BW200090, CFR 50.59 Summary Report
| ML20350B431 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 12/15/2020 |
| From: | Keenan J Exelon Generation Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| BW200090 | |
| Download: ML20350B431 (22) | |
Text
December 15, 2020 BW200090 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Braidwood Station, Units 1 and 2 Renewed Facility Operating License Nos. NPF-72 and NPF-77 NRC Docket Nos. STN 50-456 and STN 50-457
Subject:
10 CFR 50.59 Summary Report Pursuant to the requirements of 10 CFR 50.59, "Changes, tests, and experiments," paragraph (d)(2), Braidwood Station is providing the required report for Facility Operating License Numbers NPF-72 and NPF-77. This report is being provided for changes implemented during the time period of June 19, 2018 through June 18, 2020 and consists of the 10 CFR 50.59 Coversheets for changes to the facility or procedures as described in the Updated Final Safety Analysis Report (UFSAR), and tests or experiments not described in the UFSAR.
Please direct any questions regarding this submittal to Mr. Kevin Lueshen, Regulatory Assurance Manager, at (815) 417-2800.
Respectfully, J: n Keenan
~
Site Vice President Braidwood Station
Attachment:
Braidwood Station 10 CFR 50.59 Summary Report cc: NRR Project Manager - Braidwood Station Illinois Emergency Management Agency - Division of Nuclear Safety US NRC Regional Administrator, Region Ill US NRC Senior Resident Inspector (Braidwood Station)
ATTACHMENT Braidwood Station 10 CFR 50.59 Summary Report Evaluation No. Revision Title Approved on 0 09/07/2012
- Lake Screen House Travelling Screen Level Control BRW-E-2012-155 1 04/30/2014 Install/Remove Temporary Control Panel in Support of Ovation BRW-E-2018-38 1
- Digital Upgrade for 7300 NSSS and BOP Cabinets TCCP 01/31/2020 Temporary Change (Unit 2) 0 MSIV ROOMS HELB CALCULATIONS I Short-Term 03/18/2019 BRW-E-2018-80 Pressurization Sub-compartment Analysis of Main Steam Tunnel 1 and Main Steam Isolation Valve Rooms 0212412020 BRW-E-2019-32 0 Secondary Pump Trip Unit 1/2 03/06/2020
- Revision 0 was not issued.
- The 10 CFR 50.59 Summary Report to NRC includes evaluations performed in support of installed configurations. This evaluation was initiated in 2012, however, the supporting activities to install the modification and update the UFSAR weren't completed until 10/25/2019 which falls within the reporting period of this 2020 report.
50.59 REVIEW COVERSHEET FORM LS-AA-104-1001 Revision 3 Pagel of4 Stadon/Ulllit(s): Braidwood I Units l&2 Acthity/Documen¢ Number: EC 388161 / UfSAR DRP 14-087 Revhion Number: __,o._____
Title:
LAKE SCBW HOUSE TRAW ING SCREEN LEVEL CONTROL NOTE: For 50.59 Evaluations, information on this form will provide the basis for preparing the biennial summary report submitted to the NRC in accordance with the requirements of IO CFR 50.59(dX2).
Description or Activity:
(Provide a brief, concise description of what the proposed activity involves.)
The proposed activity eliminates the automatic Lake Screen House (LSH) traveling screen 12" differential level trips associated with each Circulating Water (CW) pump, the 15 second time delay associated with each CW pump's trip, and upgrades the existing traveling screen level measuring instrumentation for both units.
The following changes are included in this activity:
- Removal of six Hagan Ring Balance differential pressure (d/P) level instruments (lLDR-SWOOl, ILDR-SW007, ILDR-SWOl3, 2LDR-SW001, 2LDR-SW007, 2LDR-SW013) located in 1(2)PL02J and replaced with six new d/P Level Transmitters, six new d/P Level Recorders, and a new power supply to power the loops, which will provide the same functions as the removed obsolete instruments.
- Removal ofsix Differential Pressure Switches (ILDS-SWOOI, ILDS-SW007, lLDS-SW013, 2LDS-SW001, 2LDS-SW007, 2LDS-SW013) and six Time Delay Relays (1PL02J-TDRA. IPI..021-TDRB, IPL02J-TDRC, 2PL02J-TDRA, 2PL02J-TDRB, 2PL021-TDRC) located in l(2)PL02J, which are associated with the CW pump trips.
- Modify the Lake Traveling Screen Wash Control Panels 1(2)PL02J to reflect the removal and addition of the above instrumentation.
Reason for Activity:
(Discuss why the proposed activity is being perfonned.)
The control system that includes the measurement of traveling screen differential level and generates a trip function is non-safety related. A bubbler type system is used to measure water levels upstream (lake side) and downstream (pump side) of the traveling screens and these levels are converted to a differential level by a mechanical balance type linkage (Hagan Ring Balance) and recorder. Instrumentation issues, i.e. physical problems with the mechanical device and switch degradation, has resulted in unreliable indication and control functions for the traveling screens, Screen Wash (SW) System, and the CW pumps.
The six Hagan Ring Balance d/P instruments in the Lake Traveling Screen Wash Control Panels l(2)PL02J are obsolete and spare parts are not available in the aftermarket. There are no direct replacements for these devices. There has also been reliability issues associated with the instrumentation. These d/P instruments monitor level differential in the Braidwood cooling pond before and after the Traveling Screens. They will be replaced with new transmitters and recorders providing the same functions as the existing instruments which include control, indication, recording, and alarms.
In addition, the automatic traveling screen differential level trip function for the CW pumps is permanently being removed. This trip function is intended for CW pump protection against operation with inadequate submergence. However, this trip function has resulted in spurious trips of the CW pumps and operational transients throughout Braidwood Station's operational history.
The removal of this trip does not affect the ability to manually trip the CW pumps.
Effect or Activity:
(Discuss how the activity impacts plant operations, design bases, or safety analyses described in the UFSAR)
The proposed modification will eliminate the automatic I.SH traveling screen 12" differential level trip for the CW pumps and upgrades the existing traveling screen level measuring instrumentation for both units. The UFSAR Section 7.7.l.15 will be revised per UFSAR DRP 14-087 to reflect the change of no CW pump trips due to high differential pressures. The existing CW pump d/P indication and indicated pump motor amps will still be available in the Main Control Room. The operator(s) can manually trip the CW pump(s), if conditions are warranted, upon receiving an alarm so the proposed modification will not have an impact on plant operation from the Main Control Room. The removal of the CW pump trip performs the same function as the
50.59 REVIEW COVERSHEET FORM I..S-AA-104-1001 Revision 3 Page 2of4 Stadon/Unit(s): Braidwood I Units l & 2 Activity/Document Number: EC 388161 / UFSAR DRP 14-087 Revision Number: __,Q_ _ __
Title:
LAKE SCBffiN HOVSE TRAYfiJ ING SCBF'.EN LEVEL CQNTROL current procedure BwOP CW-26, DEFEATING CIRCULATING WATER PUMP TRIP ON TRA VEUNG SCREEN DELTA P. to defeat the CW pump trip. BwOP CW-26 currently contains a Limitation and Action (E.2) to trip the affected CW pump \\'ith an observed 30 inch differential level across its associated traveling screen. This design change permanently removes the trip such that this procedure is oo longer necessary. This activity will not affect the other traveling screen modes of operation due to a timer, differential level (normal backwash at 6" M.. and rotate in slow speed or emergency backwash at l 0.. AL and rotate in fast speed), or manual actuation. This activity will not affect any other existing CW pump trips other than the traveling screen 12" differential level. The existing level instrumentation will be replaced with new reliable transmitters and recorders providing the same functions as the existing instruments which include control. indication. recording, and alarms.
There are '> I..SH traveling screens per CW pump ( 12 screens total). The purpose of the traveling screens is to filter any small pieces of matter and debris that have passed through the I..SH bar grills before it can enter the suction for the CW pumps. The purpose of the differential pump trip is to protect the structural integrity of the traveling screens and to protect the CW pump under low suction pressure conditions. The proposed modification will not affect the design functions of the traveling screens or the CW pumps.
Although not explicitly stated within the Braidwood Design Basis (UFSAR), the existence of the 12" water differential level CW pump trip would limit the buildup of debris and resulting traveling screen differential level, thereby reducing the potential for traveling screen over-loading and failure. This condition would also reduce forebay water level such that it minimizes impact on the performance of the Essential Service Water (SX), Fire Protection (FP), and Non-Essential Service Water (WS) Systems.
