4-way pilot valves and not be detected by the pressure switches in the main header of the Instrument Air System that would de-energize solenoid valves at 66 psig. |
| | If this occurred, the pilot valves would shuttle, causing the accumulator pressure to bleed off,thewhich would prevent the valves from closing as required. Operations personnel would have no indication of accumulator low pressure other than local observations made by an auxiliary operator and possibly dual valve indication in the main control room due to the valves cycling slightly. |
| | This potential scenario constitutes potential operation outside the design basis of the plant and was immediately reported to the NRC via the emergency notification system on January 9, 1998 at 1450 hours. |
| | CAUSE: |
| | The cause of this condition was inadequate design control during development of a plant modification implemented in August 1984, which was prior to issuance of the Harris Plant Operating License. Specifically, NRC Information Notice 82-25, "Failure of Hiller Actuators Upon Gradual Loss of Air Pressure" stated that on a gradual loss of control air, the pneumatic control valves may assume some intermediate position and cause the stored air in the accumulator to vent to atmosphere and prevent the actuator from performing it's safety function of closing. To resolve this concern Field Change Request (FCR-I-992) was developed to install two pressure switches in the Instrument Air System header that would de-energize the solenoid valves (described above) at 66 psig. However, ANSI Standard N18.2.a "Nuclear Safety Criteria for Design of Stationary PWR Plants" requires that a barrier be installed between safety class interfaces, such as the air control circuit and the instrument air supply. This design requirement was not properly applied during the development of FCR-I-992; therefore, the safety related system did not meet the single failure design criteria. |
| | The investigation for this event also revealed additional missed opportunities to identify this design deficiency in the Instrument Air System. These included: (1) HNP's response development for NRC Generic Letter 88-14 in 1989, which specifically included an evaluation (Plant Change Request, PCR-4151) of the Instrument Air System and the Pre-heater Bypass Valve actuator design, (2) Adverse Condition Report ¹91-314 initiated in June 1991, which identified a leak in the supply air regulator for 1AF-81 that significantly lowered accumulator pressure, and (3) development of PCR-6158 and it's associated evaluation (PCR-6066) in 1992, which implemented a modification to the actuator air circuitry to enhance the air intensifier pump and upgrade portions of the control air piping to safety-related. In each of these cases, the "smart" leak scenario described in this LER was not identified nor considered credible due to the incorrect assumption that the system design met the single failure criteria. |
| | N M (4. ) |
| | |
| | NRC FORM 366A US. NUCLEAR REGULATORY COMMISSION F)a6) |
| | LlCENSEE EVENT REPORT (LER) |
| | TEXT CONTINUATION FACILITY NAME (I) DOCKET LER NUMBER IS) PAGE )3) |
| | SEQUENTIAL REVISION NUMBER NUMBER Shearen Harris Nuclear Plant ~ |
| | Unit //1 50400 3 OF 4 98 - Ool .- 00 TEXT P secre s~seir cere'rerf. ese errrFse)re copes el h'RC &m NQl )17) |
| | SAFETY SIGNIFICANCE: |
| | There were no actual consequences associated with this potential failure scenario. JCO 98-01 established the basis for continued operability of the pre-heater bypass valves. These valves are listed as containment isolation valves in PLP-106 (TS Equipment List and Core Operating Limits Report) and are required to automatically shut following receipt of a Main Feedwater Isolation Signal (MFIS). A MFIS is generated from the safety injection actuation logic as well as on high-high S/G level in any steam generator. As such, any accident analyzed in the FSAR that results in a safety injection or high-high S/G level may be impacted by degradation of the non-safety related air supply to these valves. |
| | The list of applicable accidents includes: Main Steamline Break (15.1.5), Feedwater Line Break (15.2.8), Steam Generator Tube Rupture (15.6.3), and Loss of Coolant (15.6.5). Two different failure scenarios must be considered. |
| | The first is the failure of the air supply to the valve actuators coincident with initiation of the accident. This is necessary because a non-safety related component cannot be credited in the mitigation of an accident, nor can the loss of a non-safety related component result in failure of a safety related component. The second condition to be considered is the undetected loss of air to the actuators during normal plant operation such that a pre-existing, degraded con'dition could be created. |
| | Loss of Air Su I Coincident with Accident Initiation Of the events listed above, only the analysis of the Main Steamline Break (MSLB) actually takes credit for isolation of main feedwater flow in mitigating the accident. However, due to the rapid event sequence of this accident scenario, the pre-heater bypass valves would remain capable of performing their safety function as analyzed in the limiting MSLB events. |
| | Loss of Air Su I as a Pre-existin Condition Over the course of the plant life, degradation of the air supply line to the pre-heater bypass valves that could result in a worst case leak (i.e., a smart leak) must be considered as a possibility. This presents a concern in that if the valves were to become disabled due to a small drop in the Instrument Air System header pressure, the inoperable condition of the pre-heater bypass valves may not be detected in the main control room. A MSLB or LOCA that occurred with this pre-existing condition would not progress within the constraints of the current analysis as described above. |
| | Therefore, the compensatory actions identified in the corrective actions section below are necessary to ensure that the air supply to the pre-heater bypass valves remains intact. This action is not intended to be a manual response credited in the mitigation of an accident in place of a normal automatic function. Rather, the increased level of surveillance improves the confidence in the integrity of the Instrument Air System header and ensures high reliability of the pre-heater bypass valves. |
| | PREVIOUS SIMILAREVENTS: |
| | There have been no previous events reported related to a newly identified failure scenario that would potentially render the S/G Pre-heater Bypass Valves inoperable. |
| | |
| | NRC FORM 36BA US. NUCLEAR RECULATORY COMMISSION |
| | )4.9S) |
| | LlCENSEE EVENT REPORT (LEB) |
| | TEXT CONTINUATION FACILITY NAME II) OOCXET LER NUMBER IB) PACE I31 SEOUENTIAL REVISION YEAR NUMBER NUMBER Shearon Harris Nuclear Plant ~ Unit //I 50400 4 OF 4 98 - 00'I - 00 TEXT Rl psdsd spssd ss ssqvi ssE ssss ddt')psl copes ol NRC ppfps 3BQI (17) |
| | CORRECTIVE ACTIONS COMPLETED: |
