ML24348A093
| ML24348A093 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 12/13/2024 |
| From: | Constellation Energy Generation |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML24348A091 | List: |
| References | |
| BW240069 | |
| Download: ML24348A093 (1) | |
Text
ATTACHMENT Braidwood Station 10 CFR 50.59 Summary Report Evaluation No.
Revision Title Approved on 0
04/07/2019 BRW-E-2019-91 Upgrade Existing AVRs with Digital ABB Unitrol Model 1
1/30/2020 BRW-E-2021-37 0
Ovation Automatic Runback of Main Turbine Upon Loss of Multiple 7/28/2022 Feed Water Pumps BRW-E-2022-16 0
U1 and U2 Turbine Load Reject Time Delay Modification 8/11/2022 BRW-E-2022-25 0
TRM Change and Procedure Revisions for change in TSR 3.1.b.4 9/12/2022 boration flow BRW-E-2022-22 0
Implementation of Braidwood BOL MTC Elimination 9/23/2022 BRW-E-2022-42 0
Updates B&B Chapter 15 Dose Analyses 4/17/2023 BRW-E-2023-001 0
Thermal Overload Protection Surveillance Interval Change for U1 4/13/2023 and U2 SX001A Valves BRW-E-2022-40 0
Reclassify ASME Ill FP PipingNalves/Components in Seismically 12/19/2022 Qualified Areas BRW-E-2024-001 0
Feedwater Regulating Valve Dual Positioner - Unit 1 / Unit 2 1/26/2024
50.59 REVIEW COVERSHEET FORM Station/Unit(s): Braidwood Units I and 2 EC 624259, Rev. 000.1 Page 1 of 2 LS-AA-104-1001 Revision 4 Page l of2 Activity/Document Number: EC 624259 {U2) and EC 624258 (Ul) / DRP 18-002 {U2) and DRP 18-001 (Ul)
Revision Number: 000 and 000 / NA for DRPs
Title:
Upgrade Existing Unit 2( l) A VR with Digital ABB Unitrol Model 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 l O CFR 50.59( d)(2).
Description of Activity:
(Provide a brief, concise description of what the proposed activity involves.)
The proposed activity replaces the existing Westinghouse/ Cutler-Hammer solid-state main generator automatic voltage regulator (A VR) with a new Asea Brown Boveri (ABB) digital A VR, Model Unitrol 6000 Medium, which performs the same critical functions in the overall excitation system as the existing A VR. In place of the single-channel design of the existing A VR, the replacement A VR uses a two-channel design to improve reliability. The AVR will continue to receive power from the permanent magnet generator (PMG). In addition, two independent 480 Vac power feeds are introduced to support the dual-channel design to enhance reliability. The PMG will supply one channel; one of the 480 Vac power feeds will supply the other channel, while the second 480 Vac will provide a backup source of power that can be manually aligned to supply the first channel (normally supplied by the PMG). The new A VR also includes a power system stabilizer (PSS) to enhance grid system stability. The new design modifies the main control room controls and provides an excitation control terminal (ECT) touchscreen panel at the A VR for local control and monitoring. Existing power sources are re-utilized where necessary but new 480 Vac power sources and an additional 125 Vdc feed are required for this modification. The new Unit l and Unit 2 A VR cabinets will be relocated approximately fifty (50) feet north-east to a location with lower sustained floor vibration levels.
DRP 18-002 (U2) and DRP 18-00 I (U l) are issued in conjunction with this activity to update the main exciter description in UFSAR Section 10.2.2.2 to indicate that the generator exciter will be controlled by a digital AVR equipped with two redundant channels.
Reason for Activity:
(Discuss why the proposed activity is being performed.)
The existing A VR is obsolete and is no longer supported by the manufacturer. The new A VR is a state-of-the-art system offering improved reliability.
Effect of Activity:
(Discuss how the activity impacts plant operations, design bases, or safety analyses described in the UFSAR.)
Instead of separate base adjuster and voltage adjuster controls, there will be one control to adjust voltage. The A VR will automatically line up the voltage of the automatic and manual control modes. The new power system stabilizer function will be actuated automatically (at approximately 10% power). Control board indications, switches, and alarms will be modified to reflect the new operational requirements of the replacement A VR system. A new excitation controls terminal (ECT) will be installed at the A VR cabinet -- this terminal can only be operated under strict administrative controls.
There is no impact on the design bases or safety analyses described in the UFSAR.
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.)
Failures in the A VR system could result in a turbine trip and challenges to the offsite power system. The replacement A VR system includes digital hardware and software, and significant changes to the human-machine interface. Since there is a potential to adversely affect UFSAR-described design functions, these aspects of the proposed activity were "screened in" for further assessment under the 50.59 Evaluation process.
Like the existing A VR, failures in the replacement A VR could result in a turbine trip and challenges to the offsite power system.
50.59 REVIEW COVERSHEET FORM Station/Unit(s): Braidwood Units l and 2 EC 624259, Rev. 000.1 Page 2 of 2 LS-AA-104-1001 Revision4 Page 2 of2 Activity/Document Number: EC 624259 (U2) and EC 624258 (Ul) / DRP 18-002 (U2) and DRP 18-001 (Ul)
Revision Number: 000 and 000 / NA for DRPs
Title:
Upgrade Existing Unit 2( l) A YR with Digital ABB Unitrol Model The failure modes and effects of the replacement A YR are bounded by those of the existing A YR. The replacement A YR is a state-of-the art system widely used in the industry and is provided by a vendor with considerable experience. The replacement A YR incorporates a dual-channel design in place of the single-channel design of the existing A YR, and additional sources of power have been provided by the station to support the dual-channel design. The operator interface with the A YR system has been simplified, and existing station practices ensure that the operators are familiar with the replacement system and with the required interface with the system.
The improvements in reliability provide assurance that there is no more than a minimal increase in the frequency of occurrence of an accident or in the likelihood of a malfunction previously evaluated in the UFSAR. Previous analyses of events which could result from an A YR malfunction remain bounding; therefore, the radiological consequences of accidents or malfunctions are not affected. The A YR system does not interface with any Ovation-based control systems, so there is no potential for a common-cause failure affecting multiple plant systems that could create the possibility for an accident or malfunction not previously evaluated.
The proposed activity does not affect a design basis limit for a fission product barrier. Supporting analyses/evaluations for this activity have been performed in a manner consistent with standard industry practices and consistent with the evaluation requirements / methodologies described in the UFSAR. This activity does not involve a test or experiment not described in the UFSAR. The proposed activity does not affect the Technical Specifications or the Facility Operating License.
Attachments:
Attach all 50.59 Review forms completed, as appropriate.
Forms Attached: (Check all that apply.)
Applicability Review 50.59 Screening 50.59 Evaluation 50.59 Screening No.
50.59 Evaluation No.
BRW-S-2019-90 Rev.
0 -------
BRW-E-2019-91 Rev.
0 -------
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 Station/Unit(s): Braidwood Units I and 2 EC 624258, Rev. 000.1 Page 1 of 2 LS-AA-I 04-100 I Revision 4 Page I of2 Activity/Document Number: EC 624259 (U2) and EC 624258 (Ul) / DRP 18-002 (U2) and DRP 18-001 (Ul)
Revision Number: 001 and 000 / NA for DRPs
Title:
Upgrade Existing Unit 2( I) A YR with Digital ABB Unitrol Model 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( d)(2).
Description of Activity:
(Provide a brief, concise description of what the proposed activity involves.)
The proposed activity replaces the existing Westinghouse/ Cutler-Hammer solid-state main generator automatic voltage regulator (A YR) with a new Asea Brown Boveri (ABB) digital A YR, Model Unitrol 6000 Medium, which performs the same critical functions in the overall excitation system as the existing A YR. In place of the single-channel design of the existing A YR, the replacement A YR uses a two-channel design to improve reliability. The A YR will continue to receive power from the permanent magnet generator (PMG). In addition, two independent 480 Yac power feeds are introduced to support the dual-channel design to enhance reliability. The PMG will supply one channel; one of the 480 Yac power feeds will supply the other channel, while the second 480 Yac will provide a backup source of power that can be manually aligned to supply the first channel (normally supplied by the PMG). The new A YR also includes a power system stabilizer (PSS) to enhance grid system stability. The new design modifies the main control room controls and provides an excitation control terminal (ECT) touchscreen panel at the A YR for local control and monitoring. Existing power sources are re-utilized where necessary but new 480 Yac power sources and an additional 125 Y de feed are required for this modification. The new Unit I and Unit 2 A YR cabinets will be relocated approximately fifty (50) feet north-east to a location with lower sustained floor vibration levels.
The proposed activity includes a data connection between the A YR and the Plant Process Computer to improve monitoring, trending, and diagnostic capabilities.
The new A YR has the ability to detect a reverse power condition and initiate a generator trip signal after the appropriate time delay. The proposed activity will abandon the existing reverse power protection provided by relay PR! I and replace it with the feature provided by the new A YR.