The implementation of this activity will have no adverse impact to the CW System for condenser operation. the FP System, the WS System for non-safety related cooling loads, and the SX System for safety-related safe-shutdown cooling loads. While the traveling screens and associated differential level instrumentation serve no safety-related function. they are designed to support General Design Criteria 2, Design bases for protection against natural phenomena. in that the traveling screens will not interfere with the successful operation of the safety-related SX System through collapse and blockage of the SX intake structure. The cooling pond intake structure (forebay walls, floors and ceiling, including the traveling screens) is seismically qualified to protect the function of the SX System.
The SX System design basis does not credit the traveling screens to perform any function in support of the SX System mission.
The SX System is utilized to reject core decay heat and safety-related equipment beat to the Ultimate Heat Sink (UHS). The UHS is a part of the cooling pond located in front of the Lake Screen Howe. The significant beat loads on the SX System include core decay heat removal via the Residual Heat (RH) Removal and Component Cooling (CC) Systems and post Loss of Coolant Accident (LOCA) containment cooling. Other loads include safety-related equipment cubicle coolers and oil coolers.
As such, the SX System does receive the appropriate Engineered Safety Features (ESF) actuation signals and is powered by ESF power supplies. The SX System draws water from the UHS, through the traveling screens, and returns the water to the far end of the UHS. The traveling screens are designed to ensure they will not fail in a way that affects the operation of the SX System.
The design of the traveling screens includes allowances for seismic loading, both Operational Basis Events (OBE) and Safe Shutdown Events (SSE). The traveling screens have oo safety-related design basis criteria. The operational requirements of the SX System and the VHS are provided in Technical Specifications 3.7.8 and 3.7.9, respectively. The SW System is not subject to any Technical Specification requirements.
The redundancy and separation at the I..SH maintains the required suppression water source for the FP System. The FP System can be supplied by the SX System, if needed in an emergency. If the traveling screens and SW System function as designed and remove debris, the likelihood of an impact on the entire FP System is not more than minimal.
The WS System is non-safety related and not required for the safe shutdown of the reactor (UFSAR Section 9.2.1.1). The WS pumps do not have a trip function on high traveling screen differential level. The loss of a WS pump or pump performance could potentially result in reaching temperature limitations on Balance of Plant equipment that warrants power reduction, pump trips, or manual reactor trips. Although the absence of the CW pump trip function on high differential level could impact the WS System in this circumstance the likelihood of this occurring is no different than ifthe existing non-safety related CW pump differential level failed to function.
50.59 REVIEW COVERSHEET FORM LS-AA-104-1001 Revision3 Page3of4 Station/Unit(s): Bmid\l!Wd I Units l & 2 Activity/Document Number. EC 388161 / UFSAR DRP 14=087 Revision Number: __,O.___ __
Tittle: LAKE SCBEfN HQUSE IRAVEUNG SCREEN LEVEL CONIROL The elimination of the traveling screen differential level trip function does reduce the defense in depth for a fouling event at the LSH as described in the Braidwood Station response to SOER 07-2, lntake Cooling Water Blockage. However, existing procedure guidance in conjunction with annunciator alarms and operator response will minimize the potential for a gross fouling event affecting plant equipment The failure of the CW System controls can fail in such a manner to prevent traveling screen operation. prevent SW System operation. and prevent CW pump trip. A review of the UFSAR revealed that the turbine trip event, as described in Section 15.2.3, is the only accident that may be affected as a result of the proposed modification to eliminate the LSH traveling screen differential level trip function for the CW pumps. The trip of a CW pump or pumps is an initiating event that results in a loss of condenser vacuum which can cause a turbine trip. Therefore, the end result of the failure is the same as the results currently described in UFSAR Section I 5.2.3 and the results are unchanged for any credible/potential malfunction due to the failure of the CW System controls. The elimination of the traveling screen differential level trip will not increase the frequency of a loss of condenser vacuum or a turbine trip.
Based on the abo ...-e. the design basis and safety analyses of the traveling screens, SW. CW, SX. FP, and WS Systems are not adversely impacted.
Swmnaey of Condmion for the Activity's 50.59 Review:
(Provide justification for the conclusion. including sufficient detail to recognize and understand the essential arguments leading to the conclusion. Provide more than a simple statement that a 50.59 Screening, 50.59 Evaluation, or a License Amendment Request, as applicable, is not required.)
This activity eliminates the automatic LSH tra ...-eling screen 12" differential level trips associated with each Circulating Water (CW) pump, the 15 second time delay associated with each pumps trip, and upgrades the existing traveling screen level measuring instrumentation for both units. Existing CW pump clJP indication and indicated pump motor amps will still be available in the Main Control Room. The operator(s) can continue to manually trip the CW pumps, if conditions are warranted, upon receiving an alarm so the proposed modification will not have an impact on plant operation from the Main Control Room.
The existing level instrumentation will be replaced with new reliable transmitters and recorders providing the same functions as the existing instruments which include control, indication, recording, and alarms. While the traveling screens serve no safety-related function. they are designed to support General Design Criteria 2, Design bases for protection again.st natural phenomena, in that the traveling screens will not interfere with the successful operation of the safety-related SX System through collapse and blockage of the SX intake structure.
The SW System traveling screens installed at the LSH are designed to filter cooling pond debris larger than approximately 3/8".
The Circulating Water System Controls is a control system not required for safety and unrelated to reactor safety as described in UFSAR Section 7.7. l.IS. Although not explicitly stated within the Braidwood Design Basis (UFSAR), the existence of the 12" water differential level CW pump trip would limit the buildup of debris and resulting traveling screen differential level, thereby reducing the potential for traveling screen over-loading and failure. This condition would also reduce forebay water level such that it minimizes impact on the performance of the SX, FP and WS Systems. The design function of the traveling screens is not being changed by the implementation of this activity.
As described above in the effects section and detailed in the attached evaluation. the proposed change does not alter the UFSAR described design function of the SW, CW, SX, FP and WS Systems. Based on a fouling event and indications in the MCR.
Operators will take action upon receiving a CW Pump Low Delta P alarm (BwAR l/2-17-Bl3) and Traveling Screen Trouble alarm (BwAR. l/2-17-El3) of tripping the CW pump(s). Since this activity installs better and more reliable instrumentation, and the alarms are fast enough for operator action on tripping a CW pump, enough margin is ensured and available for plant equipment in the forebay (SX, WS, and FP).
The 50.59 evaluation concluded that the design basis function of the SW, CW, SX, FP and WS Systems are maintained by this change since existing procedure guidance in conjunction with annunciator alarms and operator response will minimize the potential for an event affecting plant equipment, which provides an adequate degree of protection. Since the frequency of accidents or malfunctions are not increased, the consequences of accidents and malfunctions remain bounded. no new accidents
50.59 REVIEW COVERSHEET FORM LS-AA-104-1001 Revision 3 Page4of4 Station/Unit(s): Braidwood I Units I & 2 Activity/Document Number: EC 388161IUFSARDRP14--087 Revision Number: _o.._____
Title:
LAKE SCREEN HOUSE IRAVEIJNG SCBEBN LEVEL CONJRQL or malfunctions are created. and there are no changes to fission product barrier protection or evaluation methodology, the proposed activity may be implemented without prior NRC approval.
Attadiments:
Attach all 50.59 Review forms completed, as appropriate.
Forms Attached: (Check all that apply.)
0 Applicability Review D 50.59 Screenina 50.59 Screening No. Rev.
181 50.59 Evaluation 50.59 Evaluation No. BRW-E-2012-155 Rev. 0
50.59 REVIEW COVERSHEET FORM LS-AA-104-1001 Revision 3 Pagel of 4 Station/Unit(s): Braidwood I Units 1 & 2 Activity/Document Number: EC 388161 / UFSAR DRP l+o87 Revbion Number. _..___ __
Title:
LAKE SCREEN HOUSE TRAVELING SCREEN LEVEL CONIROL NOTE: For 50.59 Evaluations, information on this form will provide the basis for preparing the biennial summary report submitted to the NRC in accordance with the requirements of 10 CFR 50.59(d)(2).
Description or Activity:
(Provide a brief, concise description of what the proposed activity involves.)
The proposed activity eliminates the automatic Lake Screen House (LSH) traveling screen 12" differential level trips associated with each Circulating Water (CW) pump, the 15 second time delay associated with each CW pump's trip, and upgrades the existing traveling screen level measuring instrumentation for both units.