| | : 1. Justification for Continued Operation (JCO ¹98-01) was generated along with the associated engineering evaluation (ESR ¹9800014)..This JCO/evaluation determined that the pre-heater bypass valves remained operable based upon a once per shift (12-hour) monitoring of the discharge pressure on the inlet regulator for each valve actuator to ensure the integrity of the non-safety related instrument air piping. |
| | : 2. The Daily Operations Surveillance Test procedures (OST-1021 and OST-1022) were revised to direct 12-hour monitoring of the discharge pressure on the inlet regulator for each valve actuator. |
| | CORRECTIVE ACTIONS PLANNED: |
| | : 1. A plant modification will be developed and implemented to resolve the smart leak in the non-safety related Instrument Air System header. This will be completed prior to entering Mode< following completion of refueling outage ¹8, which is currently scheduled to begin in October 1998. |
| | : 2. The FSAR will be revised to include a description of the modified S/G Pre-heater Bypass Isolation Valves. |
| | This will be completed by December 31, 1998. |
| | : 3. Training will be provided to appropriate Engineering personnel, which addresses the design control aspects of this event. This will be completed by July 15, 1998.}} |
LER 98-001-00:on 980109,potential Condition Outside Design Basis Related to Instrument Air Sys Leak Causing SG pre- Heater Bypass Isolation Valves to Be Inoperable Was Noted. Caused by Inadequate Design Control.Generated Jco 98-01ML18016A313 |
Person / Time |
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Site: |
Harris ![Duke Energy icon.png](/w/images/7/75/Duke_Energy_icon.png) |
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Issue date: |
02/09/1998 |
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From: |
Verrilli M CAROLINA POWER & LIGHT CO. |
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To: |
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Shared Package |
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ML18016A311 |
List: |
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References |
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LER-98-001, LER-98-1, NUDOCS 9802170061 |
Download: ML18016A313 (4) |
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Category:LICENSEE EVENT REPORT (SEE ALSO AO
MONTHYEARML18017A9181999-10-0808 October 1999 LER 99-008-00:on 991008,CR Emergency Filtration Sys Tech Specs Occurred.Caused by Site Personnel Failed to Recognize That Blocking Open CR Emergency Filtration Sys.Procedures Revised.With 991008 Ltr ML18017A8671999-09-10010 September 1999 LER 99-007-00:on 990811,determined That Cvis ARMs High Alarm Setpoints Were Not within TS Limit.Caused by Not Having Procedure to Verify If Cvis ARM High Alarm Setpoints Were within TS Requirements.Revised Procedures.With 990910 Ltr ML18016B0481999-08-0404 August 1999 LER 99-006-01:on 981124,noted Failure to Comply with TS 4.0.4 & TS 3/4.6.3, Civs. Caused by post-maint Testing That Did Not Adequately Test Control Circuitry & Verify Isolation Time Following Maint.Procedure Was Revised ML18016A9801999-06-0404 June 1999 LER 99-006-00:on 981124,failed to Comply with TS 4.0.4 & TS 3/4.6.3, Civ. Caused by post-maint Testing That Did Not Adequately Test Control Circuitry & Verify Isolation Time Following Maint.Procedure Will Be Revised.With 990604 Ltr ML18016A9111999-04-12012 April 1999 LER 99-005-00:on 990313,plant Exceeded ESFAS TS 3.3.2,Action 21.Caused by Inadequate Procedure Rev Preparation.Licensee Revised Applicable Maint Surveillance Test Procedure (MST-10072) to Identify TS Required Actions.With 990412 Ltr ML18016A8971999-04-0808 April 1999 LER 99-004-00:on 990312,unit Trip Was Noted.Caused by Degraded Condition of SG Water Level Flow Control Valve. Replaced Positioners on All Three FW Regulating Valves.With 990408 Ltr ML18016A8261999-02-22022 February 1999 LER 99-003-00:on 990123,noted That Plant Was Outside Design Basis Due to Isolation of Fire Protection Containment Sprinkler Sys.Caused by Human Error.Restored Containment Sprinkler Sys to Operable Status.With 990222 Ltr ML18016A8111999-02-12012 February 1999 LER 99-002-00:on 990114,RT Due to Not Removing Temporary Device from Relay Following Calibration Was Noted.Caused by Human Error.Counseled Personnel Involved in Event.With 990212 Ltr ML18016A7971999-02-0505 February 1999 LER 99-001-00:on 990106,SF Pool Water Level Was Not Maintained Greater than 23 Feet Above Stored BWR Fuel Assemblies.Caused by Fasteners Bending Under Specific Circumstances.Increased Water Level.With 990205 Ltr ML18016A7941999-01-29029 January 1999 LER 98-004-01:on 980313,identified Design Deficiency Re Potential Runout of Tdafwp.Caused by Inadequate Original AFW Sys Design.Operability Evaluation Was Completed on 980313 & Addl Engineering Analysis Was Performed by Vendor ML18016A7211998-11-17017 November 1998 LER 98-007-00:on 981023,turbine Control Anomaly Caused Manual Rt.Caused by Failure to Incorporate Verbal Vendor Guidance in Operating Procedures.Addl Vendor Guidance Will Be Verified & Added to Procedures.With 981117 Ltr ML18016A4841998-07-0707 July 1998 LER 97-002-01:on 970207,determined That Cold Weather Conditions Resulted in Mfiv Being Potentially Inoperable During Period 970117-20.Caused by Inadequate Design of HVAC Sys.Implemented Mods to Steam Tunnel HVAC Sys ML18016A4701998-06-30030 June 1998 LER 97-021-03:on 980210,discovered That SFP Water Level Had Not Been Verified Greater than 23 Feet Above BWR Fuel Assemblies.Caused by Misinterpretation of TS Requirements. Will Submit TS Change Request to Revise TS 3.9.1.11 ML18016A4491998-06-0808 June 1998 LER 98-006-00:on 980508,failure to Perform Insp & Preventive Maint on MCCB as Required by TS Was Noted.Caused by Inadequate Sps.Tested 9 Pressurizer Heater Bank Breakers by Cycling each.W/980608 Ltr ML18022B0551998-05-20020 May 1998 LER 98-005-00:on 980420,TS Verbatim non-compliance Was Determined.Caused by Misinterpretation of TS Requirements. Issued Memo to Reemphasize Need to Comply W/Literal Meaning of TS Requirements in Verbatim manner.W/980520 Ltr ML18016A4061998-04-30030 April 1998 LER 98-002-01:on 980121,determined Ssps (P-11 Permissive) Testing Deficiency.Caused by Inadequate Review of Initial Ts.Will Revise & Perform Surveillance Test Procedures to Verify Operability of P-11 Permissive ML18016A3841998-04-13013 April 1998 LER 98-004-00:on 980313,design Deficiency Related to Indequate Runout Protection for Turbine Driven AFW Pump Was Identified.Caused by Inadequate Original AFW Sys Design. Evaluation (ESR 98-00100) Will Be completed.W/980409 Ltr ML18016A3441998-03-12012 March 1998 LER 97-021-02:on 980210,identified Failure to Properly Test non-safety Related Pressurizer Porv.Caused by Inadequate Surveillance Test Procedures.Revised Operations Surveillance Test OST-1117 to Include Testing of Subject PORV ML18016A3291998-02-27027 February 1998 LER 98-003-00:on 980129,failure to Perform Shutdown Margin Calculation Required by TS Surveillance Requirements Occurred.Caused by Ambiguity in TS 3.1.3.1.c.Procedures revised.W/980227 Ltr ML18016A3211998-02-20020 February 1998 LER 98-002-00:on 980121,solid State Protection Sys Testing Deficiency Occurred.Caused by Inadequate Review of Initial Tech Specs.Ts