The existing differential pressure switch logic associated with sensing and initiating a loss of stator water cooling generator trip signal will be processed by the new A YR. This feature improves the ability to detect the failure or actuation of a single differential pressure switch in the two-of-three trip logic for a loss of stator water cooling.
DRP 18-002 (U2) and DRP 18-00 I (U I) are issued in conjunction with this activity to update the main exciter description in UFSAR Section I 0.2.2.2 to indicate that the generator exciter will be controlled by a digital A YR equipped with two redundant channels.
Reason for Activity:
(Discuss why the proposed activity is being performed.)
The existing A YR is obsolete and is no longer supported by the manufacturer. The new A YR is a state-of-the-art system offering improved reliability. In addition, the proposed change addresses reliability issues with the existing General Electric GGP reverse power relays used in the generator protection scheme. The proposed change to the loss of stator water cooling logic provides the ability to identify a latent vulnerability associated with a single differential pressure switch failure or actuation.
Effect of Activity:
(Discuss how the activity impacts plant operations, design bases, or safety analyses described in the UFSAR.)
Instead of separate base adjuster and voltage adjuster controls, there will be one control to adjust voltage. The A YR will automatically line up the voltage of the automatic and manual control modes. The new power system stabilizer function will be actuated automatically (at approximately I 0% power). Control board indications, switches, and alarms will be modified to reflect the new operational requirements of the replacement A YR system. A new excitation controls terminal (ECT) will be installed at the A YR cabinet~ this terminal can only be operated under strict administrative controls.
50.59 REVIEW COVERSHEET FORM Station/Unit(s): Braidwood Units I and 2 EC 624258, Rev. 000.1 Page 2 of2 LS-AA-I 04-100 I Revision 4 Page 2 of2 Activity/Document Number: EC 624259 (U2) and EC 624258 (UI) / DRP 18-002 (U2) and DRP 18-001 (UI)
Revision Number: 00 I and 000 / NA for DRPs
Title:
Upgrade Existing Unit 2( I) A YR with Digital ABB Unitrol Model The proposed changes to the reverse power and loss of stator water cooling generator trip functions improve the reliability of these trip features without changing their function.
There is no impact on the design bases or safety analyses described in the UFSAR.
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.)
Failures in the A YR system and the generator protective functions modified by the proposed activity utilizing the capabilities of the new A YR system could result in a turbine trip and challenges to the offsite power system. The replacement A YR system includes digital hardware and software, and significant changes to the human-machine interface. Since there is a potential to adversely affect UFSAR-described design functions, these aspects of the proposed activity were "screened in" for further assessment under the 50.59 Evaluation process.
Like the existing A YR, failures in the replacement A YR and malfunctions of the reverse power and loss of stator cooling water generator trip functions could result in a turbine trip and challenges to the offsite power system. The failure modes and effects of the replacement A YR are bounded by those of the existing A YR. The replacement A YR is a state-of-the art system widely used in the industry and is provided by a vendor with considerable experience. The replacement A YR incorporates a dual-channel design in place of the single-channel design of the existing A YR, and additional sources of power have been provided by the station to support the dual-channel design. The operator interface with the A YR system has been simplified, and existing station practices ensure that the operators are familiar with the replacement system and with the required interface with the system.
The improvements in reliability provide assurance that there is no more than a minimal increase in the frequency of occurrence of an accident or in the likelihood of a malfunction previously evaluated in the UFSAR. Previous analyses of events which could result from an A YR malfunction or generator trips remain bounding; therefore, the radiological consequences of accidents or malfunctions are not affected. The A YR system does not interface with any Ovation-based control systems, so there is no potential for a common-cause failure affecting multiple plant systems that could create the possibility for an accident or malfunction not previously evaluated.
The proposed activity does not affect a design basis limit for a fission product barrier. Supporting analyses/evaluations for this activity have been performed in a manner consistent with standard industry practices and consistent with the evaluation requirements I methodologies described in the UFSAR. This activity does not involve a test or experiment not described in the UFSAR. The proposed activity does not affect the Technical Specifications or the Facility Operating License.
Attachments:
Attach all 50.59 Review forms completed, as appropriate.
Forms Attached: (Check all that apply.)
D Applicability Review 50.59 Screening 50.59 Evaluation 50.59 Screening No.
50.59 Evaluation No.
BRW-S-2019-90 BRW-E-2019-91 Rev.
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 Station/Unit(s): Braidwood Station Units I & 2 LS-AA-104-1001 Revision 4 Page I of3 Activity/Document Number: Engineering Changes 633143, 630851 D RPs I 9-065,19-066 Revision Number: 00
Title:
Ovation Automatic Runback of Main Turbine Upon Loss of Multiple Feed Water Pumps 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.)
The proposed activity installs a software enhancement to the Ovation Distributed Control System (DCS) which monitors the status of the main feed water pumps, detects when there is only one feed water pump in operation, and automatically initiates the existing, pre-programmed "CD/FW" main turbine load reduction (runback). The runback rate and target power level are not changed, as sensitivity analysis using the Westinghouse software model of Braidwood Station units 1 and 2 demonstrates the existing 20%/min to 700MWe to be adequate. The sensitivity analysis and testing results are used to validate the feed pump status detection logic, the automatic run back permissives, and evaluate sensitivities to lowering steam generator levels when reduced to a single feed water pump. Based on the modeling results obtained from the sensitivity studies associated with steam generator levels, additional software enhancements are made to increase the time to reaching the low steam generator administrative and protective reactor trip values. The additional control system enhancements include automatically starting the standby condensate/condensate booster (CD/CB) pump, limiting the full-open position (short stroking) of certain main feed water regulating valves (MFR Vs) during the CD/FW run back, and temporarily reducing the steam generator level setpoint during transient recovery (unit 2 only). These additional automatic control actions become enabled based on an active CD/FW runback signal and coincident with other plant inputs needed to satisfy the control logic. These additional plant inputs are designed to represent a loss of normal feed water flow condition such that the automatic start of the standby CD/CB pump and short stroking the MFR Vs does not occur when conditions do not require them. The automated CD/FW runback function can be monitored, enabled, or disabled via the control system graphics. A new, dedicated alarm tile is provided in the main control room to alert the operator to the "CD/FW AUTO RUNBACK ACTIVE".
Reason for Activity:
(Discuss why the proposed activity is being performed.)
Operating experience has shown that a significant number of reactor trips occur because of the loss (trip or unavailability) of main feed water pumps. The Braidwood Station design includes three main feed water pumps, two turbine-driven feed pumps (TDFP) and one motor-driven feed pump (MDFP). The normal operating lineup is two TDFPs in operation and providing the necessary flow to maintain steam generator levels, with the MDFP in standby and available to automatically start in the event of a TDFP trip (ref. § 10.4.7.2). The MDFP automatic start was previously a manual action required by the operator and was automated as part of the NSSS and BOP control system upgrade implemented via engineering changes 404360 and 404363.
This was implemented as a system enhancement to improve the success of keeping a unit online after a TDFP trip. However, if the MDFP fails to start or is unavailable (ref. IR 4144070), it is necessary to reduce reactor power to a level such that the one remaining feed water pump can provide the necessary flow to maintain adequate steam generator levels without a reactor trip.
The evolution of reducing power to the capacity of a single feed water pump is time-sensitive and is currently dependent on manual operator actions. To eliminate the variability of the manual operator actions and improve likelihood of maintaining unit operation under this scenario, the proposed activity automates certain control actions by utilizing the DCS to monitor feed water pump status. When reduced to a single feedwater pump, the control system initiates a turbine load reduction and automatically starts the standby CD/CB pump, thereby improving the outcome of certain loss of normal feedwater flow events.
Effect of Activity:
(Discuss how the activity impacts plant operations, design bases, or safety analyses described in the UFSAR.)
50.59 REVIEW COVERSHEET FORM Station/Unit(s): Braidwood Station Units 1 & 2 LS-AA-104-1001 Revision 4 Page 2 of3 Activity/Document Number: Engineering Changes 633143, 630851 DRPs19-065, 19-066 Revision Number: 00
Title:
Ovation Automatic Runback of Main Turbine Upon Loss of Multiple Feed Water Pumps The proposed activity impacts plant operations by implementing functional changes to the DCS which automates a main turbine load reduction (and subsequently a reactor power reduction), automatically starts the standby CD/CB pump, and limits the full-open position of certain MFRV s when the plant is reduced to a single feedwater pump in operation and loss of normal feedwater flow conditions are detected. Currently, the evolution to detect, assess, and act when plant conditions are reduced to a single feed pump require manually executed processes by the operator which can delay the initiation of a runback. The proposed activity reduces the time between detection of a single feed pump in service and actuation of turbine load reduction, improving the success of maintaining plant operation and eliminating the variability of manual operator actions. The effect of the activity improves the outcome of certain condition II events (i.e., reactor trip on loss of normal feedwater) by taking more immediate control actions to maintain the plant within conditions which are accommodated with margin to automatic or manual protective actions ( condition r events).