The following changes are included in this activity:
- Removal of six Hagan Ring Balance differential pressure (d/P) level instruments ( JLDR-SWOO I. ILDR-SW007, lLDR-SWO 13. 2LDR-SWOO I, 2LDR-SW007, 2LDR-SWO 13) located in l (2)PL02J and replaced with six new d/P Level Transmitters, six new d/P Level Recorders, and a new power supply to power the loops, which will provide the same functions as the removed obsolete instruments.
o Removal of six Differential Pressure Switches (lLDS-SWOOI, 1LDS-SW007, ILDS-SWOI3, 2LDS-SWOOI, 2LDS-SW007, 2LDS-SWOl3) and six Time Delay Relays (IPL02J-TDRA. 1PL02J-TDRB, 1PL02J-TDRC. 2PL02J-TDRA.
2PL02J-TDRB, 2PL02J-TDRC) located in 1(2)PL02J, which are associated with the CW pump trips.
o Modify the Lake Traveling Screen Wash Control Panels l (2)PL02J to reflect the removal and addition of the above instrumentation.
o Generate UFSAR DRP 14-087 to revise the affected UFSAR sections, as applicable.
Reason for Activity:
(Discuss why the proposed activity is being perfonned.)
The control system that includes the measurement of traveling screen differential level and generates a trip function is non-safety related. A bubbler type system is u.sed to measure water levels upstream (lake side) and downstream (pump side) of the traveling screens and these levels are converted to a differential level by a mechanical balance type linkage (Hagan Ring Balance) and recorder. Instrumentation issues, i.e. physical problems with the mechanical device and switch degradation, has resulted in unreliable indication and control functions for the traveling S(.'feens, Screen Wash (SW) System. and the CW pumps.
The six Hagan Ring Balance d/P instruments in the Lake Traveling Screen Wash Control Panels 1(2)PL02J are obsolete and spare parts are not available in the aftermarket. There are no direct replacements for these devices. There has also been reliability issues associated with the instrumentation. These d/P instruments monitor level differential in the Braidwood cooling pond before and after the Traveling Screens. They will be replaced with new transmitters and recorders providing the same functions as the existing instruments which include control, indication, recording, and alarms.
In addition. the automatic traveling screen differential level trip function for the CW pumps is permanently being removed. This trip function is intended for CW pump protection against operation with inadequate submergence. However, this trip function has resulted in spurious trips of the CW pumps and operational transients throughout Braidwood Station *s operational history.
The removal of this trip does not affect the ability to manually trip the CW pumps.
Effect of Activity:
(Discuss how the activity impacts plant operations, design bases, or safety analyses described in the UFSAR.)
The proposed modification will eliminate the automatic LSH traveling screen 12" differential level trip for the CW pumps and upgrades the existing traveling screen level measuring instrumentation for both units. The UFSAR Section 7.7.1.15 will be revised per UFSAR DRP I +o87 to reflect the change of no CW pump trips due to high differential pressures. The existing CW pump d/P indication and indicated pump motor amps will still be available in the Main Control Room. The operator(s) can manually trip the CW pump(s), if conditions are W"dJTanted, upon receiving an alarm so the proposed modification will not have an impact on plant operation from the Main Control Room. The removal of the CW pump trip performs the same function as rhe
50.59 REVIEW COVERSHEET FORM LS-AA-104-1001 Revision 3 Page 2 of4 Station/Unit(s): Braidwood I Units l & 2 Activity/Document Number: EC 388161 I UFSAR DRP I 4-087 Revbion Number: _.....___ __
Title:
LAKE SCREEN HOUSE TRAVELING SCREEN LEVEL CONTRQL current procedure BM>P CW-26, DEFEATING CIRCULATING WATER PUMP TRIP ON TRA YEUNG SCREEN DELTA P, to defeat the CW pump trip. BwOP CW-26 currently contains a Limitation and Action (E.2) to trip the affected CW pump with an observed 30 inch differential level across its associated traveling screen. This design change permanently removes the trip such that this procedure is no longer necessary. This activity will not affect the other traveling screen modes of operation due to a timer, differential level (normal backwash at 6" .1L and rotate in slow speed or emergency backwash at to" .1L and rotate in fast speed), or manual actuation. This activity will not affect any other existing CW pump trips other than the traveling screen 12" differential level. The existing level instrumentation will be replaced with new reliable transmitters and recorders providing the same functions as the existing instruments which include control, indication, recording, and alarms.
There are two LSH traveling screens per CW pump (12 screens total). The purpose of the traveling screens is to filter any small pieces of matter and debris that have passed through the LSH bar grills before it can enter the suction for the CW pumps.
Although the UFSAR mentions the existence of the CW pump trip, there is no discussion or evaluation of any impact on SX or other interfacing systems. Therefore. there is no impact to UFSAR-described design functions, which are UFSAR described design bases functions and other SSC functions described in the UFSAR that support or impact design bases functions. There is no safety design basis associated with the Circulating Water System, i.e. specifically the CW pump trip.
Although not explicitly stated within the Braidwood Design Basis (UFSAR). the e:idstence of the 12" water differential level CW pump trip would limit the buildup of debris and resulting traveling screen differential level, thereby reducing the potential for traveling screen over-loading and failure. This condition would also reduce forebay water level such that it minimizes impact on the perfonnance of the Essential Service Water (SX). Fire Protection (FP), and Non-Essential Service Water (WS) Systems.
The implementation of this activity will have no adverse impact to the CW System for condenser operation. the FP System, the WS System for non-safety related cooling loads, and the SX System for safety-related safe-shutdown cooling loads. While the traveling screens and associated differential level instrumentation serve no safety-related function. they meet the requirements of General Design Criteria 2. Design bases for protection against natural phenomena, in that the traveling screens will not interfere with the successful operation of the safety-related SX System through collapse and blockage of the SX intake structure. The cooling pond intake structure (forebay walls, floors and ceiling, including the traveling screens) is seismically qualified to protect the function of the SX System.
The SX System design basis does not credit the traveling screens to perform any function in support of the SX System mission.
The SX System is utilized to reject core decay heat and safety-related equipment heat to the Ultimate Heat Sink (UHS). The UHS is a part of the cooling pond located in front of the Lake Screen House. The significant heat loads on the SX System include core decay heat removal via the Residual Heat (RH) Removal and Component Cooling (CC) Systems and post Loss of Coolant Accident (LOCA) containment cooling. Other loads include safety-related equipment cubicle coolers and oil coolers.
As such, the SX System does receive the appropriate Engineered Safety Features (ESF) actuation signals and is powered by ESF power supplies. The SX System draws water from the UHS. through the traveling screens, and returns the water to the far end of the UHS. The traveling screens are designed to ensure they will not fail in a way that affects the operation of the SX System.
The design of the traveling screens includes allowances for seismic loading, both Operational Basis Events (OBE) and Safe Shutdown Events (SSE). The traveling screens have no safety-related design basis criteria. The operational requirements of the SX System and the UHS are provided in Technical Specifications 3.7.8 and 3.7.9. respectively. The SW System is not subject to any Technical Specification requirements.
The redundancy and separation at the LSH maintains the required suppression water source for the FP System. The FP System can be supplied by the SX System. if needed in an emergency. If the traveling screens and SW System function as designed and remove debris, the likelihood of an impact on the entire FP System is not more than minimal.
The WS System is non-safety related and not required for the safe shutdown of the reac1or (UFSAR Se<..1ion 9.2.1.1 ). The WS pumps do not have a trip function on high traveling screen differential level. The loss of a WS pump or pump performance could potentially result in reaching temperature limitations on Balance of Plant equipment that warrants power reduction, pump trips, or manual reactor trips. Although the absence of the CW pump trip function on high differential level could impact the WS System in this circumstance the likelihood of this occurring is no different than if the existing non-safety related CW pump differential level failed lo function.
50.59 REVIEW COVERSHEET FORM l.S-AA-104-1001 Revision3 Page 3 of 4 Stadon/Unit(s): Braidwood I Unjts I & 2 Activity/Document Number: EC 388161 I UFSAR DRP 14-087 Rev~ion Number: - - - - -
Title:
LA.KE SCREEN HOUSE 'fRM'EI ING SCBfEN LEVEL CONTROL The elimination of the traveling screen differential level trip function does eliminate a non-safety related protective measure in a fouling event at the l.SH as described in the Braidwood Station response to SOER 07-2, Intake Cooling Water Blockage. It should be noted that SOER 07-2 was based on intake cooling water blockage that adversely affects safety-related systems and plant reliability. The proposed design change removes the automatic traveling screen differential level CW pump trip which removes a non-safety related automatic function and is acceptable because multiple alternate non-safety related protective measures for the CW pump trip function continue to be accomplished via other existing trips and existing procedure guidance in conjunction with annunciator alarms and operator response which will minimize the potential for a gross fouling event affecting plant equipment. Therefore, a manual action of tripping the CW pumps in the event of a high differential level is not required to maintain interfacing system operability (i.e., SX system, etc).