Testing Frequency for P-11 Permissive Revised. W/980217 Ltr ML18016A3131998-02-0909 February 1998 LER 98-001-00:on 980109,potential Condition Outside Design Basis Related to Instrument Air Sys Leak Causing SG pre- Heater Bypass Isolation Valves to Be Inoperable Was Noted. Caused by Inadequate Design Control.Generated Jco 98-01 ML18016A2641997-12-18018 December 1997 LER 97-024-00:on 971118,SSPS Testing Deficiency Was Noted. Caused by Inadequate Testing Scheme Provided by Ssps Vendor. Revised procedure.W/971218 Ltr ML18016A2501997-11-24024 November 1997 LER 97-023-00:on 920721,RCS PIV Testing Deficiency Was Noted.Caused by Failure to Consider All Testing Variables During Initial Sp Development.Surveillance Tp OST-1506 Was Revised to Incorporate Correction factor.W/971124 Ltr ML18016A2201997-10-22022 October 1997 LER 97-021-01:on 970922,discovered That Spent Fuel Pool Water Level Was Not Maintained Greater than 23 Feet Above Stored BWR Fuel Assemblies.Caused by Misinterpretation of TS Requirements.Revised Daily Surveillance Procedures ML18016A2081997-10-14014 October 1997 LER 97-016-01:on 970608,reactor Trip Occurred,Due to Personnel Error While Attempting to Adjust Power Range Nuclear Instrumentation Channel Following Performance of Calorimetric.Procedures revised.W/971014 Ltr ML18016A2111997-10-14014 October 1997 LER 96-008-02:on 960425,turbine Trip/Reactor Trip Occurred. Caused by High Resistance Connection Resulting from a Phase Switch Jaw & Blade Contacts.Failed a Phase Disconnect Switch on Breaker 52-7 Replaced ML18016A1931997-09-29029 September 1997 LER 97-022-00:on 970829,TS Required Shutdown Due to Expiration of AFW Lco.Caused by Personnel Error.Completed Repairs TDAFW Pump & Returned Plant to Svc on 970831. W/970926 Ltr ML18016A1891997-09-12012 September 1997 LER 97-020-00:on 970814,inadequate Fire Protection Provided for safety-related EDG Fuel Oil Transfer Pump Cables Resulted in Operation Outside Design Basis.Caused by Engineering Oversight.Established Fire watches.W/970912 Ltr ML18016A1881997-09-12012 September 1997 LER 97-021-00:on 970814,spent Fuel Pool Water Level Was Not Maintained Greater than 23 Feet Above Stored BWR Fuel Assemblies.Caused by Misinterpretation of Ts.Directions Provided to Operations.W/970912 Ltr ML18012A8641997-08-18018 August 1997 LER 97-019-00:on 970720,turbine Trip/Reactor Trip Occurred. Caused by Three Phase Fault That Collapsed Excitation Field in Main Generator,Resulting in Generator Lockout.Exciter Rotor Assembly Was replaced.W/970818 Ltr ML18012A8581997-08-0808 August 1997 LER 96-018-01:on 960903,manual Reactor Trip Occurred Due to Loss of Normal Sw.Caused by Mechanical Failure of B Water Pump & a Normal SW to Remain Running Once Manually Started.Restored a Normal SW Pump to Svc ML18012A8551997-08-0808 August 1997 LER 96-013-02:on 961028,condition Outside of Design Basis Where RWST Had Been Aligned w/non-seismically Qualified Sys Was Identified.Caused by Failure to Reconcile Operating Procedure Lineups.Established Administrative Controls ML18012A8471997-07-31031 July 1997 LER 97-018-00:on 970701,determined That Plant Procedures Had Not Received Proper Reviews & Approvals.Caused by Failure to Comply W/Plant Administrative Procedure AP-006.Counseled Involved individuals.W/970731 Ltr ML18012A8371997-07-24024 July 1997 LER 97-S01-00:on 970405,unescorted Access Inappropriately Granted to Contract Outage Workers Was Determined.Caused by Personnel Error.Access Files for Individuals Inappropriately Granted Unescorted Access Were Placed on Access Hold ML18012A8291997-07-11011 July 1997 LER 97-017-00:on 970612,failed to Recognize Inoperable Reactor Afd Monitor.Caused by Personnel Error.Operators Involved in Event Will Be Counseled Prior to Assuming Shift duties.W/970711 Ltr ML18012A8301997-07-0808 July 1997 LER 97-016-00:on 970608,reactor Trip Occurred Due to Personnel Error in Adjusting Power Range (Pr) Nuclear Instrumentation (Ni).Issued Night Order Prohibiting Pr Ni Adjustment When Redundant Channel inoperable.W/970708 Ltr ML18012A8241997-07-0202 July 1997 LER 97-015-00:on 970602,inadequate Auxiliary Feedwater Sys Flow Control Valve Surveillance Testing Deficiency Was Identified.Caused by Failure to Recognize Impact on TS 4.7.1.2.1.Readjusted AFW FCV Actuator spring.W/970702 Ltr ML18022B0181997-06-13013 June 1997 LER 97-014-00:on 970514,SI Occurred During Ssps Surveillance Testing.Caused by Inattention to Detail During Recent Rev to Surveillance Test Procedure Being Used.Revised Deficient Surveillance procedures.W/970613 Ltr ML18012A8081997-06-0909 June 1997 LER 97-013-00:on 970508,entry Into Mode-6 Without Operable Components,Resulting in TS 3.0.4 Violation Occurred.Caused by Personnel Error.Personnel Involved counseled.W/970609 Ltr ML18012A8021997-06-0606 June 1997 LER 97-023-02:on 961114,design Deficiency Was Identified in Emergency DG Protection Circuitry.Caused by Inadequate Plant Design.Revised Surveillance Test Procedures OST-1013 & OST-1073.W/970606 Ltr ML18012A8011997-06-0404 June 1997 LER 97-012-00:on 970505,determined That Previous Auxiliary Control Panel Had Not Verified Operability of Interposing Relays.Caused by Misinterpretation of Tss.Reviewed Other Remote Shutdown Panel Transfer circuitry.W/970604 Ltr ML18012A7951997-05-29029 May 1997 LER 96-023-01:on 961114,design Deficiency in EDG Protection Circuitry Was Identified.Caused by Inadequate Original Plant Design.Surveillance Test Procedures OST-1013 & OST-1073 revised.W/970529 Ltr ML18012A7891997-05-22022 May 1997 LER 97-011-00:on 970422,inappropriate TS Interpretation Resulted in Violations of ECCS Accumulator TS & Entry Into TS 3.0.3.Caused by Procedural Inadequacy.Tsi 88-001 Cancelled 970508 & Procedures revised.W/970522 Ltr ML18012A7871997-05-19019 May 1997 LER 97-010-00:on 970418,design Deficiency Determined Re Reactor Coolant Pump Motor Oil Collection Sys.Caused by RCP Ocs Design Detail.Rcp Ocs Enclosures for Each of Three Installed RCP Motors Have Been modified.W/970519 Ltr ML18012A7761997-05-0707 May 1997 LER 97-009-00:on 970407,fuse Was Removed from CR Ventilation Isolation Signal Power Supply Circuitry Due to Personnel Error.Individuals Involved Were counseled.W/970507 Ltr ML18012A7751997-05-0505 May 1997 LER 97-008-00:on 970404,safety-related AHU Not Declared Inoperable During Maintenance on Associated Temperature Switches Resulting in Violation of Ts.Caused by Incorrect Interpretation.Operations Night Order issued.W/970505 Ltr ML18012A6291997-04-24024 April 1997 LER 97-007-00:on 970325,inoperable CCW Sys TS 3.0.3 Entry Made.Caused by Combination of Procedural Inadequacies, Improper Use of Procedure Guidance & Poor Communication. Applicable Individuals counseled.W/970423 Ltr ML18022B0151997-04-17017 April 1997 LER 97-006-00:on 970318,breach Was Identified in Thermo-Lag Fire Barrier Wall Due to Inadequate Initial Design,Poor Construction Methods & Incomplete as-built Design.Visual Insp of Thermo-Lag Barrier Walls performed.W/970417 Ltr ML18012A6041997-04-0303 April 1997 LER 97-004-00:on 970304,in-plant Spent Fuel Cask Handling Activities Conducted Outside Design Basis.Caused by Lack of Understanding of Requirements.Operations Placed on Hold Pending NRC Review & Approval of procedures.W/970331 Ltr ML18012A6031997-03-31031 March 1997 LER 97-003-00:on 970227,steam Generator Low Level Protection Circuitry Outside Design Basis Occurred.Caused by Inadequate Failure Modes & Effects Analysis Performed as-built Piping Configuration for S/G Level.Review performed.W/970331 Ltr 1999-09-10