Maintaining plant operation within condition I events (normal operational transients, particularly regarding turbine load reductions) does not utilize or control SSCs in a manner inconsistent with descriptions in the UFSAR nor impact the plant design bases. The safety analyses, as described and previously evaluated in the UFSAR, remain bounding.
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.)
50.59 screening questions 1 and 2 were answered "yes" since the proposed activity changes the method of controlling SSCs described in the UFSAR from manual to automatic and requires change to associated station procedures that fundamentally alter existing means of performing or controlling design functions. As this constitutes an adverse change to the facility, a 50.59 Evaluation was performed to address these conditions.
The 50.59 screening determined that, with the exception of conditions noted above, implementation of this activity does not adversely affect any UFSAR described functions, does not involve a change to the methods of evaluation used to establish the design bases or safety analyses as described in the UFSAR, and does not result in system operation outside the referenced bounds of the design or inconsistent with the analyses or inconsistent with the description in the UFSAR. Implementation of this activity does not require a change to the Technical Specifications or Facility Operating License.
The 50.59 evaluation determined that implementation of this activity will exhibit a sufficiently low likelihood of failure and thus will not result in more than a minimal increase of the frequency of an accident, or likelihood of occurrence of a malfunction of an SSC, nor does it create a possibility for an accident of a different type, nor does it create the possibility for a malfunction of an SSC important to safety with a different result from any previously evaluated.
The proposed activity involves control system software enhancements which automate previously manual actions in order to provide a reduction in response time for feed water pump failures and decrease the frequency of unit trips. Facility changes which convert manual actions to automatic are considered adverse in the screening process and must be evaluated to ensure the proposed change remains bounded by the design and licensing bases of the plant. Furthermore, the use of software to perform system functions requires special considerations for ensuring reliability and dependability that is different than that typically provided by hardware or analog solutions.
The changes associated with the proposed activity were reviewed and further evaluated to determine that the design and licensing basis of the plant remains bounding. The evaluation includes:
- Changing main turbine load reduction from manual to automatic
- Changing the start of the fourth (standby) condensate/condensate booster pump from manual to automatic
- Procedure changes that fundamentally alter existing means of performing or controlling design functions
50.59 REVIEW COVERSHEET FORM Station/Unit(s): Braidwood Station Units l & 2 LS-AA-104-100 I Revision 4 Page 3 of3 Activity/Document Number: Engineering Changes 633143, 630851 DRPs19-065, 19-066 Revision Number: 00
Title:
Ovation Automatic Run back of Main Turbine Upon Loss of Multiple Feed Water Pumps These elements are interdependent by means of control system logic changes to achieve a common design function, and therefore evaluated collectively. While the objective of the proposed activity improves the plant response to feedwater pump failures by precluding a unit trip, the postulated failures in the control system are assumed to result in a reactor trip. The safety system equipment continues to operate in the same manner regardless of the initiating event which caused the trip and remain capable of ensuring that fission product barrier limits are not exceeded. The existing safety analysis for accidents that could be initiated by failures in the control systems remain bounding and no radiological consequences beyond that which has been previously evaluated would occur.
Thorough testing of the software and analysis of the new failure modes and their effects ensures the design is adequate by testing system functionality to the extent possible prior to installation in the plant. The use of modeling allows testing to normal and abnormal conditions that could not practically be performed in the plant. Likelihood of failures in the control system is sufficiently reduced using signal redundancy, permissives, hardware redundancy, a quality design and testing process, operating experience in similar applications, and integration on the existing DCS.
The means by which the operator interfaces with the DCS remains essentially unchanged. By automating control tasks which were previously performed by the operator through prompt response actions, the proposed activity reduces operator burden and allows for more focus on situational assessment and response. Affected procedures remain applicable but revised based on the new automatic control response and validated as part of the design change activities. Operators are key stakeholders in the design change and ensure timing, indication, data presentation, and usability are not adversely affected. Existing controls remain in place, at the control board or the control system graphics, that allow the operator to manually initiate a runback as well as manually start the standby CD/CB pump.
There is no adverse change to an element of an UFSAR-described evaluation methodology, or use of an alternative methodology, that is used in establishing the design bases or used in the safety analyses.
Therefore, based on this 50.59 screening & evaluation, implementation of this activity does not require a License Amendment Request (LAR) and may be implemented under the governing Station procedures.
Attachments:
Attach all 50.59 Review forms completed, as appropriate.
Forms Attached: (Check all that apply.)
Applicability Review 50.59 Screening 50.59 Evaluation 50.59 Screening No.
50.59 Evaluation No.
BRW-S-2021-36 Rev.
00 BRW-E-2021-37 Rev.
00 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 Station/Unit(s):
Braidwood Station/ Units 1 and 2 Activity /Document Number: =-EC-=--=6=3-=-6-'--7-'--70;::.._ ______________ _
Title:
Ul and U2 Turbine Load Reject Trip Time Delay LS-AA-104-1001 Revision 4 Page 1 of2 Revision Number: ~O __ _
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:
The Turbine Load Reject protection scheme is used to protect the turbine from significant overspeed following a rapid loss of load if the Main Generator is unexpectedly disconnected from the switchyard/transmission system. The Load Reject scheme uses position indicating limit switches associated with breakers and disconnects to sense when the Main Generator becomes separated from the switchyard. The current load reject protection logic directly actuates the 86G IA Turbine Trip Lockout Relay to trip the turbine control valves. The subject design change will install an interposing time delay relay (TOR) to the existing Turbine Load Reject protection scheme. The new time delay relay will be actuated by the existing load reject logic and then actuate the 86G 1 A Lockout Relay after a short time delay.
Reason for Activity:
A lightning strike in the switchyard caused a false, spurious indication that the Bus Tie (BT) 1-3, 1-8 and 7-8 breakers had opened. BT 1-8 and BT 7-8 breakers are the generator output breakers and are inputs to the load reject logic. Opening of generator output breakers indicates the main generator has separated from the transmission system. The lightning induced false simultaneous open indication for BT 1-8 and BT 7-8 initiated a load reject turbine trip even though the main generator had remained connected to the transmission system through the event. The lightning strike was significantly more energetic and closer than previous strikes in the area. Since it is difficult to bound and design protection from lightning strikes, the purpose of the time delay is to block potential spurious signals from causing an unnecessary turbine trip. Since the duration of the transient from a lightning strike is short, a short time delay can block spurious trips without compromising turbine overspeed protection.
Effect of Activity:
Load rejection is a sudden loss of load on the main generator and will result in a rapid increase in turbine speed until the turbine control valves are closed. In addition to the load reject logic there are other redundant trips that detect the loss of load or resultant overspeed condition and trip the turbine by closing the control valves. However, the load reject logic responds to the cause of the loss of load and trips the turbine preemptively before the overs peed detection logic responds, which limits the overshoot that can occur. The time delay added by the design change to the logic will delay the turbine trip slightly, but the resultant overspeed is still within the rating of the turbine.
Loss of Load is also a design basis accident discussed in the UFSAR (section 15.2.2). The Turbine Trip event (UFSAR Section 15.2.3) bounds the Loss of Load event. The analyzed event is initiated by a turbine trip, which results in a rapid closure of the turbine stop valves. Further, the reactor trip on turbine trip signal is not credited in either analysis so the timing of the closure would not impact the subsequent analysis. A specific actuation signal is not modeled since the turbine trip and not the cause of the turbine trip is the initiating event in the bounding analysis. Since the turbine trip is the initiating event, the delay added to the specific Load Reject scheme does not change the Loss of External Load analysis and remains bounded by Turbine Trips in section 15.2.3. The proposed modification to delay the turbine trip signal from the Load Reject will not impact the analyses discussed in UFSAR Sections 15.2.3 through 15.2.5.
50.59 REVIEW COVERSHEET FORM Station/Unit(s):
Braidwood Station/ Units 1 and 2 Activity/Document Number: ~E~C~6~3~6_77~0~---------------
LS-AA-104-1001 Revision 4 Page 2 of2 Revision Number: ~O __ _
Title:
"-U=l--=a"'-n=d'--U""'2=--'-T=u-'-'rb=i""'-n-=-e--=L=o=a=d--'-R=e;.,,.je=c=-=t-'T--'-r'""ip'-T-'-'i"'"'m..;..;e:....=c..D-=-e=la'-'-y _____________________ _
Summary of Conclusion for the Activity's 50.59 Review:
Based on the following, the design change may be implemented without prior NRC approval.