The failure of the CW System controls can fail in such a manner to prevent traveling screen operation. prevent SW System operation. and prevent CW pump trip. A review of the UFSAR revealed that the turbine trip event. as described in Section 15.2.3, is the only accident that may be affected as a result of the proposed modification to eliminate the LSH traveling screen differential level trip function for the CW pumps. The trip of a CW pump or pumps is an initiating event that results in a loss of condenser vacuum which can cause a turbine trip. Therefore, the end result of the failure is the same as the results currently described in UFSAR Section 15.2.3 and the results are unchanged for any credible/potential malfunction due to the failure of the CW System controls. The elimination of the traveling screen differential level trip will not increase the frequency of a loss of condenser vacuum or a turbine trip.
Based on the above, the design basis and safety analyses of the traveling S(,,'feens, SW, CW, SX, FP, and WS Systems are not adversely impacted.
Summary of Conclusion for the Activity's 50.59 Review:
(Provide justification for the conclusion. including sufficient detail to recognize and understand the essential arguments leading to the conclusion. Provide more than a simple statement that a 50.59 Screening, 50.59 Evaluation. or a License Amendment Request, as applicable, is not required.)
This activity eliminates the automatic l.SH traveling screen 12" differential level trips associated with each Circulating Water (CW) pump, the 15 second time delay associated with each pumps trip, and upgrades the existing traveling screen level measuring instrumentation for both units. Existing CW pump d/P indication and indicated pump motor amps will still be available in the Main Control Room. The operator{s) can continue to manually trip the CW pumps, if conditions are warranted, upon receiving an alarm so the proposed modification will not have an impact on plant operation from the Main Control Room.
The existing level instrumentation will be replaced with new reliable transmitters and recorders providing the same functions as the existing instruments which include control, indication, recording, and alarms. While the traveling screens serve no safety-related function, they meet the requirements of General Design Criteria 2, Design bases for protection against natural phenomena. in that the traveling screens will not interfere with the successful operation of the safety-relaied SX System through collapse and blockage of the SX intake structure.
The SW System traveling screens installed at the l.SH are designed to filter cooling pond debris larger than approximately 3/8".
The Circulating Water System Controls is a control system not required for safely and unrelated to reactor safety as described in UFSAR Section 7.7.1.15. Although not explicitly stated within the Braidwood Design Basis (UFSAR), the existence of the 12 04 water differential level CW pump trip would limit the buildup of debris and resulting traveling screen differential level, thereby reducing the potential for traveling screen over-loading and failure. This condition would also reduce forebay water level such that it minimizes impact on lhe performance of the SX. FP and WS Systems. The function of the traveling screens is not being changed by the implementation of this activity.
As described above in the effects section and detailed in the attached evaluation, the proposed change does not alter the UFSAR described design function of the SW, CW, SX, FP and WS Systems. Based on a fouling event and indications in the MCR, Operators will take action upon receiving a CW Pump Low Della P alarm (BwAR l/2-l7*B13) and Traveling Screen Trouble alarm (BwAR 112-17-El 3) of tripping the CW pump(s). Since this activity installs better and more reliable instrumentation, and the alarms are fast enough for operator action on tripping a CW pump. enough margin is ensured and available for plant equipment in the forebay (SX. WS, and FP).
50.59 REVIEW COVERSHEET FORM LS*AA* l M-100 l Revision 3 Page4of4 Station/Unit(s): Braidwood I Units I & 2 Activity/Document Number: EC 388161/UFSARDRP14-087 Revision Number: __,.___ __
Title:
LAKE SCREEN HOUSE IRAVELING SCREEN LEVEL CONTRQL The 50.59 evaluation concluded that the design function of the SW, CW, SX, FP and WS Systems are maintained by this change since existing procedure guidance in conjunction with annunciator alarms and operator response will minimize the potential for an event affecting plant equipment, which provides an adequate degree of protection. Since the frequency of accidents or malfunctions are not increased. the consequences of accidents and malfunctions remain bounded, no new accidents or malfunctions are created, and there are no changes to fission product barrier protection or evaluation methodology, the proposed activity may be implemented without prior NRC approval.
Attachments:
Attach all 50.59 Review forms completed, as appropriate.
Forms Attached: (Check all that apply.)
D Applicability Review D 50.59 Screening 50.59 Screening No. Rev.
181 50.59 Evaluation 50.59 Evaluation No. BRW-E-2012-155 Rev.
50.59 REVIEW COVERSHEET FORM LS-AA- I 04-100 I Revision 4 Page I of3 Station/Unit(s): Braidwood Unit 2 Activity/Document Number: EC 404450 (U2) Revision Number:
Title:
lnstalll:'Remove Temporarv Control Panel in Support of Ovation Digital Upgrade for 7300 NSSS and BOP Cabinets TCCP Temporary Change (Unit 2)
NOTE: For 50.59 Evaluations, information on this fonn will provide the basis for preparing the biennial summary report submitted to the NRC in accordance with the requirements of l 0 CFR 50.59(d)(2).
Description of Activity:
(Provide a brief, concise description of what the proposed activity involves.)
The proposed activity is a temporary configuration change (TCC) that will be perfonned in support of separate activities which are upgrading multiple nuclear steam supply system (NSSS) and balance of plant (BOP) control systems and transferring them from the existing 7300-series system to an Ovation-based distributed control system (DCS) under EC 400436 and EC 404362.
The TCC involves the transfer of instrument loops which are to be upgraded but which are also required andfor desired to be available during the transition of the plant into the refueling outage. During the refueling outage, the permanent changes to the NSSS and BOP control systems will be made.
The proposed activity involves the installation of:
- a temporary control panel (TCP) in the auxiliary electrical equipment room (AEER) that will house the necessary controllers and inputloutput (1/0) modules
- temporary cables from the existing NSSS (PA05J, PA06J, PA07J, PAOSJ) and BOP (PA20JA, PA20JB) cabinets to the TCP for the affected instrument loops
- two temporary workstations in the control room
- connections of the temporary workstations and TCP to the Ovation network This configuration will connect the affected instrument loops to the Ovation-based DCS and allow for their operation via the temporary workstations. This will ensure that key process variables can be monitored and controlled during the transition to and from the EC 404450 implementing refueling outages. The TCC allows work required for the permanent upgrade of the NSSS and BOP control systems to the Ovation-based DCS to be initiated prior to commencement of the implementing refueling outages.
The TCP will be removed once it is no longer required, and the affected control systems will then be a part of the pennanently upgraded NSSS and BOP control systems.
The proposed activity involves the following systems:
I. Distributed Control System (DCS)
- 2. Reactor Coolant System (RC/RY)
- 3. Chemical and Volume Control System (CV)
- 4. Residual Heat Removal (RH)
- 5. Safety Injection (SI)
- 6. Component Cooling (CC)
- 7. Condensate (CD)
- 8. Essential Service Water (SX)
- 9. Reactor and Containment Floor Drains (RF)
The proposed activity does not involve the major NSSS or BOP control systems (i.e., control ofTavg, pressurizer pressure level, steam dumps, steam generator level, feedwater pump speed, or heater drain level). The proposed activity does not involve the reactor protection or engineered safety features actuation systems. Most of the affected instrument loops are used to monitor the status of equipment and provide indication and alarm functions only. The affected instrument loops with the potential to impact plant processes -such as boric acid flow, charging flow, or VCT level - are not transferred until the reactor is shut down and the
50.59 REVIEW COVERSHEET FORM LS-AA- I04-!00 I Revision 4 Page 2 of3 Station/Unit(s): Braidwood Unit 2 Activity/Document Number: EC 404450 (U2) Revision Number:
Title:
Insta!URemove Temporarv Control Panel in Support of Ovation Digital Upgrade for 7300 NSSS and BOP Cabinets TCCP Temporarv Change <Unit 2)
RCS is borated to the cold shutdown condition. Some of the affected instrument loops do involve UFSAR-described design functions, as discussed in the 50.59 screening.
Reason for Activity:
(Discuss why the proposed activity is being performed.)
The activity is part of an overall phased project to upgrade key process control systems integrating them into a DCS and eliminating the existing Westinghouse 7300-series process control systems - to address equipment reliability and obsolescence issues. This TCC will ensure that key process variables can be monitored and controlled during the transition to and from the EC implementing refueling outages. The TCC allows work required for the permanent upgrade of the NSSS and BOP control systems to the Ovation-based DCS to be initiated prior to commencement of the implementing refueling outages.