[Table view] Category:RO)
MONTHYEARML18017A9181999-10-0808 October 1999 LER 99-008-00:on 991008,CR Emergency Filtration Sys Tech Specs Occurred.Caused by Site Personnel Failed to Recognize That Blocking Open CR Emergency Filtration Sys.Procedures Revised.With 991008 Ltr ML18017A8671999-09-10010 September 1999 LER 99-007-00:on 990811,determined That Cvis ARMs High Alarm Setpoints Were Not within TS Limit.Caused by Not Having Procedure to Verify If Cvis ARM High Alarm Setpoints Were within TS Requirements.Revised Procedures.With 990910 Ltr ML18016B0481999-08-0404 August 1999 LER 99-006-01:on 981124,noted Failure to Comply with TS 4.0.4 & TS 3/4.6.3, Civs. Caused by post-maint Testing That Did Not Adequately Test Control Circuitry & Verify Isolation Time Following Maint.Procedure Was Revised ML18016A9801999-06-0404 June 1999 LER 99-006-00:on 981124,failed to Comply with TS 4.0.4 & TS 3/4.6.3, Civ. Caused by post-maint Testing That Did Not Adequately Test Control Circuitry & Verify Isolation Time Following Maint.Procedure Will Be Revised.With 990604 Ltr ML18016A9111999-04-12012 April 1999 LER 99-005-00:on 990313,plant Exceeded ESFAS TS 3.3.2,Action 21.Caused by Inadequate Procedure Rev Preparation.Licensee Revised Applicable Maint Surveillance Test Procedure (MST-10072) to Identify TS Required Actions.With 990412 Ltr ML18016A8971999-04-0808 April 1999 LER 99-004-00:on 990312,unit Trip Was Noted.Caused by Degraded Condition of SG Water Level Flow Control Valve. Replaced Positioners on All Three FW Regulating Valves.With 990408 Ltr ML18016A8261999-02-22022 February 1999 LER 99-003-00:on 990123,noted That Plant Was Outside Design Basis Due to Isolation of Fire Protection Containment Sprinkler Sys.Caused by Human Error.Restored Containment Sprinkler Sys to Operable Status.With 990222 Ltr ML18016A8111999-02-12012 February 1999 LER 99-002-00:on 990114,RT Due to Not Removing Temporary Device from Relay Following Calibration Was Noted.Caused by Human Error.Counseled Personnel Involved in Event.With 990212 Ltr ML18016A7971999-02-0505 February 1999 LER 99-001-00:on 990106,SF Pool Water Level Was Not Maintained Greater than 23 Feet Above Stored BWR Fuel Assemblies.Caused by Fasteners Bending Under Specific Circumstances.Increased Water Level.With 990205 Ltr ML18016A7941999-01-29029 January 1999 LER 98-004-01:on 980313,identified Design Deficiency Re Potential Runout of Tdafwp.Caused by Inadequate Original AFW Sys Design.Operability Evaluation Was Completed on 980313 & Addl Engineering Analysis Was Performed by Vendor ML18016A7211998-11-17017 November 1998 LER 98-007-00:on 981023,turbine Control Anomaly Caused Manual Rt.Caused by Failure to Incorporate Verbal Vendor Guidance in Operating Procedures.Addl Vendor Guidance Will Be Verified & Added to Procedures.With 981117 Ltr ML18016A4841998-07-0707 July 1998 LER 97-002-01:on 970207,determined That Cold Weather Conditions Resulted in Mfiv Being Potentially Inoperable During Period 970117-20.Caused by Inadequate Design of HVAC Sys.Implemented Mods to Steam Tunnel HVAC Sys ML18016A4701998-06-30030 June 1998 LER 97-021-03:on 980210,discovered That SFP Water Level Had Not Been Verified Greater than 23 Feet Above BWR Fuel Assemblies.Caused by Misinterpretation of TS Requirements. Will Submit TS Change Request to Revise TS 3.9.1.11 ML18016A4491998-06-0808 June 1998 LER 98-006-00:on 980508,failure to Perform Insp & Preventive Maint on MCCB as Required by TS Was Noted.Caused by Inadequate Sps.Tested 9 Pressurizer Heater Bank Breakers by Cycling each.W/980608 Ltr ML18022B0551998-05-20020 May 1998 LER 98-005-00:on 980420,TS Verbatim non-compliance Was Determined.Caused by Misinterpretation of TS Requirements. Issued Memo to Reemphasize Need to Comply W/Literal Meaning of TS Requirements in Verbatim manner.W/980520 Ltr ML18016A4061998-04-30030 April 1998 LER 98-002-01:on 980121,determined Ssps (P-11 Permissive) Testing Deficiency.Caused by Inadequate Review of Initial Ts.Will Revise & Perform Surveillance Test Procedures to Verify Operability of P-11 Permissive ML18016A3841998-04-13013 April 1998 LER 98-004-00:on 980313,design Deficiency Related to Indequate Runout Protection for Turbine Driven AFW Pump Was Identified.Caused by Inadequate Original AFW Sys Design. Evaluation (ESR 98-00100) Will Be completed.W/980409 Ltr ML18016A3441998-03-12012 March 1998 LER 97-021-02:on 980210,identified Failure to Properly Test non-safety Related Pressurizer Porv.Caused by Inadequate Surveillance Test Procedures.Revised Operations Surveillance Test OST-1117 to Include Testing of Subject PORV ML18016A3291998-02-27027 February 1998 LER 98-003-00:on 980129,failure to Perform Shutdown Margin Calculation Required by TS Surveillance Requirements Occurred.Caused by Ambiguity in TS 3.1.3.1.c.Procedures revised.W/980227 Ltr ML18016A3211998-02-20020 February 1998 LER 98-002-00:on 980121,solid State Protection Sys Testing Deficiency Occurred.Caused by Inadequate Review of Initial Tech Specs.Ts Testing Frequency for P-11 Permissive Revised. W/980217 Ltr ML18016A3131998-02-0909 February 1998 LER 98-001-00:on 980109,potential Condition Outside Design Basis Related to Instrument Air Sys Leak Causing SG pre- Heater Bypass Isolation Valves to Be Inoperable Was Noted. Caused by Inadequate Design Control.Generated Jco 98-01 ML18016A2641997-12-18018 December 1997 LER 97-024-00:on 971118,SSPS Testing Deficiency Was Noted. Caused by Inadequate Testing Scheme Provided by Ssps Vendor. Revised procedure.W/971218 Ltr ML18016A2501997-11-24024 November 1997 LER 97-023-00:on 920721,RCS PIV Testing Deficiency Was Noted.Caused by Failure to Consider All Testing Variables During Initial Sp Development.Surveillance Tp OST-1506 Was Revised to Incorporate Correction factor.W/971124 Ltr ML18016A2201997-10-22022 October 1997 LER 97-021-01:on 970922,discovered That Spent Fuel Pool Water Level Was Not Maintained Greater than 23 Feet Above Stored BWR Fuel Assemblies.Caused by Misinterpretation of TS Requirements.Revised Daily Surveillance Procedures ML18016A2081997-10-14014 October 1997 LER 97-016-01:on 970608,reactor Trip Occurred,Due to Personnel Error While Attempting to Adjust Power Range Nuclear Instrumentation Channel Following Performance of Calorimetric.Procedures revised.W/971014 Ltr ML18016A2111997-10-14014 October 1997 LER 96-008-02:on 960425,turbine Trip/Reactor Trip Occurred. Caused by High Resistance Connection Resulting from a Phase Switch Jaw & Blade Contacts.Failed a Phase Disconnect Switch on Breaker 52-7 Replaced ML18016A1931997-09-29029 September 1997 LER 97-022-00:on 970829,TS Required Shutdown Due to Expiration of AFW Lco.Caused by Personnel Error.Completed Repairs TDAFW Pump & Returned Plant to Svc on 970831. W/970926 Ltr ML18016A1891997-09-12012 