The modification adds a time delay relay to delay a turbine trip specifically from the Load Reject logic. The Load Reject is a sudden loss of load on the main generator if it becomes disconnected from the switchyard/transmission system. The Load Reject is one possible initiator of the Loss of External Load accident in the UFSAR. The Load Reject logic responds to a loss of load but is not an initiator of a loss of load. Therefore, a change to the Load Reject protection scheme does not change the likelihood of a Loss of External Load. The Loss of External Load accident is bounded by the Turbine Trip events described in UFSAR section 15.2.3. The Turbine Trip events continue to bound the Loss of External Load with the addition of the time delay to load reject. Therefore, the consequence of the Loss of External Load is not altered by the new time delay to the Load Reject logic.
The load reject logic responds to the cause of the loss ofload to trip the turbine preemptively before the overspeed detection logic responds. Therefore, the Load Reject logic responds to a cause of loss of load on the turbine to limit the overshoot that can occur when the generator is disconnected from the transmission system. However, the logic is not an initiator to the loss of load. Therefore, the likelihood of a turbine overspeed from loss of load is not affected. The time delay added by the design change to the logic will delay the turbine trip slightly, but the resultant overspeed is still within the rating of the turbine. Therefore, there is no change to the consequence of the turbine overspeed following load reject.
There are no new accidents created by the modification. The failure effects associated with the new relay are bounded by the existing failure effects and results associated with the existing logic. No fission product barriers are altered or exceeded.
There are no changes to methodologies associated with the analyses impacted by the modification.
Attachments:
Attach all 50.59 Review forms completed, as appropriate.
Forms Attached: (Check all that apply.)
D Applicability Review 50.59 Screening 50.59 Evaluation 50.59 Screening No.
50.59 Evaluation No.
BRW-E-2022-16 Rev.
Rev.
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.
50.59 REVIEW COVERSHEET FORM Station/Unit(s): Braidwood Station Units I & 2, Byron Units I & 2 LS-AA-I 04-100 I Revision 4 Page 1 of2 Activity/Document Number: TRM Change #22-008 (BRW), #22-008 (BYR)
Revision Number: 00 DRP #20-036 (BRW), #20-036 (BYR), Procedures in Activity Description
Title:
TRM Change and Procedure Revisions for change in TSR 3.1.b.4 boration flow 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 l O CFR 50.59(d)(2).
Description of Activity:
(Provide a brief, concise description of what the proposed activity involves.)
The Nuclear Safety Requirements for the Boric Acid Transfer Pump boration flowpath are to provide sufficient boron to the core to restore shutdown margin and offset Xenon burnout during transients. This is accomplished through the BORDER analysis performed by Westinghouse, which compares the maximum Xenon burnout worth to the boration flow value defined in TSR 3.1.b.4. The previous value of 30 gpm will be adjusted to 33 gpm based on changes in Xenon worth determined for future core cycles. This requires changes in both the surveillance for measurement and comparison to the analysis in addition to changes in Operating procedures that use the boration flowrate in off-normal conditions.
The changes under this screening and evaluation include: revision ofTSR 3.1.b.4, UFSAR 9.3.4.1.2.5 referenced flowrate, and revisions to the following procedures consistent with the change:
lBwOS TRM 3.1.b.4 Rev 11 EMERGENCY BORATION FLOW RATE VERIFICATION 2BwOS TRM 3.1.b.4 lBwOL 3.1.1 2BwOL 3.1.1 lBwOL 3.9.1 2BwOL 3.9.1 Rev 11 EMERGENCY BORATION FLOW RATE VERIFICATION LCOAR SHUTDOWN MARGIN (SOM) TECH SPEC LCO 3.1.1 LCOAR SHUTDOWN MARGIN (SOM) TECH SPEC LCO 3.1.1 LCOAR BORON CONCENTRATION TECH SPEC LCO 3.9.1 LCOAR BORON CONCENTRATION TECH SPEC LCO 3.9.1 lBwOL TRM 3.1.h TECHNICAL REQUIREMENTS MANUAL (TRM) LCOAR SHUTDOWN MARGIN (SOM) -
MODE 1 AND MODE 2 WITH KEFF :?:1.0 TRM TLCO 3.1.h 2BwOL TRM 3.1.h TECHNICAL REQUIREMENTS MANUAL (TRM) LCOAR SHUTDOWN MARGIN (SOM) -
MODE 1 AND MODE 2 WITH KEFF :?:1.0 TRM TLCO 3.1.h lBwOL TRM 3.1.i TECHNICAL REQUIREMENTS MANUAL (TRM) LCOAR SHUTDOWN MARGIN (SOM) -
MODE 5 TRM TLCO 3.1.i 2BwOL TRM 3.1.i TECHNICAL REQUIREMENTS MANUAL (TRM) LCOAR SHUTDOWN MARGIN (SOM) -
MODE 5 TRM TLCO 3.1.i lBwFR-S.1 RESPONSE TO NUCLEAR POWER GENERATION/ATWS UNIT 1 2BwFR-S.1 RESPONSE TO NUCLEAR POWER GENERATION/ATWS UNIT 2 lBwOA PRl-2 EMERGENCY BORATION UNIT 1 2BwOA PRl-2 BwOP CV-41 lBFR S.l lBOA PRl-2 EMERGENCY BORATION UNIT 2 MAKEUP TO THE RCS WITH THE REACTOR MAKEUP SYSTEM UNAVAILABLE RESPONSE TO NUCLEAR POWER GENERATION/ ATWS UNIT 1 EMERGENCY BORATION UNIT 1 lBOL 1.1 LCOAR SHUTDOWN MARGIN (SOM) TECH SPEC LCO 3.1.1 lBOL l.H LCOAR SHUTDOWN MARGIN (SOM) - MODE 1 AND MODE 2 WITH KEFF GREATER THAN 1.0 TRM LCO 3.1.H lBOL 1.1 LCOAR SHUTDOWN MARGIN (SDM)-MODE 5 TRM LCO 3.1.1 lBOSR 0.5-2.CV.2-2 UNIT ONE 1CV112D 1CV112E AND 1CV8104 STROKE TIME, POSITION INDICATION TEST AND EMERGENCY BORATION FLOW TEST 2BFR S.1 2BOA PRl-2 RESPONSE TO NUCLEAR POWER GENERATION/ ATWS UNIT 2 EMERGENCY BORATION UNIT 2
50.59 REVIEW COVERSHEET FORM Station/Unit(s): Braidwood Station Units 1 & 2, Byron Units 1 & 2 LS-AA-I 04-100 I Revision 4 Page 2 of2 Activity/Document Number: TRM Change #22-008 (BRW), #22-008 (BYR)
Revision Number: 00 DRP #20-036 (BRW), #20-036 (BYR), Procedures in Activity Description
Title:
TRM Change and Procedure Revisions for change in TSR 3.1.b.4 boration flow 2BOL 1.1 2BOL l.H LCO 3.1.H LCOAR SHUTDOWN MARGIN (SDM) TECH SPEC LCO 3.1.1 LCOAR SHUTDOWN MARGIN (SDM) - MODE 1 AND MODE 2 WITH KEFF GT 1.0 TRM 2BOL 1.1 LCOAR SHUTDOWN MARGIN (SDM) - MODE 5 TRM LCO 3.1.1 2BOSR 0.5-2.CV.2-2 UNIT TWO 2CV112D, 2CV112E, AND 2CV8104 STROKE TIME, POSITION INDICATION, AND EMERGENCY BORATION FLOW TEST Reason for Activity:
(Discuss why the proposed activity is being performed.)
During performance of the Braidwood 1 Cycle 24 reload analysis (specifically the BORDER analysis), Westinghouse identified that the required boration flow rate to offset Xenon burnout exceeded the previous flowrate of 30 gpm. This change supports operation into the future by bounding the expected required flowrates.
Effect of Activity:
(Discuss how the activity impacts plant operations, design bases, or safety analyses described in the UFSAR.)
The boration flow rate surveilled in accordance with the analysis is changed from 30 gpm to 33 gpm. In addition, Operating procedures that require emergency boration for off-normal conditions will ensure the boration flowrate is greater than 33 gpm.
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 boration flowrate being changed from 30 gpm to 33 gpm remains well within the pump performance defined in Table 9.3-3 of the UFSAR of75 gpm design flow rate. The testing changes remain within the bounds of those existing design ranges. The change is minimal as compared to the existing design conditions, and does not result in a change in methodology.
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-2022-24 Rev.
00 6E-22-016
~
50.59 Evaluation 50.59 Evaluation No.
BRW-E-2022-25 Rev.
00 6G-22-001 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 Station/Unit(s): Braidwood Units 1 & 2 LS-AA-104-1001 Revision 4 Page 1 of2 Activity/Document Number: =E~C_6~3~7~2=3~0 ________________ _
Revision Number: ~0 ____ _
Title:
Implementation of Braidwood BOL MTC elimination 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 implements a new method for confirming low power physics testing as it relates to the Beginning of Life (BOL)
Moderator Temperature Coefficient (MTC) check.
ma PWROG-19014-P is a PWR Owner's Group (PW ROG) report which outlines a method for confirming BOL startup MTC without performing any explicit ITC testing. In the report, it demonstrated that all BOL MTC measurements were found to be within +/-2.0 pcm/°F, therefore if the BOL MTC calculation has at least 2.0 pcm/°F margin to the TS MTC limit, the startup boron concentration is tested to be within 50 ppm of predicted and the core design reload analysis is performed using an NRC approved neutronic code outlined in PWROG-190 I 4-P, the explicit startup MTC check via ITC testing is not needed.