Effect of Activity:
(Discuss how the activity impacts plant operations, design bases, or safety analyses described in the UFSAR.)
The TCC will install a temporary control panel in the AEER and two temporary workstations in the control room. Most of the affected instrument loops are used to monitor the status of equipment and provide indication and alarm functions only. These instrument loops will be transferred to the TCP 1-4 weeks before the refueling outage. The affected instrument loops with the potential to impact plant processes such as boric acid flow, charging flow, or volume control tank level - are not transferred until the reactor is shut down and the RCS is borated to the cold shutdown condition.
The TCC will ensure that key process variables can be monitored and controlled during the transition to and from the refueling outage via the temporary workstations. The transfer of the affected instrument loops to the TCP will result in the loss of certain main control board indication, control, or alarm (annunciator) functions. In some cases, the alarm or control functions are maintained via the Ovation-based system. If an existing alarm function triggers an Ovation-based alarm, that alarm will also actuate the "Ovation System Trouble" annunciator. In addition, reactor operators will periodically monitor the affected instrument loops for proper behavior and will respond to Ovation alarms per approved station procedures.
Summary of Conclusion for the Activity's 50.59 Review:
(Provide justification for the conclusion, including sufficient detail to recognize and understand the essential arguments leading to the conclusion. Provide more than a simple statement that a 50.59 Screening, 50.59 Evaluation, or a License Amendment Request, as applicable, is not required.)
The proposed activity involves the installation of a substantial amount of digital hardware and software, as well as significant changes to the human-machine interface. In accordance with the guidance provided in NEI 01-0 I, these aspects of the proposed activity were "screened in" for further evaluation under the 50.59 process for Question I regarding a change to an SSC that adversely affects an UFSAR-described design function and Question 2 regarding a change to a procedure that adversely affects how UFSAR-described SSC design functions are performed or controlled. The 50.59 Screening concluded in the response to Questions 3, 4 and 5 that the proposed activity: does not involve an adverse change to an element of a UFSAR-described evaluation methodology nor use an alternate evaluation methodology used in establishing the design bases or used in the safety analyses, does not involve a test or experiment described in the UFSAR where an SSC is utilized or controlled in a manner that is outside of the reference bounds of the design for that SSC, and does not require a change to the Technical Specifications or Facility Operating License.
The rigorous process used in developing the digital hardware and software were credited in the 50.59 Evaluation with ensuring the modified systems would perform as required. The involvement of the plant operations staff in the development of the human-machine interface and the Braidwood operator training process were credited with ensuring a successful operator interface with
50.59 REVIEW COVERSHEET FORM LS-AA-104-100 l Revision 4 Page 3 of3 Station/Unit(s): Braidwood Unit 2 Activity/Document Number: EC 404450 (U2) Revision Number:
Title:
lnstalliRemove Temporary Control Panel in Support of Ovation Digital Up!!rade for 7300 NSSS and BOP Cabinets TCCP Temporary Change (Unit 2) the new system. As a result, the proposed activity will not result in more than a minimal increase in the frequency of occurrence of an accident previously evaluated in the UFSAR or more than a minimal increase in the likelihood of a malfunction of an SSC important to safety previously evaluated in the UFSAR. Since the affected instrument loops with the potential to impact plant processes - such as boric acid and primary water flow, charging flow, or volume control tank level - are not transferred until the reactor is shut down and the RCS is borated to the cold shutdown condition, the accidents with the potential to be impacted are limited to boron dilution events, and the possibility for an accident of a different type is not created. ~~~~l'!filL~~~§!
radiological consequences would occur, and no design basis limit for a fission product barrier as described in the UFSAR is exceeded or altered. Since administrative controls to isolate the RCS from potential sources of unborated water will be implemented in Modes 3, 4, and 5 to ensure that an inadvertent dilution cannot occur, the possibility for a malfunction with a different result is not created. The proposed activity does not involve a method of evaluation described in the UFSAR.
Attachments:
Attach all 50.59 Review forms completed, as appropriate.
Forms Attached: (Check all that apply.)
D Applicability Review rg] 50.59 Screening 50.59 Screening No. BRW-S-2018-37 Rev.
rg] 50.59 Evaluation 50.59 Evaluation No. BRW-E-2018-38 Rev.
See LS-AA-I 04, Section 5, Documentation, for record retention requirements for this and all other 50.59 forms associated with the Activity.
50.59 REVIEW COVERSHEET FORM LS-AA-104-1001 Revision 4 Page 1 of2 Station/Unit(s): Braidwood Station/Units 1 and 2 Activity/Document Number: EC 396348 / DRP #17-056 Revision Number: 000 I 000
Title:
MSIV ROOMS HELB CALCULATIONS I Short-Term Pressurization Subcompartment Analvsis of Main Steam Tunnel and Main Steam Isolation Valve Rooms NOTE: For 50.59 Evaluations, information on this form will provide the basis for preparing the biennial summary report submitted to the NRC in accordance with the requirements of 10 CFR 50.59(d)(2).
Description of Activity:
(Provide a brief, concise description of what the proposed activity involves.)
This activity is EC 396348 which is a design change document change request (DCR). The changes identified in EC 396348 include updates to the UFSAR, therefore, DRP #17-056 is included in this activity.
Reason for Activity:
(Discuss why the proposed activity is being performed.)
UFSAR Section 3.6 describes the protection against dynamic effects associated with the postulated break of piping. One effect is described as structural loading due to pressurization resulting from a pipe break. Calculation 3C8-0282-001, Rev. 003 calculates the short-term peak and transient pressures in the Main Steam Tunnel (MST) and Main Steam Isolation Valve (MSIV)
Room subcompartments following a high energy line break (HELB). Results from this calculation are discussed in UFSAR Section 3.6 and UFSAR Attachments A3.6 and C3.6. Design errors have been identified in this short-term pressurization subcompartment analysis. These errors are documented in several AR's (e.g., IR 4085634 and 4085649 for Byron [also applicable for Braidwood]). In order to resolve the identified errors, calculation 3C8-0282-001 has been updated and determines new short-term peak and transient pressures in the Main Steam Tunnel (MST) and Main Steam Isolation Valve (MSIV) Room subcompartments following a high energy line break (HELB).
Effect of Activity:
(Discuss how the activity impacts plant operations, design bases, or safety analyses described in the UFSAR.)
The effect of this activity is to determine bounding mass & energy (M&E) releases for each MST and MSIV Room subcompartment and to determine the peak and transient short-term pressures and update the UFSAR with the new results.
There is no impact on plant operations or response to any accidents. The change to the UFSAR provides the updated safety analysis as described in UFSAR Section 3.6.
Summary of Conclusion for the Activity's 50.59 Review:
(Provide justification for the conclusion, including sufficient detail to recognize and understand the essential arguments leading to the conclusion. Provide more than a simple statement that a 50.59 Screening, 50.59 Evaluation, or a License Amendment Request, as applicable, is not required.)
An Applicability Review was performed and determined a portion of the activity involves a change to the UFSAR that removes excessive detail as discussed in NEI 98-03. The information removed is design information from UFSAR Attachment C3.6 that is not important to the description of the facility or of the IOCFR50.2 design bases or safety analyses of the facility.
Screening Conclusions The new thermal-hydraulic analysis for evaluating the short-term pressurization event in the MST and MSIV Room subcornpartments, contained in calculation 3C8-0282-001, Rev. 004 and implemented via EC 396348, involves the use of an alternative methodology from what was previously used in establishing the effect of the design bases HELBs in these areas.
This is evaluated in the 50.59 Evaluation in Question 8.
All other aspects of the proposed activity, including resolution of the errors in the short-term pressurization subcompartment analysis, do not alter the UFSAR described function of any SSCs.
There is no adverse effect on the way that the UFSAR described design function is performed or controlled. There is no testing involved with this change.
This change does not require a change to the Technical Specifications or Operating License.
50.59 REVIEW COVERSHEET FORM LS-AA-104-1001 Revision 4 Page 2 of2 Station/Unit(s): Braidwood Station/Units l and 2 Activity/Document Number: EC 396348 / DRP #17-056 Revision Number: "'"00""'0,,_/,__00"'-=0'-----
Title:
MSIV ROOMS HELB CALCULATIONS I Short-Term Pressurization Subcompartment Analysis of Main Steam Tunnel and Main Steam Isolation Valve Rooms Evaluation Conclusions The attached evaluation addresses the effect involving the use of an alternative methodology from what was previously used in establishing the effect of design bases HELBs in the MST and MSIV Room subcompartments. However, 50.59 Evaluation BRW-E-2018-80, determined that the use of GOTHIC to evaluate the effects of HELBs in the MST and MSIV Room subcompartments, rather than RELAP, is not departure from a method described in the UFSAR because it is appropriate for the intended application and the method has been approved by the NRC.