September 1997 LER 97-020-00:on 970814,inadequate Fire Protection Provided for safety-related EDG Fuel Oil Transfer Pump Cables Resulted in Operation Outside Design Basis.Caused by Engineering Oversight.Established Fire watches.W/970912 Ltr ML18016A1881997-09-12012 September 1997 LER 97-021-00:on 970814,spent Fuel Pool Water Level Was Not Maintained Greater than 23 Feet Above Stored BWR Fuel Assemblies.Caused by Misinterpretation of Ts.Directions Provided to Operations.W/970912 Ltr ML18012A8641997-08-18018 August 1997 LER 97-019-00:on 970720,turbine Trip/Reactor Trip Occurred. Caused by Three Phase Fault That Collapsed Excitation Field in Main Generator,Resulting in Generator Lockout.Exciter Rotor Assembly Was replaced.W/970818 Ltr ML18012A8581997-08-0808 August 1997 LER 96-018-01:on 960903,manual Reactor Trip Occurred Due to Loss of Normal Sw.Caused by Mechanical Failure of B Water Pump & a Normal SW to Remain Running Once Manually Started.Restored a Normal SW Pump to Svc ML18012A8551997-08-0808 August 1997 LER 96-013-02:on 961028,condition Outside of Design Basis Where RWST Had Been Aligned w/non-seismically Qualified Sys Was Identified.Caused by Failure to Reconcile Operating Procedure Lineups.Established Administrative Controls ML18012A8471997-07-31031 July 1997 LER 97-018-00:on 970701,determined That Plant Procedures Had Not Received Proper Reviews & Approvals.Caused by Failure to Comply W/Plant Administrative Procedure AP-006.Counseled Involved individuals.W/970731 Ltr ML18012A8371997-07-24024 July 1997 LER 97-S01-00:on 970405,unescorted Access Inappropriately Granted to Contract Outage Workers Was Determined.Caused by Personnel Error.Access Files for Individuals Inappropriately Granted Unescorted Access Were Placed on Access Hold ML18012A8291997-07-11011 July 1997 LER 97-017-00:on 970612,failed to Recognize Inoperable Reactor Afd Monitor.Caused by Personnel Error.Operators Involved in Event Will Be Counseled Prior to Assuming Shift duties.W/970711 Ltr ML18012A8301997-07-0808 July 1997 LER 97-016-00:on 970608,reactor Trip Occurred Due to Personnel Error in Adjusting Power Range (Pr) Nuclear Instrumentation (Ni).Issued Night Order Prohibiting Pr Ni Adjustment When Redundant Channel inoperable.W/970708 Ltr ML18012A8241997-07-0202 July 1997 LER 97-015-00:on 970602,inadequate Auxiliary Feedwater Sys Flow Control Valve Surveillance Testing Deficiency Was Identified.Caused by Failure to Recognize Impact on TS 4.7.1.2.1.Readjusted AFW FCV Actuator spring.W/970702 Ltr ML18022B0181997-06-13013 June 1997 LER 97-014-00:on 970514,SI Occurred During Ssps Surveillance Testing.Caused by Inattention to Detail During Recent Rev to Surveillance Test Procedure Being Used.Revised Deficient Surveillance procedures.W/970613 Ltr ML18012A8081997-06-0909 June 1997 LER 97-013-00:on 970508,entry Into Mode-6 Without Operable Components,Resulting in TS 3.0.4 Violation Occurred.Caused by Personnel Error.Personnel Involved counseled.W/970609 Ltr ML18012A8021997-06-0606 June 1997 LER 97-023-02:on 961114,design Deficiency Was Identified in Emergency DG Protection Circuitry.Caused by Inadequate Plant Design.Revised Surveillance Test Procedures OST-1013 & OST-1073.W/970606 Ltr ML18012A8011997-06-0404 June 1997 LER 97-012-00:on 970505,determined That Previous Auxiliary Control Panel Had Not Verified Operability of Interposing Relays.Caused by Misinterpretation of Tss.Reviewed Other Remote Shutdown Panel Transfer circuitry.W/970604 Ltr ML18012A7951997-05-29029 May 1997 LER 96-023-01:on 961114,design Deficiency in EDG Protection Circuitry Was Identified.Caused by Inadequate Original Plant Design.Surveillance Test Procedures OST-1013 & OST-1073 revised.W/970529 Ltr ML18012A7891997-05-22022 May 1997 LER 97-011-00:on 970422,inappropriate TS Interpretation Resulted in Violations of ECCS Accumulator TS & Entry Into TS 3.0.3.Caused by Procedural Inadequacy.Tsi 88-001 Cancelled 970508 & Procedures revised.W/970522 Ltr ML18012A7871997-05-19019 May 1997 LER 97-010-00:on 970418,design Deficiency Determined Re Reactor Coolant Pump Motor Oil Collection Sys.Caused by RCP Ocs Design Detail.Rcp Ocs Enclosures for Each of Three Installed RCP Motors Have Been modified.W/970519 Ltr ML18012A7761997-05-0707 May 1997 LER 97-009-00:on 970407,fuse Was Removed from CR Ventilation Isolation Signal Power Supply Circuitry Due to Personnel Error.Individuals Involved Were counseled.W/970507 Ltr ML18012A7751997-05-0505 May 1997 LER 97-008-00:on 970404,safety-related AHU Not Declared Inoperable During Maintenance on Associated Temperature Switches Resulting in Violation of Ts.Caused by Incorrect Interpretation.Operations Night Order issued.W/970505 Ltr ML18012A6291997-04-24024 April 1997 LER 97-007-00:on 970325,inoperable CCW Sys TS 3.0.3 Entry Made.Caused by Combination of Procedural Inadequacies, Improper Use of Procedure Guidance & Poor Communication. Applicable Individuals counseled.W/970423 Ltr ML18022B0151997-04-17017 April 1997 LER 97-006-00:on 970318,breach Was Identified in Thermo-Lag Fire Barrier Wall Due to Inadequate Initial Design,Poor Construction Methods & Incomplete as-built Design.Visual Insp of Thermo-Lag Barrier Walls performed.W/970417 Ltr ML18012A6041997-04-0303 April 1997 LER 97-004-00:on 970304,in-plant Spent Fuel Cask Handling Activities Conducted Outside Design Basis.Caused by Lack of Understanding of Requirements.Operations Placed on Hold Pending NRC Review & Approval of procedures.W/970331 Ltr ML18012A6031997-03-31031 March 1997 LER 97-003-00:on 970227,steam Generator Low Level Protection Circuitry Outside Design Basis Occurred.Caused by Inadequate Failure Modes & Effects Analysis Performed as-built Piping Configuration for S/G Level.Review performed.W/970331 Ltr 1999-09-10
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML18017A9181999-10-0808 October 1999 LER 99-008-00:on 991008,CR Emergency Filtration Sys Tech Specs Occurred.Caused by Site Personnel Failed to Recognize That Blocking Open CR Emergency Filtration Sys.Procedures Revised.With 991008 Ltr ML18017A9151999-09-30030 September 1999 Monthly Operating Rept for Sept 1999 for Shearon Harris Npp. with 991012 Ltr ML20217A9931999-09-30030 September 1999 NRC Regulatory Assessment & Oversight Pilot Program, Performance Indicator Data ML18017A8671999-09-10010 September 1999 LER 99-007-00:on 990811,determined That Cvis ARMs High Alarm Setpoints Were Not within TS Limit.Caused by Not Having Procedure to Verify If Cvis ARM High Alarm Setpoints Were within TS Requirements.Revised Procedures.With 990910 Ltr ML18017A8621999-08-31031 August 1999 Monthly Operating Rept for Aug 1999 for Harris Nuclear Plant.With 990908 Ltr ML18016B0481999-08-0404 August 1999 LER 99-006-01:on 981124,noted Failure to Comply with TS 4.0.4 & TS 3/4.6.3, Civs. Caused by post-maint Testing That Did Not Adequately Test Control Circuitry & Verify Isolation Time Following Maint.Procedure Was Revised ML18017A8361999-07-31031 July 1999 Monthly Operating Rept for July 1999 for Shearon Harris Nuclear Power Plant.With 990811 Ltr ML18016B0151999-06-30030 June 1999 Monthly Operating Rept for June 1999 for Shearon Harris