Using this requirement for BOL MTC confirmation, the physical plant manipulations/testing for MTC can be bypassed.
Note that this activity does not modify or change the process of End of Life (EOL) MTC confirmation (SR 3.1.3.2). PWROG-19014-P has a separate EOL MTC elimination method that is NOT being implemented here.
Reason for Activity:
(Discuss why the proposed activity is being performed.)
This activity is being made to reduce the frequency of BOL MTC testing during startup. A reduction in the time to perform MTC will result in reducing the critical path time associated with the return to power operations after a refueling/ maintenance outage which in turn will reduce the time spent in unique plant configurations and performing operations maneuvers at lower power conditions. Furthermore, this will minimize unnecessary maneuvers that could put undue wear and tear on the plant components.
Effect of Activity:
(Discuss how the activity impacts plant operations, design bases, or safety analyses described in the UFSAR.)
The effect of this activity would be to minimize the frequency of MTC testing during startup. This activity will affect TS Bases and require updates to procedures BwVS 500-6 "Low Power Physics Test Program" at Braidwood and NF PWR Core Design procedure updates (tracked by AT-4385717-22).
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 involves changes to methodologies for confirming the BOL MTC prior to full power operation. The 50.59 process has determined that alternate methodologies will be used for the MTC confirmation thus requiring a 50.59 evaluation in which question #8 must be answered. The 50.59 evaluation determined that the new MTC confirmation process utilizes a different but conservative methodology in which excess conservatism between predicted MTC values and the MTC limit have been shown to bound testing variations found in recent industry experience. Therefore, prior NRC approval is not required to implement the new MTC confirmation process.
Attachments:
Attach all 50.59 Review forms completed, as appropriate.
50.59 REVIEW COVERSHEET FORM LS-AA-104-1001 Revision 4 Page 2 of2 Station/Unit(s): Braidwood Units I & 2 Activity/Document Number: =E=C'-'6"""'3'---'7--=2=3=0 _______________ _
Title:
Implementation of Braidwood BOL MTC elimination Forms Attached: (Check all that apply.)
Applicability Review 50.59 Screening 50.59 Evaluation 50.59 Screening No.
BRW-S-2022-21 50.59 Evaluation No.
BRW-E-2022-22 Revision Number: ~O ____ _
Rev.
0 Rev.
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.
50.59 REVIEW COVERSHEET FORM Station/Unit(s): Byron & Braidwood Units I & 2 LS-AA-I 04-100 I Revision 4 Page I of2 Activity/Document Number: =E~C_6~2~8~6=2~7~&~6=2~86~2~8~------------
Title:
Updates to B&B Chapter 15 Dose Analyses Revision Number: -=-0 ____ _
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 implements updates to the following Byron and Braidwood UFSAR Chapter 15 Dose Analyses:
BRW-19-0019-M, Rev. 0 / BYRI 9-027, Rev. 0, "Byron & Braidwood Source Terms for Use in Alternative Source Term Analyses" dated July 8, 2019.
BYR04-047, Rev. 6 / BRW-04-0041-M, Rev. 6, "Re-analysis of Fuel Handling Accident (FHA) Using Alternative Source Terms" dated October 27, 2020.
BRW-04-0038-M, Rev. 7 / BYR04-051, Rev. 7, "Re-analysis of Loss of Coolant Accident (LOCA) Using Alternative Source Terms" dated May 2, 2022.
BRW-04-0029-M, Rev. 5 / BYR04-045, Rev. 5, "Re-analysis of Control Rod Ejection Accident (CREA) Using Alternative Source Terms" dated May 13, 2022.
BRW-04-0043-M, Rev. 5 / BYR04-049, Rev. 5, "Re-analysis of Locked Rotor Accident (LRA) Using Alternative Source Terms" dated April 27, 2022.
BRW-04-0040-M, Rev. 6 / BYR04-046, Rev. 6, "Re-analysis of Main Steam Line Break (MSLB) Accident Using Alternative Source Terms" dated May 13, 2022.
BRW-22-0006-M, Rev. 0 / BYR22-0l 4, Rev. 0 "Re-analysis of Steam Generator Tube Rupture (SGTR) Accident Using Alternate Source Terms" dated April 28, 2022.
These new revisions cover the implementation of an updated source term along with several error corrections and their associated changes. The summary of all changes and which cal cs were impacted are found below:
New Source Term reflective of more advanced core design practices first identified in IR 2602135. All calculations were impacted.
Corrected RadTrad Error Notice 17 discussed in IR 3983381. CREA was impacted.
Corrected the "Decay plus Daughter" error discussed in IR 4077095. LRA & SGTR were impacted.
Control Room flow rates were corrected based on IR 4282529. All calculations were impacted.
Control Room inleakage was reduced to 436 cfm along with the other Control Room flow rate corrections. CREA, MSLB, FHA, LRA and SGTR were impacted.
Control Room HY AC isolation time was lowered in conjunction with the control room flow rate changes. This impacted Operations procedures which control Time Critical Actions, specifically OP-BR-I 02-106 (TCA #25) and OP-BY-102-106 (TCA #29).
Reason for Activity:
(Discuss why the proposed activity is being performed.)
This activity is done to both update the UFSAR Chapter 15 dose analyses to account for the new source term based on IR 2602135 as well as to correct the errors identified in IR 4077095, 3983381, 4282529. Doing both of these ensures the UFSAR Chapter 15 dose analyses are accurate and align with core design practices for Byron and Braidwood.
Effect of Activity:
(Discuss how the activity impacts plant operations, design bases, or safety analyses described in the UFSAR.)
The fuel-related inputs and control room HYAC parameters of the UFSAR Chapter 15 dose analyses are being updated along with additional error corrections. All final results demonstrated that there was a less than minimal increase in dose releases while also staying below their prescribed limits as outlined in the Table below. Site operations procedures OP-BR-I 02-106 &
OP-BY-I 02-106 are being updated to account for the change of Control Room HY AC isolation.
50.59 REVIEW COVERSHEET FORM Station/Unit(s): Byron & Braidwood Units I & 2 LS-AA-104-1001 Revision 4 Page 2 of2 Activity/Document Number: _E~C_6~2~8~6~2~7~&~6~2~86~2~8~------------
Title:
Updates to B&B Chapter 15 Dose Analyses Revision Number: ~0 ____ _
UFSAR Ch. 15 Analysis AOR New Revision Limit FHA Control Room 4.40 3.37 5.0 EAB 5.31 5.35 6.3 LPZ 0.94 0.95 6.3 LOCA Control Room 4.94 4.94 5
EAB 15.03 15.27 25 LPZ 4.99 5.16 25 CREA Control Room 4.54 4.52 5
EAB 5.36 5.37 6.3 LPZ 2.28 2.59 6.3 LRA Control Room 2.79 2.98 5
EAB 1.68 1.69 2.5 LPZ 0.60 0.61 2.5 MSLB Control Room 0.26 I 0.65 0.26 I 0.60 515 EAB 0.15/0.20 0.15/0.20 25 I 2.5 LPZ 0.08 I 0.46 0.08 I 0.46 25 I 2.5 SGTR Control Room 2.0 I 0.56 I.93 I 0.55 5/5 EAB 3.7 / 2.1 3.5 I 2.0 25 I 2.5 LPZ 0.69 I 0.41 0.65 I 0.38 25 I 2.5 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 involves changes to UFSAR Chapter 15 dose analyses. The total effective dose equivalent (TEDE) results for the dose-significant scenarios have increased slightly at select locations of interest. These dose analyses utilize the same methodologies currently in the licensing basis. For each of these accidents, the final results ofMSLB and SGTR remained at or below the Analysis of Record values. For the LOCA, FHA, CREA and LRA, the final limiting values all increases are no more than minimal. Furthermore, despite those increases, all values remain within their acceptance criteria. Therefore, these updated dose analyses can be implemented under the 50.59 process and do not require a licensing amendment.
Attachments:
Attach all 50.59 Review forms completed, as appropriate.
Forms Attached: (Check all that apply.)
Applicability Review 50.59 Screening 50.59 Evaluation 50.59 Screening No.
BRW-S-2022-43 I 6E-22-024 50.59 Evaluation No.
BRW-E-2022-42 I 6G-22-004 Rev.
0 / 0 Rev.