Based on the above, and as detailed in the attached 50.59 screening and evaluation, this activity may proceed per normal plant processes and procedures without requesting prior approval from the NRC.
Attachments:
Attach all 50.59 Review forms completed, as appropriate.
Forms Attached: (Check all that apply.)
~ Applicability Review
~ 50.59 Screening 50.59 Screening No. BRW-S-2018-79 Rev. --'-o______
~ 50.59 Evaluation 50.59 Evaluation No. BRW-E-2018-80 Rev. _..;;.o_ _ _ _ __
See LS-AA-104, Section 5, Documentation, for record retention requirements for this and all other 50.59 forms associated with the Activity.
50.59 REVIEW COVERSHEET FORM LS-AA-104-1001 Revision 4 Page 1 of2 Station/Unit(s): Braidwood Station/Units 1 and 2 Activity/Document Number: EC 396348 I DRP #17-056 Revision Number: ""'00~0._./_,0..,,0"'"0_ __
Title:
MSIV ROOMS HELB CALCULATIONS I Short-Term Pressurization Subcompartment Analysis of Main Steam Tunnel and Main Steam Isolation Valve Rooms NOTE: For 50.59 Evaluations, information on this form will provide the basis for preparing the biennial summary report submitted to the NRC in accordance with the requirements of 10 CFR 50.59(d)(2).
Description of Activity:
(Provide a brief, concise description of what the proposed activity involves.)
This activity is EC 396348 which is a design change document change request (DCR). The changes identified in EC 396348 include updates to the UFSAR, therefore, DRP#17-056 is included in this activity.
Reason for Activity:
(Discuss why the proposed activity is being performed.)
UFSAR Section 3.6 describes the protection against dynamic effects associated with the postulated break of piping. One effect is described as structural loading due to pressurization resulting from a pipe break. Calculation 3C8-0282-001, Rev. 003 calculates the short-term peak and transient pressures in the Main Steam Tunnel (MST) and Main Steam Isolation Valve (MSIV)
Room subcompartments following a high energy line break (HELB). Results from this calculation are discussed in UFSAR Section 3.6 and UFSAR Attachments A3.6 and C3.6. Design errors have been identified in this short-term pressurization subcompartment analysis. These e1Tors are documented in several AR's (e.g., IR 4085634 and 4085649 for Byron [also applicable for Braidwood]). In order to resolve the identified errors, calculation 3C8-0282-001 has been updated and determines new short-term peak and transient pressures in the Main Steam Tunnel (MST) and Main Steam Isolation Valve (MSIV) Room subcompartments following a high energy line break (HELB).
Effect of Activity:
(Discuss how the activity impacts plant operations, design bases, or safety analyses described in the UFSAR.)
The effect of this activity is to determine bounding mass & energy (M&E) releases for each MST and MSIV Room subcompartment and to determine the peak and transient short-term pressures and update the UFSAR with the new results.
There is no impact on plant operations or response to any accidents. The change to the UFSAR provides the updated safety analysis as described in UFSAR Section 3.6.
Summary of Conclusion for the Activity's 50.59 Review:
(Provide justification for the conclusion, including sufficient detail to recognize and understand the essential arguments leading to the conclusion. Provide more than a simple statement that a 50.59 Screening, 50.59 Evaluation, or a License Amendment Request, as applicable, is not required.)
An Applicability Review was performed and determined a portion of the activity involves a change to the UFSAR that removes excessive detail as discussed in NEI 98-03. The information removed is design information from UFSAR Attachment C3.6 that is not important to the description of the facility or of the 10CFR50.2 design bases or safety analyses of the facility.
Screening Conclusions The new thermal-hydraulic analysis for evaluating the short-term pressurization event in the MST and MSIV Room subcompartments, contained in calculation 3C8-0282-001, Rev. 004 and implemented via EC 396348, involves the use of an alternative methodology from what was previously used in establishing the effect of the design bases HELBs in these areas.
This is evaluated in the 50.59 Evaluation in Question 8.
All other aspects of the proposed activity, including resolution of the errors in the short-term pressurization subcompartment analysis, do not alter the UFSAR described function of any SSCs.
There is no adverse effect on the way that the UFSAR described design function is performed or controlled. There is no testing involved with this change.
This change does not require a change to the Technical Specifications or Operating License.
50.59 REVIEW COVERSHEET FORM LS-AA-104-100 I Revision 4 Page 2 of2 Station/Unit(s): Braidwood Station/Units 1 and 2 Activity/Document Number: EC 396348 / DRP #17-056 Revision Number: ""'00""'0'""'/_,0~0""0_ __
Title:
MSIV ROOMS HELB CALCULATIONS I Short-Term Pressurization Subcompartment Analysis of Main Steam Tunnel and Main Steam Isolation Valve Rooms Evaluation Conclusions The attached evaluation addresses the effect involving the use of an alternative methodology from what was previously used in establishing the effect of design bases HELBs in the MST and MSIV Room subcompartments. However, 50.59 Evaluation BRW-E-2018-80, determined that the use of GOTHIC to evaluate the effects of HELBs in the MST and MSIV Room subcompartments, rather than RELAP, is not departure from a method described in the UFSAR because it is appropriate for the intended application and the method has been approved by the NRC.
Based on the above, and as detailed in the attached 50.59 screening and evaluation, this activity may proceed per normal plant processes and procedures without requesting prior approval from the NRC.
Attachments:
Attach all 50.59 Review forms completed, as appropriate.
Forms Attached: (Check all that apply.)
~ Applicability Review
~ 50.59 Screening 50.59 Screening No. BRW-S-2018-79 Rev. _o"-------
~ 50.59 Evaluation 50.59 Evaluation No. BRW-E-2018-80 Rev. -""1_ _ _ _ __
See LS-AA-104, Section 5, Documentation, for record retention requirements for this and all other 50.59 forms associated with the Activity.
50.59 REVIEW COVERSHEET FORM LS-AA-104- JOO I Revision 4 Page I of5 Station/Unit(s): Byron & Braidwood/Units 1&2 Activity/Document Number: 1/2BOA SEC-I, I/2Bw0A SEC-I Revision Number: I I4/115/I 12/I 10
Title:
Secondarv Pump Trip Unit 1/2 NOTE: For 50.59 Evaluations, infonnation on this form will provide the basis for preparing the biennial summary report submitted to the NRC in accordance with the requirements of I 0 CFR 50.59( d)(2).
Description of Activity:
(Provide a brief, concise description of what the proposed activity involves.)
This Activity revises abnormal operating procedures 1/2BOA SEC- I and J/2BwOA SEC- I to include actions to operate the Auxiliary Feedwater (AF) system during unit startup in Mode !(reactor power <15% and prior to synchronization of the main generator to the grid) and Mode 2 to address a loss of the nonnal feedwater (FW) flow without requiring a reactor trip. Current procedural direction requires a manual reactor trip when the normal FW system is lost in Modes I and 2. The procedure changes being implemented under this Activity will direct Operations to start the AF system and stabilize the unit at a power level in Mode 2 within the capability of the AF pumps following a loss of the normal FW with reactor power initially less than 15%1 during a unit startup and prior to synchronization of the main generator to the grid. Operation will continue in Mode 2 until the normal FW system can be restored or the unit will be shutdown if required by operational or secondary chemistry conditions.
Continued operation without hydrazine injection with the AF system feeding steam generators is limited to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
In accordance with current abnormal procedures, the reactor will be manually tripped ifthe normal FW flow is completely lost in Mode I at power levels greater than 15%. Procedural direction already exists in l/2BOA SEC-I and l/2Bw0A SEC-I to operate the AF system to address a loss of the normal FW flow in Mode 3.
As documented in the supporting 50.59 Evaluation No. 60-19-005/BRW-E-2019-32, technical justification to support the operation of the Auxiliary Feedwater (AF) system in the event that nonnal feedwater is lost during a unit startup prior to synchronizing the main generator to the grid is provided in EC 628034 for Byron and EC 627959 for Braidwood.
Reason for Activity:
(Discuss why the proposed activity is being perfonned.)