Npp. with 990713 Ltr ML20196H8621999-06-30030 June 1999 NRC Regulatory Assessment & Oversight Pilot Program, Performance Indicator Data, June 1999 Rept ML18016A9801999-06-0404 June 1999 LER 99-006-00:on 981124,failed to Comply with TS 4.0.4 & TS 3/4.6.3, Civ. Caused by post-maint Testing That Did Not Adequately Test Control Circuitry & Verify Isolation Time Following Maint.Procedure Will Be Revised.With 990604 Ltr ML18016A9851999-05-31031 May 1999 Monthly Operating Rept for May 1999 for Shearon Harris Nuclear Plant,Unit 1.With 990614 Ltr ML18017A8981999-05-12012 May 1999 Technical Rept Entitled, Harris Nuclear Plant-Bacteria Detection in Water from C&D Spent Fuel Pool Cooling Lines. ML18016A9581999-04-30030 April 1999 Monthly Operating Rept for Apr 1999 for Shearon Harris Nuclear Plant,Unit 1.With 990513 Ltr ML18016A9011999-04-12012 April 1999 Part 21 Rept Re Defect in Component of DSRV-16-4,Enterprise DG Sys.Caused by Potential Problem with Connecting Rod Assemblies Built Since 1986,that Have Been Converted to Use Prestressed Fasteners.Affected Rods Should Be Inspected ML18016A9111999-04-12012 April 1999 LER 99-005-00:on 990313,plant Exceeded ESFAS TS 3.3.2,Action 21.Caused by Inadequate Procedure Rev Preparation.Licensee Revised Applicable Maint Surveillance Test Procedure (MST-10072) to Identify TS Required Actions.With 990412 Ltr ML18016A8971999-04-0808 April 1999 LER 99-004-00:on 990312,unit Trip Was Noted.Caused by Degraded Condition of SG Water Level Flow Control Valve. Replaced Positioners on All Three FW Regulating Valves.With 990408 Ltr ML18016A8941999-04-0505 April 1999 Revised Pages 20-25 to App 4A of non-proprietary Version of Rev 3 to HI-971760 ML18016A9101999-03-31031 March 1999 Monthly Operating Rept for Mar 1999 for Shearon Harris Nuclear Power Plant.With 990413 Ltr ML18016A8661999-03-31031 March 1999 Shnpp Operator Training Simulator,Simulator Certification Quadrennial Rept. ML18017A8931999-02-28028 February 1999 Risks & Alternative Options Associated with Spent Fuel Storage at Shearon Harris Nuclear Power Plant. ML18016A8551999-02-28028 February 1999 Monthly Operating Rept for Feb 1999 for Shearon Harris Npp. with 990312 Ltr ML18016A8261999-02-22022 February 1999 LER 99-003-00:on 990123,noted That Plant Was Outside Design Basis Due to Isolation of Fire Protection Containment Sprinkler Sys.Caused by Human Error.Restored Containment Sprinkler Sys to Operable Status.With 990222 Ltr ML18016A8531999-02-18018 February 1999 Non-proprietary Rev 3 to HI-971760, Licensing Rept for Expanding Storage Capacity in Harris SFP 'C' & 'D'. ML18016A8111999-02-12012 February 1999 LER 99-002-00:on 990114,RT Due to Not Removing Temporary Device from Relay Following Calibration Was Noted.Caused by Human Error.Counseled Personnel Involved in Event.With 990212 Ltr ML18016A7971999-02-0505 February 1999 LER 99-001-00:on 990106,SF Pool Water Level Was Not Maintained Greater than 23 Feet Above Stored BWR Fuel Assemblies.Caused by Fasteners Bending Under Specific Circumstances.Increased Water Level.With 990205 Ltr ML18022B0631999-02-0404 February 1999 Rev 0 to Nuclear NDE Manual. with 28 Oversize Uncodable Drawings of Alternative Plan Scope & 4 Oversize Codable Drawings ML20202J1161999-02-0101 February 1999 SER Accepting Relief Requests Associated with Second 10-year Interval Inservice Testing Program ML18016A8041999-01-31031 January 1999 Monthly Operating Rept for Jan 1999 for Shearon Harris Nuclear Power Plant.With 990211 Ltr ML18016A7941999-01-29029 January 1999 LER 98-004-01:on 980313,identified Design Deficiency Re Potential Runout of Tdafwp.Caused by Inadequate Original AFW Sys Design.Operability Evaluation Was Completed on 980313 & Addl Engineering Analysis Was Performed by Vendor ML18016A7801998-12-31031 December 1998 Monthly Operating Rept for Dec 1998 for Shearon Harris Npp. with 990113 Ltr ML18016A7671998-11-30030 November 1998 Monthly Operating Rept for Nov 1998 for Shnpp,Unit 1.With 981215 Ltr ML18016A9731998-11-28028 November 1998 Changes,Tests & Experiments, for Harris Nuclear Plant.Rept Provides Brief Description of Changes to Facility & Summary & of SE for Each Item That Was Implemented Under 10CFR50.59 Between 970608-981128.With 990527 Ltr ML18016A8351998-11-28028 November 1998 ISI Summary 8th Refueling Outage for Shearon Harris Power Plant,Unit 1. ML18016A7411998-11-25025 November 1998 Rev 1 to Shnpp Cycle 9 Colr. ML18016A7211998-11-17017 November 1998 LER 98-007-00:on 981023,turbine Control Anomaly Caused Manual Rt.Caused by Failure to Incorporate Verbal Vendor Guidance in Operating Procedures.Addl Vendor Guidance Will Be Verified & Added to Procedures.With 981117 Ltr ML18016A7071998-11-0303 November 1998 Rev 0 to Harris Unit 1 Cycle 9 Colr. ML18016A7201998-10-31031 October 1998 Monthly Operating Rept for Oct 1998 for Shearon Harris Nuclear Power Plant.With 981113 Ltr ML20154F8701998-10-0606 October 1998 Safety Evaluation Authorizing Proposed Alternative to Requirements of OMa-1988,Part 10,Section 4.2.2.3 for 21 Category a Reactor Coolant Sys Pressure Isolation Valves ML18016A6201998-09-30030 September 1998 Monthly Operating Rept for Sept 1998 for Harris Nuclear Power Plant.With 981012 Ltr ML18016A5971998-09-21021 September 1998 Rev 1 to Harris Unit 1 Cycle 8 Colr. ML18016A5881998-08-31031 August 1998 Monthly Operating Rept for Aug 1998 for Shnpp,Unit 1.With 980914 Ltr ML18016A5071998-07-31031 July 1998 Monthly Operating Rept for Jul 1998 for Shearon Harris Nuclear Plant.W/980811 Ltr ML18016A9431998-07-0707 July 1998 Rev 1 to QAP Manual. ML18016A4841998-07-0707 July 1998 LER 97-002-01:on 970207,determined That Cold Weather Conditions Resulted in Mfiv Being Potentially Inoperable During Period 970117-20.Caused by Inadequate Design of HVAC Sys.Implemented Mods to Steam Tunnel HVAC Sys ML18016A9371998-06-30030 June 1998 Technical Rept on Matl Identification of Spent Fuel Piping Welds at Hnp. ML18016A4861998-06-30030 June 1998 Monthly Operating Rept for June 1998 for SHNPP.W/980715 Ltr ML18016A4701998-06-30030 June 1998 LER 97-021-03:on 980210,discovered That SFP Water Level Had Not Been Verified Greater than 23 Feet Above BWR Fuel Assemblies.Caused by Misinterpretation of TS Requirements. Will Submit TS Change Request to Revise TS 3.9.1.11 ML18016A4491998-06-0808 June 1998 LER 98-006-00:on 980508,failure to Perform Insp & Preventive Maint on MCCB as Required by TS Was Noted.Caused by Inadequate Sps.Tested 9 Pressurizer Heater Bank Breakers by Cycling each.W/980608 Ltr ML18016A4521998-05-31031 May 1998 Monthly Operating Rept for May 1998 for Shearon Harris Nuclear Power Plant.W/980612 Ltr ML18016A7711998-05-26026 May 1998 Non-proprietary Rev 2 to HI-971760, Licensing Rept for Expanding Storage Capacity in Harris Spent Fuel Pools 'C' & 'D'. 1999-09-30
[Table view] |
Text
NRC FORM 366 U.s. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB No. 3150-0104 (4.95)
EXPIRES 04/30/96
' ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS MANDATORY INFORMATION COLLECTION REDUEST. 509 HRS. REPORTED LESSONS LEARNED ARE LXCENSEE EVENT REPORT (LER) INCORPORATED INTO THE LICENSING PROCESS AND FEO BACK TO UIOUSTRY.