0 / 0 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 Sta tion/U nit(s ): _....;B::;;ra.aa;;.;;.id""w""-"-o"""odaa..a:S=ta=t=io=n"-'-/ -=-U.a.an~it""s... l....:aa.:n.:.:d=-2=--
LS-AA-I 04-100 I Revision 4 Page I of2 Activity/Document Number: ~T~R-=-M---=C~h=a=n0ge=--=23;..-....:0c.:0.... 1 ___________ _
Revision Number: ___ 0~--
Title:
Thermal Overload Protection Surveillance Interval Change for Ul and U2 SX00lA Valves 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(d)(2).
Description of Activity:
(Provide a brief, concise description of what the proposed activity involves.)
Revise Technical Requirements Manual (TRM) Table T3.8.b-2 to extend the interval for the Thermal Overload (TOL)
CHANNEL CALIBRATION for two Motor Operated Valves (MOVs): I SXO0 I A (Essential Service Water Pump I A Suction Valve) and 2SX0OIA (Essential Service Water Pump 2A Suction Valve). These PMs are tracked under PMIDs 188679 and 188998, respectively.
In TRM Section 3.8.b, Motor Operated Valves Thermal Overload Protection Devices, TSR 3.8.b. l contains a NOTE stating, "Each thermal overload is calibrated and each valve is cycled through at least one complete cycle of full travel with the motor operator when the thermal overload is OPERABLE." The TSR requirement is to perform CHANNEL CALIBRATION at a frequency of6 years. With the 25% grace allowed per TRM Surveillance Requirement (TSR) 3.0.b and the 6-year frequency specified in TSR 3.8.b. l, the maximum current interval is 90 months. With TRM Change 23-00 I, the two MOY TOLs will be allowed to go a maximum of I 08 months during the current operating cycle between CHANNEL CALIBRATIONS to bridge to the next Refueling Outage and align the future PM frequency to outage schedulces. This is the maximum duration required from the last credited PM to reach A2R24 (97 months for UI to reach AIR24).
The TOL is designed to protect the motor from damage under prolonged overload condition.
TRM Change 23-00 l also removes the note related to the extended TOL calibration interval for valves I SI880 I B, 1 S188028 and 2Sl8804B. This note is no longer needed.
Reason for Activity:
(Discuss why the proposed activity is being performed.)
The TOL CHANNEL CALIBRATIONS for I SX00 I A and 2SX00 I A have been performed previously online for their respected Preventive Maintenance Identifiers (PMID 188679 and I 88998). The calibration requires a post maintenance test to stroke the valve to confirm rotation and circuit continuity since the calibration requires the TOL terminations to be lifted and landed. Operations has concerns with the potential plant impact of stroking these MOVs at power. The TOL tests for I SX00 I A and 2SX00 I A are due 10/3/2023 and 4/20/2023, respectively. In order to minimize risk and maintain divisional separation of work, outages are generally scheduled to avoid cross train work being performed in the same outage. The proposed one time extension to the frequency interval would align the performance of the TOL CHANNEL CALIBRATION with the next associated outage for the specific train. The next performance of the TOL CHANNEL CALIBRATION will be performed prior to the completion of A I R24 (prior to reaching mode 4) for l SX00 I A and A2R24 (prior to reaching mode 4) for 2SX001A.
Effect of Activity:
(Discuss how the activity impacts plant operations, design bases, or safety analyses described in the UFSAR.)
This activity extends the frequency of the Thermal Overload (TOL) CHANNEL CALIBRATION for Motor Operated Valves (MOVs) I SXO0 IA (Essential Service Water Pump I A Suction Valve) and 2SX00 1 A (Essential Service Water Pump 2A Suction Valve). The change does not impact plant operation.
The TRM change screens as adverse since it is extending an existing surveillance interval. The TRM 3.8.b and Tech Spec requirements are based on Regulatory Guide I. I 06, THERMAL OVERLOAD PROTECTION FOR ELCTRIC MOTORS ON MOTOR-OPERA TED VALVES. The primary objective of the Reg Guide is to ensure the TOL does not prevent MOVs from performing their support function under accident conditions.
50.59 REVIEW COVERSHEET FORM Station/Unit(s):
Braidwood Station / Units 1 and 2 LS-AA-I 04-100 I Revision 4 Page 2 of2 Activity/Document Num her: ----'T:....:R~M:..:..::....C=h=a.:.:ng.._e=---=23=----=0-"0-=-1 __________ _
Revision Number: __
__,_,O'-----
Title:
_--.::..T.:.:h.:cer:....:m=al=-O=.:.v.::.e:..:rl""o=-a=-d-=-P-=-r-"o.:.::te:.=cc.:..:ti""o.::.n....:aS:..::u:.:.r-'-v.:cei"""ll=a.:..::n.=;ce:...::.:In:a.:t~e"-rv'""aaa.:l-'C::.:h""a""n.,,.g..,e'""faa,o.a..r--=U"-'1"""a""n:a.:d::...a:U'-=2'-'S::.:X-"0.;..0""l'-"A.a....aV--=a:.:.lv~ea.::s ______ _
These valves are motor operated (MOY), butterfly style, and fail as is. The valves are in the open position during normal operation, with power locked out. The valves remain in the open position and support the operation of the SX system during a Design Basis Accident (OBA) or transient. This configuration is not changed by the TRM revision so that the design function of the SX system is maintained. The plant safety analyses are therefore not affected.
TRM Change 23-00 I also removes the note related to the extended TOL calibration interval for valves I Sl880 I B, I SI8802B and 2Sl8804B. This is an administrative change as the note is no longer needed.
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 lSX00IA and 2SXOO!A valves are motor operated (MOV), butterfly type valves, and fail as is. The valves are in the open position during normal operation, with power locked out. The TRM change does not alter this configuration.
Therefore, these valves are not susceptible to a malfunction that closes the valve. The valves remain in the open position and support SX pumps operation at the onset of a Design Basis Accident (DBA).
Pursuant to the 50.59 Resource Manual, accidents evaluated in the UFSAR also include other events for which the plant is designed to cope and are described in the UFSAR. The Auxiliary Building flooding is such an event. The SX system is classified as a moderate energy fluid system with breaks postulated in piping locations that exceed the Moderate Energy Line Break (MELB) stress criteria. Flooding calculation 3C8-0685-002 determined that MELBs do not need to be postulated for any of the SX piping lines in the Auxiliary Building.
Calculation 3C8-0685-002 calculates the flood level due to leakage from a postulated piping break assuming isolation in 30 minutes. The 30 minute isolation time discussed in the UFSAR is related to the isolation of design basis breaks. Since SX piping breaks are not design basis breaks, the isolation method for the Auxiliary Building Flooding Analysis is not impacted. Therefore, this change does not result in a more than minimal increase in consequences for an accident previously evaluated in the UFSAR.
The proposed activity does not introduce new accidents because the valves are not initiators of any accident and no new failure modes are introduced. No fission product barriers are altered or exceeded. There are no changes to methodologies associated with the analyses impacted by the modification.
Based on the above, the proposed change 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 l8l 50.59 Screening 50.59 Evaluation 50.59 Screening No.
50.59 Evaluation No.
BRW-E-2023-001 Rev.
Rev.
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.
EC 637092 Rev 000, Attachment 2 50.59 REVIEW COVERSHEET FORM Station/Unit(s): Braidwood/ Units 0, t, 2 Page 1 of 3 LS-AA-I 04-100 I Revision 4 Page I of3 Activity/Document Number: EC# 637092, DRP 20-044 Revision Number: ~
Title:
Reclassify ASME III FP Piping/ Valves/ Components In Seismically Qualified Areas 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( d)(2).
Description of Activity:
(Provide a brief, concise description of what the proposed activity involves.)
Engineering Change (EC) 637092 is a Design Change Document Change Request (OCR) that incorporates the supporting documentation for the reclassification of Fire Protection (FP) piping and components in the Auxiliary Building, Fuel Handling Building, and Containment from Safety Category I, Quality Group C (ASME Section III, Class 3) to Safety Category II, Quality Group D (Non-Safety Related ANSI 831.1) with the exception of the containment piping from check valves I (2)FP345 up to and including relief valves I (2)FP450 and selected other supports which will remain classified as Safety Category I, Quality Group C. All the affected FP piping and components will continue to be classified as Seismic Category I. The classification of the containment isolation portion of the FP System which is Safety Category I, Quality Group B (ASME Section III, Class 2) is not affected by this OCR.
The Safety Classification of the FP System SSCs impacted by this OCR, as documented in PassPort, will be Augmented Quality (AQ) in accordance with CC-AA-304 and NO-AA-10, Section A.2.4 to meet 10 CFR 50 Appendix A, General Design Criterion (GDC) 3. This classification is consistent with the remainder of the FP System in non-seismically qualified areas. In addition, the affected SSCs have the augmented requirement of being qualified as Seismic Category I.