The purpose of the procedure changes is to manually start the AF pumps and not immediately trip the reactor following a loss of the normal feedwater during unit startup in Mode I (at reactor <15% and prior to synchronization of the main generator to the grid) and Mode 2. During startup, the normal FW flow is initially from only one FW pump placed in service. In some circumstances, the availability of the other FW pumps may be limited due to continuation of maintenance following scheduled or forced outages. This condition reduces the normal redundancy provided by the FW system during the unit startup and increases the possibility ofa loss of normal FW to occur.
The procedure changes will allow operators to stabilize steam generator levels with the AF system and eliminate need to trip the reactor, which in turn eliminates the plant transient associated with a reactor trip.
Effect of Activity:
(Discuss how the activity impacts plant operations, design bases, or safety analyses described in the UFSAR.)
The procedure changes implemented under this Activity will allow Operators to stabilize steam generator levels using the AF system following a loss of the normal FW flow in Mode I(reactor power <15% and prior to synchronization of the main generator to the grid) and Mode 2 and restore a normal FW flow source without placing an operational transient on the plant resulting from a reactor trip. Since current procedural direction and automatic design features exist to start the AF pumps following the loss of the normal FW flow and reactor trip, there is no change in the demands on the AF system and the AF system would be operated consistent with the current procedures and design basis.
The loss of the normal FW flow is an accident analyzed in Chapter 15 of the UFSAR. However, the accident is analyzed for the bounding case with reactor power initially at I 00% of rated power. The analysis considers a reactor trip occurring on low steam generator level and an automatic start of the AF pumps. The UFSAR does not specifically analyze a loss of the normal FW flow at low power/start-up since this condition would be bounded by the I00% event. The use of the AF system to pre-emptively respond to stabilize steam generator levels is not considered in the UFSAR analyzed event since the AF system does not have adequate flow capacity to support plant operation at full power.
50.59 REVIEW COVERSHEET FORM LS-AA-104-1001 Revision 4 Page 2 of5 Station/Unit(s): Byron & Braidwood/Units 1&2 Activity/Document Number: l/2BOA SEC-1, 1/2Bw0A SEC-I Revision Number: 114/115/112/110
Title:
Secondary Pump Trip Unit 112 The design flow capability of the AF system is based on a feedline break downstream of the feedwater isolation valves to prevent an over-pressurization of the Reactor Coolant System (RCS). During the response to the upset condition of a loss of normal feedwater at low power, it may be necessary to stabilize steam generator levels by manually adjusting AF system flow to less than that assumed in the UFSAR Chapter 15 analysis for a feedline break event. Similarly, it may be necessary to stabilize steam generator levels by manually adjusting AF system flow to greater than that assumed in the Steam Generator Tube Rupture Margin-to-Overfill analysis in UFSAR Chapter 15. The loss of feed water is a Category II event as discussed in the UFSAR Chapter 15. Additional concurrent design basis events are not postulated.
For postulated piping failures, the AF system was not evaluated/designed as a high-energy system based on guidance established in BTP MEB 3-1. Specifically, since the AF system is not operated during normal plant startup and since the fraction of time that the AF system operates within the pressure-temperature conditions specified for high-energy fluid system is less than 2 percent of the time that the system operates, the AF system was classified as a moderate-energy system. The guidance in the BTP MEB 3-1 includes amplification that systems such as the reactor decay heat removal system qualify as moderate-energy fluid systems based on limited operation at high-energy conditions; however, systems such as auxiliary feedwater systems operated during normal PWR reactor startup, hot standby, or shutdown would still qualify as high-energy fluid systems.
The procedure changes implemented under this Activity do not change the requirement to use the AF system in response to a loss of normal FW. The change is that the reactor is not tripped before starting the system for the upset condition resulting from a loss of normal FW in Mode I (reactor power <l 5% and prior to synchronization of the main generator to the grid) and Mode 2.
The AF system would still be used as before to control steam generator levels following the loss of normal FW event.
Therefore, the procedural changes would not significantly increase the time that the AF system is operated at a high-energy condition. The procedure changes do not allow use of the AF system in lieu of the normal FW source for plant startup. As documented in EC 628034 and EC 627959, the AF system is capable of supporting a steady-state reactor power level of 3% with a single AF pump in operation and 6.8% reactor power with two AF pumps in operation. Given the limited capacity of the AF system, continuing with the startup is not practical and procedure steps will require the normal FW to be restored before power ascension is continued. The changes to the procedures will direct actions following a loss of the normal FW system to stabilize steam generator levels until the normal FW system can be restored. This application is consistent with the use of the AF system following any loss of normal FW system transients, where the AF system is used to maintain decay heat removal with plant in hot standby or shutting down. Therefore, the change does not require re-evaluation of the AF system for postulated high energy pipe breaks since the AF system is operated during upset conditions only.
Summary of Conclusion for the Activity's 50.59 Review:
(Provide justification for the conclusion, including sufficient detail to recognize and understand the essential arguments leading to the conclusion. Provide more than a simple statement that a 50.59 Screening, 50.59 Evaluation, or a License Amendment Request, as applicable, is not required.)
The following is a summary of the conclusions in Evaluation 6G-19-005/BRW-E-2019-32:
The changes to the procedures used in response to loss of the normal FW system in Mode !(reactor power <15% and prior to synchronization of the main generator to the .grid) and Mode 2 do not change the frequency ofoccurrence of a loss of the normal FW system. The changes are intended to eliminate the plant transient associated with a manual reactor trip. Since the plant is maintained at normal conditions for the startup and the reactor trip eliminated, there are no additional challenges placed on the plant that could lead to accidents. In response to a loss of the normal FW system in Mode !(reactor power <15% and prior to synchronization of the main generator to the grid) and Mode 2, the AF system will be operated in a manner consistent with its design capabilities. The operator actions necessary to control reactor power in the response to the event are not complex and do not increase the likelihood of errors which could increase the frequency of reactivity events. Therefore, the change is not considered to change the frequency ofoccurrence for any accidents evaluated in the UFSAR.
Since current procedures start or verify the start of the AF system following a loss of the normal FW system and reactor trip, there is no change in the demands on the AF system and the AF system would be operated consistent with the current procedures and design basis. Therefore, there is no change in the likelihood ofa failure with AF system components, including the likelihood of pipe breaks. The proposed procedure changes do not degrade the performance of a safety system
50.59 REVIEW COVERSHEET FORM LS-AA-104- IOO 1 Revision 4 Page 3 of5 Station/Unit(s): Byron & Braidwood/Units 1&2 Activity/Document Number: l/2BOA SEC-I, 1/2BwOA SEC-1 Revision Number: 114/115/112/110
Title:
Secondary Pump Trip Unit 1/2 assumed to function in the safety analyses below the level of performance assumed in the safety analyses. The changes do not degrade the performance of the AF system such that it cannot perform its safety functions at the reactor power levels associated with the change. The subsequent restoration of normal FW does not increase the likelihood of a FW system malfunction as analyzed in the UFSAR. The operator burden associated with the proposed procedure changes do not increase the likelihood of an SSC malfunction previously evaluated in the UFSAR. A limiting duration for operating in Mode 2 with the AF system in operation is provided to prevent the degradation of steam generator tube integrity due to injecting water that is not treated with hydrazine.
The loss of normal FW flow is an accident analyzed in Chapter 15 of the UFSAR. The accident is analyzed for the bounding case with reactor power initially at I 00% of rated power. The analysis considers a reactor trip occurring on low steam generator level and an automatic start of the AF system. The UFSAR does not specifically analyze a loss of the normal FW system at low power/start-up since this condition would be bounded by the I 00% event. The use of the AF system to pre-emptively respond to stabilize steam generator levels is not considered in the UFSAR analyzed event since the AF system does not have adequate flow capacity to support operation at full power. In the 100% power event, the delay in establishing AF system flow was limiting for decay heat removal and for maintaining acceptable plant conditions. The procedure changes would manually start the AF system prior to low steam generator levels occurring and would maintain normal steam generator levels. The plant remains operating under normal parameters so there are no consequences for the loss of the normal FW system in Mode !(reactor power
<15% and prior to synchronization of the main generator to the grid) and Mode 2. If the AF system is not able to maintain steam generator levels, the operator will proactively initiate a manual reactor trip in response to the degrading level trend prior to reaching the Lo-2 reactor trip setpoint. In the absence of any operator action, the reactor will trip automatically on Lo-2 steam generator level and automatically start the AF system. However, the AF system would already be running and there would be no delays in establishing AF flow for heat removal. Since the currently analyzed loss ofnormal FW accident bounds the condition considered in this Evaluation, the dose consequences are not impacted by the proposed Activity.