FORWARD COMMENTS REGAROIHG BURDEN ESTIMATE TO THE INFORMATION AND (Sae reverse for required number of RECORDS MANAGEMENT BRANCH Irk F33L US. NUCLEAR REGULATORY COMMLSSION, WASHINGTON, OC 20555000l, ANO TO THE PAPERV/ORK REDUCTION PRDIECT ISI50.
digits/characters for each block) OIOIL OFFICE OF MANAGEMENT ANO BUDGET, WASHUIGTON, OC 205IL FACIUTY NAME lll DOCKET NUMBER I2) PAGE I3I Harris Nuclear Plant Unit-1 50-400 1 OF4 TITLE (4I Potential Condition Outside Design Basis related to Instrument Air System Leak causing the S/G Pre-heater Bypass Isolation Valves to be inoperable.
EVENT DATE (5) LER NUMBER {6) REPORT DATE {7) OTHER FACILITIES INVOLVED (6)
FACIUTY NAME DOCKET NUMBER MONTH DAY SEOUENTIAL REVISION MONTH DAY YEAR NUMBER NUMBER FACIUTY NAME 1 9 98 98 001 00 2 9 98 DOCKET NUMBER 05000 OPERATING THIS REPORT IS SUBMITTED PURSUANT To THE REQUIREMENTS OF 10 CFR 5: {Chock one or more) (11)
MODE (9) 20.2201 {b) 20.2203(a)(2)(v) 50.73(a) {2)(i) 50.73(a)(2)(viii)
POWER 20.2203(a) (1) 20.2203(a) {3){i) 50.73(a) (2) {ii) 50.73(a) {2)(x)
LEVEL (10) 100%
20.2203(a) (2) (i) 20.2203(a) (3) (ii) 50.73(a)(2)(iii) 73.71 20.2203(a) (2)(ii) 20.2203(a) (4) 50.73(a) (2)(iv) OTHER 20.2203(a) (2) (iii) 50.36(c) {1) 50.73(a)(2) (v) Specify in Abstract below or in NRC Form 36BA 20.2203(a) (2)(iv) 50.36(c) (2) 50.73(a) (2) (vii)
LICENSEE CONTACT FOR THIS LER (12)
NAME TELEPHONE NUMBER Ilnalvde Ates Cade)
Michael Verrilli Sr. Analyst - Licensing (919) 362-2303 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)
CAUSE SYSTEM COMPONENT REPORTABLE REPORTABLE MANUFACTURER CAUSE SYSTEM COMPONENT MANUFACTURER TO NPRDS TO NPRDS SUPPLEMENTAL REPORT EXPECTED (14) MONTH DAY YEAR EXPECTED YES SUBMISSION
{Ifyas, complete EXPECTED SUBMISSION DATE). X No DATE {15)
ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 singla-spaced typawrittan lines) (16)
On January 9, 1998, a condition was identified during operation that results in the plant being potentially outside it' design basis. Specifically, a potential failure mechanism exists where a leak in the non-safety Instrument Air System could result in the inoperabihty of the Steam Generator Pre-heater Bypass Isolation Valves. These valves are safety-related containment isolation valves that are required by plant procedures to automatically shut in 10 seconds or less upon receipt of a Main Feedwater Isolation Signal. These valves are positioned by a pneumatic piston-operated actuator which is supplied by the non-safety related Instrument Air System. They are designed to automatically close if control air supply is lost. However, a "smart" air leak has been postulated in the Instrument Air system that could possibly reduce the air inlet pressure to just low enough to affect proper operation of the actuator's 3-way and 4-way pilot valves and not be detected by Operations personnel. If this occurred, the pilot valves would shuttle, causing the accumulator pressure to bleed off, which would prevent the valves from closing as required. This potential scenario constitutes operation outside the design basis of the plant and was reported to the NRC via the emergency notification system on January 9, 1998 at 1450 hours0.0168 days <br />0.403 hours <br />0.0024 weeks <br />5.51725e-4 months <br />.
The cause of this condition was inadequate design control during development of a plant modification implemented in August 1984 in response to NRC Information Notice 82-25. The investigation for this event also revealed several other missed opportunities to identify this condition during subsequent plant modifications and/or related evaluations.
Immediate corrective actions included development of a Justification for Continued Operation (JCO), with the required evaluation and compensatory measures to ensure continued operability. The JCO evaluation determined that the isolation valves were operable dependent upon once per 12-hour monitoring of the discharge pressure on the inlet regulator for each valve actuator. Additional corrective actions will include a plant modification to resolve the design deficiency, a FSAR revision and training for appropriate Engineering personnel.
9802170061 980209 PDR ADCICK 05000400 S PDR
NRC FORM 366A US. NUCLEAR REGULATORY COMMISSION M-SS)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATlON FACIUTY NAME (I) DOCKET LER NUMBER (6) PAGE (3)
SEOUENTIAL REVISION YEAR NUMBER NUMBER Shearon Harris NUclear Plant - Unit ¹1 50400 2 OF 4 98 - 001 - 00 TEXT Pr more spscers nod. vso oddMmsl copies of ABC Povv 3IFQI (11)
EVENT DESCRIPTION:
On January 9, 1998, with the plant operating in mode 1 at 100% power, a condition was identified that results in the plant being potentially outside the design basis. Specifically, a potential failure mechanism exists where a leak in the non-safety Instrument Air System (EIIS Code: LD) could result in the inoperability of the Steam Generator (S/G) Pre-heater Bypass Isolation Valves (1AF-64, 1AF-102, and 1AF-81, EIIS Code: SJ-V). These valves are safety-related containment isolation valves that are required by plant procedure PLP-106 (Technical Specification Equipment List and Core Operating Limit Report) to automatically shut in 10 seconds or less upon receipt of a Main Feedwater Isolation Signal (MFIS). These valves are positioned by a pneumatic piston-operated actuator and are opened and closed by high pressure air (-150 psig) from the actuator's accumulator. The accumulator is maintained at a higher air pressure by an air intensifier pump. The intensifier is a double piston compressor that is driven by control air. The air intensifier is needed to boost inlet control air pressure to approximately 150 psig, since normal control air pressure (approx. 75 psig) will not close the valve alone. With the use of pneumatic pilot valves located on the actuator, the intensified air in the accumulator is used to position the actuator piston, which is connected to the valve disc by a common shaft. The pneumatic pilot valves guide the intensified air from the accumulator to either face of the piston to position the operator. Directing the high pressure air below the piston will open the S/G pre-heater bypass valves; and directing the high pressure air to the top of the piston will cause the valves to close.