The documentation changes incorporated under this OCR include drawing changes, a new calculation, UFSAR DRP 20-044, PassPort equipment/component data changes, and piping specification changes. Fire Protection Program changes, such as the revision to the Fire Protection Report, are evaluated separately under LS-AA-128. lnservice Inspection Plan changes are evaluated separately under its associated program.
Reason for Activity:
(Discuss why the proposed activity is being performed.)
A portion of the FP System piping at the Braidwood Generating Station was conservatively classified as Safety Category I, Quality Group C during plant construction. This classification maintained consistency between the Seismic Category I requirements for the FP System as delineated in the Fire Protection Report and the Safety Category classification requirements included in UFSAR Section 3.2.1.1. This classification, however, imposes unnecessary ASME code requirements and special treatment under IO CFR 50 Appendix B during preventive & corrective maintenance, design change control, procurement, work control, testing, quality inspection and documentation for components that do not perform a function important to safety. The changes incorporated under EC 637092 relax process burden by controlling the FP System commensurate with the Safety Category II, Quality Group D functions it performs. This permits utilizing plant resources to focus on safety significant SSCs.
Effect of Activity:
(Discuss how the activity impacts plant operations, design bases, or safety analyses described in the UFSAR.)
The changes incorporated under this OCR do not change the function of the FP System SSCs or the functions supported by the FP System. The FP System continues to provide a Seismic Category I standpipe system to supply hose stations within the Auxiliary Building, Fuel Handling Building, and Containment which can be supplied from the Essential Service Water System (SX), if necessary, during a safe shutdown earthquake event. The Seismic Category I standpipe system remains capable of providing a backup source of makeup water to the spent fuel pool. There is no change to the alternate cooling water supply provided by the standpipe system for the Chemical and Volume Control System (CV) Pumps. The portion of the FP System that provides containment isolation and penetration over-pressure protection is not changed by this OCR.
The analytical methods of demonstrating FP SSC structural qualification and ability to remain functional during design basis events are unchanged. The Seismic Category I classification of these SSCs is maintained. Therefore, the structural requirements for a Safety Category I remain applicable to the affected portion of the FP System that is being reclassified to Safety Category II.
EC 637092 Rev 000, Attachment 2 50.59 REVIEW COVERSHEET FORM Station/Unit(s): Braidwood/ Units 0, 1, 2 Page 2 of 3 LS-AA-I 04-100 I Revision 4 Page 2 of3 Activity/Document Number: EC# 637092, DRP 20-044 Revision Number: !l,J!
Title:
Reclassify ASME III FP Piping/ Valves/ Components In Seismically Qualified Areas As such, no undesirable structural interaction is created with interfacing Safety Category I SSCs such as the supply from the SX System and the FP System containment penetration.
No physical or functional changes are implemented under EC 637092. Repair and replacement activities following the implementation of this DCR will be governed by ANSI B31.1 and the augmented quality requirements associated with the Fire Protection Program and the maintenance of the Seismic Category I classification of the affected SSCs. These requirements establish the necessary process controls and treatment for the system with respect to maintaining pressure boundary integrity, structural integrity and functionality following a safe shutdown earthquake event.
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 Screening BRW-S-2022-41:
Although the screening and justification included in EC 637092 concluded that the reclassification of the seismically qualified portions of the FP System to Safety Category II, Quality Group D was appropriate, the change was considered adverse. The UFSAR specifically references the Safety Category I classification of the FP System in the discussion of its design functions.
The changes ultimately relax process requirements and special treatment under the station's 10 CFR 50 Appendix B program for future repair and replacement activities that could potentially reduce FP System reliability which in turn could impact other safety related SSCs.
The document changes incorporated under EC 637092 do not change procedures that impact the operation of the FP System as described in the UFSAR.
The UFSAR described evaluation methodology associated with the piping, components, and supports for the FP System in seismically qualified areas is unaffected by the changes implemented under EC 637092.
This activity does not represent a test or experiment. There are no physical, analytical, or operational changes being implemented for the FP System that would be inconsistent with the analyses or descriptions in the UFSAR.
The FP System containment isolation function associated with Technical Specifications is not impacted by EC 637092. In addition, the containment penetration over-pressure protection provided by relief valves I (2)FP450 are unaffected by this DCR.
No Technical Specification or Facility Operating Changes are necessary to implement the changes under EC 637092.
Based on the adverse change in component classification to Safety Category II, Quality Group D, it was concluded that the activity required a IO CFR 50.59 Evaluation.
The following is a summary of the conclusions in Safety Evaluation BRW-E-2022-40:
As justified in EC 637092 and Screening BRW-S-2022-41, the changes incorporated under EC 637092 do not represent a departure from design, fabrication, construction, testing, and performance standards specified in the General Design Criteria and other regulatory requirements for the appropriate Safety Category II, Quality Group D classification of the affected portion of the FP System. EC 63 7092 does not introduce the possibility of a change in the frequency of an accident because the seismically qualified portion of the FP System and associated SSCs is not an initiator of any accident and no new failure modes are introduced.
The affected portion of the FP System will meet the design requirements for material and construction practices commensurate with its non-safety related design function. Safety analyses are unaffected, and the proposed changes will not degrade the system performance or reliability. Therefore, the proposed activity will not increase the likelihood of a malfunction.
EC 637092 Rev 000, Attachment 2 50.59 REVIEW COVERSHEET FORM Station/Unit(s): Braidwood/ Units 0, 1, 2 Page 3 of 3 LS-AA-104-1001 Revision 4 Page 3 of3 Activity/Document Number: EC# 637092, DRP 20-044 Revision Number: !h.Q
Title:
Reclassify ASME III FP Piping/ Valves/ Components In Seismically Qualified Areas Since FP System integrity is maintained and containment penetration and SX System integrity is unaffected by the proposed change, the change will not result in an increase in the consequences of previously analyzed accidents.
There are no physical changes introduced by the proposed activity. The consequences of malfunctions previously evaluated are not affected because the mitigation functions performed by the SSCs assumed in the safety analysis are not being modified. The FP System is not credited to perform any accident mitigation functions as documented in the UFSAR.
The augmented controls maintained for the affected SSCs, Seismic Category I and as required by GDC 3, ensure no new failure modes are created that could impact or reduce the availability, operability, or effectiveness of equipment important to safety and introduce an accident different from any previously evaluated in the UFSAR.
Since no new failure modes are introduced by the proposed activity, there is no potential for a different result associated with a malfunction of the seismically qualified portions of the FP System affected.
The reclassification of the seismically qualified portion of the FP System under EC 637092 does not result in a change that would cause any system parameter to change. The changes implemented under the proposed activity do not impact a fission product barrier. Therefore, the proposed does not result in design basis limit for a fission product barrier (DBLFPB) as described in the UFSAR being exceeded or altered.
The UFSAR described evaluation methodology associated with the piping, components, and supports for the FP System in seismically qualified areas is unaffected by the changes implemented under EC 637092. The changes do 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, IO CFR 50.59 Evaluation concluded that the activity can be implemented per plant procedures without obtaining a License Amendment.
Attachments:
Attach all 50.59 Review forms completed, as appropriate.
Forms Attached: (Check all that apply.)
~
Applicability Review 50.59 Screening 50.59 Evaluation 50.59 Screening No.
50.59 Evaluation No.
BR W-S-2022-41 Rev.
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BRW-E-2022-40 Rev.
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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-104-100 I Revision 4 Page I of 4 Station/Unit(s): Braidwood/Unit I and 2 Activity/Document Number: EC 638350 / EC 638351 Revision Number: 000/000
Title:
Feedwater Regulating Valve Dual Positioner - Unit I / Unit 2 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.)
The proposed activity installs redundant Fisher DVC6200 series digital valve controllers (DVC) for each Feedwater Regulating Valve (FWRV) to provide modulating control based on flow demands from the Ovation Distributed Control System (DCS). The dual positioner redundancy configuration uses two separate analog control outputs connected to separate, redundant valve positioners. The use of two separate control circuits (control output and valve positioner) prevents the failure of one circuit from affecting the control valve since there is a redundant output circuit available to maintain valve position control. This configuration eliminates the effects of a single failure that could initiate a plant transient.
The proposed change utilizes two redundant loop interface (RLI) modules whose analog outputs are wired individually to each positioner, primary and backup. The positioner switchover/failover control is implemented using an automatic or manual switchover. Redundant digital outputs from the control system, wired in a 1/2 configuration, are used to demand the switchover and redundant digital outputs from the control system, wired in a 1/2 configuration, are used to demand the lockup.
FWRV position feedback is required to detect if there is valve movement when not demanded. The position feedback from each positioner is derived from hard-wired position sensors connected to the valve.
The control demand signal is connected in series with an analog input to sense the demand. The demand sense signal is compared against the controller demand output signal to ensure the two values are within a normal operating tolerance. When the positioner is in control, a deviation beyond normal operating tolerance will initiate a switchover to the alternate control circuit. When in standby, the positioner will be disarmed. Automatic failover to a disarmed positioner is not possible.