The design flow capability of the AF system is based on a feedline break downstream of the feedwater isolation valves to prevent an over-pressurization of the RCS. During the response to the upset condition ofa loss of normal feedwater, it may be necessary to stabilize steam generator levels by manually adjusting AF system flow to less than assumed in the UFSAR Chapter 15 analysis for a feedline break event. Similarly, it may be necessary to stabilize steam generator levels by manually adjusting AF system flow to greater than that assumed in the Steam Generator Tube Rupture Margin-to-Overfill analysis in UFSAR Chapter 15. However, these design basis events are not postulated concurrent with the upset condition of a loss of nonnal feedwater during unit startup and the dose consequences of these events are not affected.
Therefore, the changes in the procedures to manually start the AF system prior to tripping the reactor on a loss of the normal FW system during unit startup in Mode l(reactor power <15% and prior to synchronization of the main generator to the grid) and Mode 2 will not change the consequence of the loss of normal FW, feed line break, or steam generator tube rupture accidents as evaluated in the UFSAR.
The procedure changes implemented under this Activity do not change the requirement to use the AF system in response to a loss of normal FW, only that the reactor is not tripped before starting the system for the upset condition resulting from a loss of the normal FW system. Therefore, the AF system would still be used as before to control steam generator levels following the loss ofnormal FW event. Therefore, the procedural changes would not significantly increase the time that the AF system is operated at a high-energy condition. The procedure changes do not allow use of the AF system in lieu of the normal FW source for plant startup. Given the limited capacity of the AF system, continuing with the startup is not practical and procedure steps will require the nonnal FW system to be restored before power ascension is continued and will not change use of the AF system for normal startup. The changes to the procedures will direct actions following a loss of the normal FW system to stabilize steam generator levels until the normal FW system can be restored. This application is consistent with the use of the AF system following any loss of the normal FW system transients, where the AF system is used to maintain decay heat removal with plant in hot standby or shutting down. The operation of the AF system in response to the loss of the normal FW system in Mode l (reactor power <15% and prior to synchronization of the main generator to the grid) and Mode 2 does not introduce any new system failure modes and the dose consequences of any accident that the AF system is designed to mitigate is unchanged.
Therefore, the changes implemented under the proposed Activity will not result in more than a minimal increase in the dose consequences of a malfunction of an SSC important to safety previously evaluated in the UFSAR
50.59 REVIEW COVERSHEET FORM LS-AA-104-1001 Revision 4 Page 4 of 5 Station/Unit(s): Byron & Braidwood/Units 1&2 Activity/Document Number: 1/2BOA SEC-I, 1/2BwOA SEC-I Revision Number: 114/115/112/110
Title:
Secondary Pump Trip Unit 1/2 The proposed procedure changes impact the abnormal operating procedures used for a loss of the normal FW system events.
The loss of the normal FW system is already an accident considered in the UFSAR. The accident is analyzed for the bounding case at an initial power level of 100% of rated power and does not specifically analyze a loss of the normal FW system at low power/start-up since this condition would be bounded by the I 00% event. The change in the procedures to manually start the AF system prior to tripping the reactor on loss of the normal FW system in Mode I (reactor power <15%
and prior to synchronization of the main generator to the grid) and Mode 2 is in response to the loss of the normal FW system to stabilize and maintain steam generator levels and does not cause the loss of the normal FW system.
The procedural strategy for a loss of normal feedwater utilizes the AF system to maintain steam generator levels and a heat sink for the reactor. At low reactor power levels, in particular with the reactor critical and power less than the point of adding heat (POAH), the injection of colder water into the steam generators from the Condensate Storage Tank has the potential to result in an excessive RCS cooldown and reactivity transient. To address this, the procedures establish manual reactor trip criteria based on startup rate and RCS temperature limitations consistent with existing guidance. In addition, prior to the start of the first AF pump, the operators will be cautioned to limit the injection flow rate to ensure that an excessive RCS cooldown does not occur. The concern for excessive cooldown with reactor power greater than the POAH is minimal based on the AF system injection flowrates being less than or equal to the normal FW system flowrate prior to its loss and the differential temperatures between the AF and FW systems. At very low power levels, the flow and temperature difference is insignificant. At higher reactor power levels (<15%) and for extended unloaded turbine operation, the temperature difference increases, but the AF system injection flow rates are significantly less than the pre-event normal FW system flow rates such that there is no RCS cooldown. In addition, due to the initial loss of normal FW, RCS temperature increases until either the condenser steam dump system responds or AF system injection turns temperature. Therefore, the proposed procedural strategy, the operational limitations established for a manual reactor trip, and the plant's inherent response results in a minimal impact on RCS temperature and core reactivity when the AF system is used to recover from a loss of normal FW during startup with reactor power less than 15% and prior to synchronizing the main generator to the grid including Mode 2 operation above and below the POAH.
Therefore, the procedure changes do not create a new accident from those already considered in the UFSAR.
For postulated piping failures, the AF system was not classified as a high-energy system based on guidance established in BTP MEB 3-1. Specifically, since the fraction of time that the AF system operates within the pressure-temperature conditions specified for high-energy fluid system is less than 2 percent of the time that the system operates, the AF system was evaluated as a moderate-energy system. The guidance in the BTP MEB 3-1 includes amplification that systems such as the reactor decay heat removal system qualify as moderate-energy fluid systems based on limited operation at high-energy conditions; however, systems such as auxiliary feedwater systems operated during PWR reactor startup, hot standby, or shutdown would still qualify as high-energy fluid systems. The proposed procedure changes direct actions following a loss of the normal FW system in Mode 1(reactor power <15% reactor power) and Mode 2, an upset condition, to stabilize steam generator levels until the normal FW system can be restored in Mode 2. The procedure changes do not allow use of the AF system in lieu of the normal FW source for normal plant startup. Given the limited capacity of the AF system, continuing the startup is not practical and procedure steps will require the normal FW system be established before power ascension is continued beyond Mode 2 and will not change use of the AF system for normal start-up. This application is consistent with the use of the AF system following any loss of the normal FW systems, where the AF system is used to maintain decay heat removal with the plant in hot standby and shutdown and does not create any new failure modes for the system. Therefore, the changes do not require re-evaluation of the AF system for postulated high energy pipe breaks since the procedure changes are not altering the design basis status of the AF system as a moderate energy system and the changes will not create the possibility for a malfunction of an SSC important to safety with a different result than any previously evaluated in UFSAR. Furthermore, no credible operator errors could result in reactivity addition rates that exceed those evaluated in Section 15.4 of the UFSAR. Therefore, the proposed procedure changes do not produce a different result or change the consequences of reactivity anomalies at low reactor power levels as described in the UFSAR.
The changes to the procedures will modify the response to the actions taken in response to loss of the normal FW system during startup in Mode !(reactor power <15% and prior to synchronization of the main generator to the grid) and Mode 2.
These changes do not alter any fission product barriers. In addition, the changes to the procedures did not require a change to any of the analyses used to establish the design basis limits for fission product barriers. Therefore, the proposed Activity does not result in a design basis limit for a fission product barrier as described in the UFSAR being exceeded or altered.
50.59 REVIEW COVERSHEET FORM LS-AA- !04-100 I Revision 4 Page 5 of5 Station/Unit(s): Byron & Braidwood/Units 1&2 Activity/Document Number: l/2BOA SEC-I, 1/2BwOA SEC-1 Revision Number: 114/115/112/110
Title:
Secondary Pump Trip Unit 1/2 No design basis analyses were revised for the procedure changes. Postulated line breaks in the AF system were originally evaluated as a moderate-energy system since the operation of AF system at high-energy condition is less than 2 percent of the time that the system operates. The procedure changes do not result in a deviation from the guidance in the BTP MEB 3-1 and re-evaluation of the AF system for high-energy pipe breaks is not required. Therefore, the proposed Activity does not result in a departure from a method of evaluation described in the UFSAR used in establishing the design bases or in the safety analyses.
Based on the above, 50.59 Evaluation 6G- I 9-005JBRW-E-2019-32 concluded that the Activity can be implemented per plant procedures without obtaining a License Amendment.
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D Applicability Review D 50.59 Screening 50.59 Screening No. Rev.
~ 50.59 Evaluation 50.59 Evaluation No. 6G-19-005 Rev. 0 BRW-E-2019-32 0 See LS-AA- I 04, Section 5, Documentation, for record retention requirements for this and all other 50.59 forms associated with the Activity.