The S/G pre-heater bypass valves are designed to automatically close if control air supply is lost. This is accomplished by two in-series solenoid valves that are energized open to supply the control air from the Instrument Air System. Ifa Main Feedwater Isolation signal is generated the solenoid valves will de-energize, securing the control air supply, which causes the S/G pre-heater bypass isolation valves to close. The solenoid valves will also de-energize if Instrument Air System pressure drops to 66 psig.
However, a "smart" air leak has been postulated in the Instrument Air System piping that could possibly reduce the air inlet pressure to just low enough to affect proper operation of the actuator's 3-way and 4-way pilot valves and not be detected by the pressure switches in the main header of the Instrument Air System that would de-energize solenoid valves at 66 psig.
If this occurred, the pilot valves would shuttle, causing the accumulator pressure to bleed off,thewhich would prevent the valves from closing as required. Operations personnel would have no indication of accumulator low pressure other than local observations made by an auxiliary operator and possibly dual valve indication in the main control room due to the valves cycling slightly.
This potential scenario constitutes potential operation outside the design basis of the plant and was immediately reported to the NRC via the emergency notification system on January 9, 1998 at 1450 hours0.0168 days <br />0.403 hours <br />0.0024 weeks <br />5.51725e-4 months <br />.
CAUSE:
The cause of this condition was inadequate design control during development of a plant modification implemented in August 1984, which was prior to issuance of the Harris Plant Operating License. Specifically, NRC Information Notice 82-25, "Failure of Hiller Actuators Upon Gradual Loss of Air Pressure" stated that on a gradual loss of control air, the pneumatic control valves may assume some intermediate position and cause the stored air in the accumulator to vent to atmosphere and prevent the actuator from performing it's safety function of closing. To resolve this concern Field Change Request (FCR-I-992) was developed to install two pressure switches in the Instrument Air System header that would de-energize the solenoid valves (described above) at 66 psig. However, ANSI Standard N18.2.a "Nuclear Safety Criteria for Design of Stationary PWR Plants" requires that a barrier be installed between safety class interfaces, such as the air control circuit and the instrument air supply. This design requirement was not properly applied during the development of FCR-I-992; therefore, the safety related system did not meet the single failure design criteria.
The investigation for this event also revealed additional missed opportunities to identify this design deficiency in the Instrument Air System. These included: (1) HNP's response development for NRC Generic Letter 88-14 in 1989, which specifically included an evaluation (Plant Change Request, PCR-4151) of the Instrument Air System and the Pre-heater Bypass Valve actuator design, (2) Adverse Condition Report ¹91-314 initiated in June 1991, which identified a leak in the supply air regulator for 1AF-81 that significantly lowered accumulator pressure, and (3) development of PCR-6158 and it's associated evaluation (PCR-6066) in 1992, which implemented a modification to the actuator air circuitry to enhance the air intensifier pump and upgrade portions of the control air piping to safety-related. In each of these cases, the "smart" leak scenario described in this LER was not identified nor considered credible due to the incorrect assumption that the system design met the single failure criteria.
N M (4. )
NRC FORM 366A US. NUCLEAR REGULATORY COMMISSION F)a6)
LlCENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (I) DOCKET LER NUMBER IS) PAGE )3)
SEQUENTIAL REVISION NUMBER NUMBER Shearen Harris Nuclear Plant ~
Unit //1 50400 3 OF 4 98 - Ool .- 00 TEXT P secre s~seir cere'rerf. ese errrFse)re copes el h'RC &m NQl )17)
SAFETY SIGNIFICANCE:
There were no actual consequences associated with this potential failure scenario. JCO 98-01 established the basis for continued operability of the pre-heater bypass valves. These valves are listed as containment isolation valves in PLP-106 (TS Equipment List and Core Operating Limits Report) and are required to automatically shut following receipt of a Main Feedwater Isolation Signal (MFIS). A MFIS is generated from the safety injection actuation logic as well as on high-high S/G level in any steam generator. As such, any accident analyzed in the FSAR that results in a safety injection or high-high S/G level may be impacted by degradation of the non-safety related air supply to these valves.
The list of applicable accidents includes: Main Steamline Break (15.1.5), Feedwater Line Break (15.2.8), Steam Generator Tube Rupture (15.6.3), and Loss of Coolant (15.6.5). Two different failure scenarios must be considered.
The first is the failure of the air supply to the valve actuators coincident with initiation of the accident. This is necessary because a non-safety related component cannot be credited in the mitigation of an accident, nor can the loss of a non-safety related component result in failure of a safety related component. The second condition to be considered is the undetected loss of air to the actuators during normal plant operation such that a pre-existing, degraded con'dition could be created.
Loss of Air Su I Coincident with Accident Initiation Of the events listed above, only the analysis of the Main Steamline Break (MSLB) actually takes credit for isolation of main feedwater flow in mitigating the accident. However, due to the rapid event sequence of this accident scenario, the pre-heater bypass valves would remain capable of performing their safety function as analyzed in the limiting MSLB events.
Loss of Air Su I as a Pre-existin Condition Over the course of the plant life, degradation of the air supply line to the pre-heater bypass valves that could result in a worst case leak (i.e., a smart leak) must be considered as a possibility. This presents a concern in that if the valves were to become disabled due to a small drop in the Instrument Air System header pressure, the inoperable condition of the pre-heater bypass valves may not be detected in the main control room. A MSLB or LOCA that occurred with this pre-existing condition would not progress within the constraints of the current analysis as described above.
Therefore, the compensatory actions identified in the corrective actions section below are necessary to ensure that the air supply to the pre-heater bypass valves remains intact. This action is not intended to be a manual response credited in the mitigation of an accident in place of a normal automatic function. Rather, the increased level of surveillance improves the confidence in the integrity of the Instrument Air System header and ensures high reliability of the pre-heater bypass valves.
PREVIOUS SIMILAREVENTS:
There have been no previous events reported related to a newly identified failure scenario that would potentially render the S/G Pre-heater Bypass Valves inoperable.
NRC FORM 36BA US. NUCLEAR RECULATORY COMMISSION
)4.9S)
LlCENSEE EVENT REPORT (LEB)
TEXT CONTINUATION FACILITY NAME II) OOCXET LER NUMBER IB) PACE I31 SEOUENTIAL REVISION YEAR NUMBER NUMBER Shearon Harris Nuclear Plant ~ Unit //I 50400 4 OF 4 98 - 00'I - 00 TEXT Rl psdsd spssd ss ssqvi ssE ssss ddt')psl copes ol NRC ppfps 3BQI (17)
CORRECTIVE ACTIONS COMPLETED:
- 1. Justification for Continued Operation (JCO ¹98-01) was generated along with the associated engineering evaluation (ESR ¹9800014)..This JCO/evaluation determined that the pre-heater bypass valves remained operable based upon a once per shift (12-hour) monitoring of the discharge pressure on the inlet regulator for each valve actuator to ensure the integrity of the non-safety related instrument air piping.
- 2. The Daily Operations Surveillance Test procedures (OST-1021 and OST-1022) were revised to direct 12-hour monitoring of the discharge pressure on the inlet regulator for each valve actuator.
CORRECTIVE ACTIONS PLANNED:
- 1. A plant modification will be developed and implemented to resolve the smart leak in the non-safety related Instrument Air System header. This will be completed prior to entering Mode< following completion of refueling outage ¹8, which is currently scheduled to begin in October 1998.
- 2. The FSAR will be revised to include a description of the modified S/G Pre-heater Bypass Isolation Valves.
This will be completed by December 31, 1998.
- 3. Training will be provided to appropriate Engineering personnel, which addresses the design control aspects of this event. This will be completed by July 15, 1998.