The position feedback signal is compared against demand sense signal to ensure the two values are within a normal operating tolerance. When the positioner is in control, a deviation beyond normal operating tolerance will initiate a switchover to the alternate control circuit. When in standby, the positioner will be disarmed. Automatic failover to a disarmed positioner is not possible.
Pressure switches installed in the primary and backup instrument air headers will sense a loss of instrument air supply. When the positioner is in control, a loss of its instrument air supply will initiate a switchover to the alternate control circuit. When in standby, the positioner will be disarmed. Automatic failover to a disarmed positioner is not possible.
A new feature, lockup, will be incorporated in the proposed activity that will prevent all FWRV movement when failures of both control channels are detected. This could include any combination of the conditions mentioned above. Lockup effectively fails the FWRV in the as-is position and permits operators to take manual actions to maintain steam generator levels.
The four primary DVCs will be located in existing local panel l/2PLI00J. The four secondary/backup DVCs will be installed in an identical new local panel 1 /2PL IO 1 J physically located near the existing panel. Each FWRV will have two position feedback sensors mounted on the valve actuator, one for each DVC control circuit.
New local panel l/2PL102J will contain failover solenoid valves which determine which DVC (primary or backup) control output is sent to the FWRV. New panel 1/2PL102J will also contain four lockup solenoid valves that will energize and isolate air to the affected FWRV when a lockup is demanded.
On a lockup demand, the Ovation DCS places the FWRV in manual control. Placing the controller in manual enables automatic control of the associated FWRV bypass valve. Normally the FWRV bypass valves are in automatic control, but closed by the Ovation DCS at steam flows greater than 30% ofrated during power ascension. Enabling automatic control of the initially closed FWRV bypass valves provides a mechanism to address an underfeed condition for the affected steam generator that may have existed at the time of lockup. Operator action is solely relied on to address an overfeed condition at the time of lockup.
New HMI display graphics will be created for Operations personnel to interact with the DVCs for the FWRVs. The ability to swap to a DVC or lock up a FWRV will be performed from the new display graphic. Through the new HMI displays, Operations personnel will be able to determine valve position demand, position feedback, and DVC status to better understand field conditions. These displays allow the Operator to dis-arm/re-arm any DVC, lock/reset the lockup function, and
50.59 REVIEW COVERSHEET FORM LS-AA-104-100 I Revision 4 Page 2 of 4 Station/Unit(s): Braidwood/Unit I and 2 Activity/Document Number: EC 638350 / EC 638351 Revision Number: 000/000
Title:
Feedwater Regulating Valve Dual Positioner - Unit I/ Unit 2 enable/disable the lockup function. Additionally, a manual hold function is provided that allows the Operator to prevent an automatic failover while working with the controls.
Procedures will be updated to reflect the changes indicated above.
Reason for Activity:
(Discuss why the proposed activity is being performed.)
The FWRVs at Braidwood Unit l and 2 are each controlled by a single DVC with a flow demand signal from the Ovation DCS.
The existing design presents a single-point vulnerability in that the failure of a single DVC for a FWRV could initiate a steam generator level transient resulting in a reactor trip. There is Operating Experience across the Constellation fleet for FWRV controller failures resulting in a reactor trip or scram. The lockup feature for the FWRVs provides an additional mechanism to prevent a reactor trip by permitting operators to take action to control and maintain steam generator levels in the event both controllers fail or lose their instrument air supply.
Effect of Activity:
(Discuss how the activity impacts plant operations, design bases, or safety analyses described in the UFSAR.)
The Ovation DCS will monitor the position demand to each DVC and valve position feedback signals for each FWRV. If a deviation between demand signal and feedback signal becomes indicative of the failure of a DVC or if instrument air is lost, the Ovation DCS will initiate the control swap (single DVC failure or loss of single Instrument Air) by energizing a failover solenoid valve which blocks the control air signal from the failed/failing control path to the redundant DVC control path. The Ovation DCS will lock the affected valve in place by energizing a lockup solenoid valve (Redundant DVC failure or loss of both instrument air sources) depending on other circumstances described in the detailed design requirements. Although these new features are intended to improve the overall reliability of the steam generator water level control (SGWLC) system design functions as described in the UFSAR, the proposed change introduces a new failure mode for the FWRVs that impacts that impacts the design functions and the method of performing and controlling the design functions.
The safety analyses consider accidents and malfunctions that result in a complete loss of feedwater to a condition which results in the maximum feedwater flow that can be delivered to the steam generators. Since the proposed activity affects the delivery of feedwater to the steam generators, these events establish the bounding conditions for the design.
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 changes to management and administrative procedures (Applicability Review Section 111.2), and maintenance procedures (Applicability Review Section 111.3) are exempt from further review under 50.59 as they fall into those criteria and are governed by other processes.
The following paragraphs summarize the conclusions of the 50.59 Screening prepared for the proposed change; The installation ofredundant DVCs with automatic switchover on sensed DVC failure or loss of instrument air improves overall system reliability and has no adverse impact on the SGWLC system and support system design functions. The new failure mode (fail as-is) associated with the lockup function for the FWRVs, however, adversely impacts SGWLC design functions and the method of performing or controlling the design functions. These adverse impacts require further review under a 50.59 Evaluation.
The addition of the lockup feature impacts Control Room Operator response time with respect to the control and operation of the SGWLC system. These impacts are considered adverse requiring further review under a 50.59 Evaluation.
The type ofpositioner/controller used to control the Non-Safety Related FWRVs is not specified or referenced in the UFSAR and is not related to any UFSAR described evaluation methodology or used in establishing the design bases or safety analyses.
The proposed change does not adversely impact calculations or supporting analyses described in the UFSAR.
Impacted calculations that require revision to support this activity have maintained existing methodology. Therefore, this activity does not involve an adverse change to an element of a UFSAR described evaluation methodology, or use of an alternative evaluation methodology, that is used in establishing the design bases or used in the safety analyses.
50.59 REVIEW COVERSHEET FORM LS-AA-I 04-100 I Revision 4 Page 3 of 4 Station/Unit(s): Braidwood/Unit I and 2 Activity/Document Number: EC 638350 I EC 638351 Revision Number: 000/000
Title:
Feedwater Regulating Valve Dual Positioner - Unit I / Unit 2 The proposed activity will not result in any equipment being used outside its referenced operating bounds or limit. As such, this is not a test or experiment. Therefore, this activity does not involve a test or experiment not described in the UFSAR, where an SSC is controlled in a manner outside the reference bounds of the design.
The Non-Safety Related FWRVs are not described in the Technical Specifications or Operating License. These valves are listed as feed water isolation valves in the basis for Technical Specification 3.3.2. However, the proposed changes to the control system for the valves do not impact feedwater isolation in any way. Therefore, no Technical Specification or Operating License change is required for this activity.
Based on the adverse impacts identified with the lockup feature for the FWRVs on the SGWLC design functions and the method of performing and controlling the design functions, a 50.59 Evaluation is required.
The following paragraphs summarize the conclusions of the 50.59 Evaluation prepared for the proposed change. The 50.59 Evaluation for this digital upgrade considered the supplemental guidance in NEI-01-0 I for the digital upgrade aspects of the proposed change. In addition, a Qualitative Assessment was prepared consistent with RJS 2002-22 Supplement I to support the conclusion that the likelihood of failure of the digital upgrade is sufficiently low.
Based on the component failure analysis, the design attributes incorporated to reduce the likelihood of failure, the quality of the design process, and operating experience, the likelihood of a failure or malfunction of multiple DVCs occurring that requires a lockup and the likelihood of a spurious Ovation DCS component failure that initiates an undesired lockup, is sufficiently low.
The selection of logic setpoints that initiate a lockup is optimized to provide the best possibility ofrecovering steam generator level to preclude a reactor trip. Furthermore, the setpoint sensitivity is such that they preclude an undesired lockup during normal operations and transients. The likelihood of failure or malfunction is such that the occurrence of a lockup is sufficiently low such that the actual increase in operator burden is insignificant. Therefore, the likelihood of a component or system failure associated with the proposed change is sufficiently low such that there is no more than a minimal increase in the frequency of the occurrence of the Loss of Normal Feedwater accident or Feedwater System Malfunction Causing an Increase in Feedwater Flow malfunction previously evaluated in the UFSAR.
The radiological consequences of a Loss of Normal Feedwater accident and Feedwater System Malfunction event are bounded by the secondary system steam release and associated doses for a Main Steamline Break at Zero Power. No aspect of the proposed change results in a condition that would affect secondary system steam releases. Therefore, the steam release during the Main Steamline Break event at zero power would remain the bounding case for evaluating the radiological consequences of a Loss of Normal Feedwater accident and a Feedwater System Malfunction event. The radiological consequences of these events remain unchanged.
The only credible accidents/events associated with the lockup feature for the FWRV s are those which cause a reduction in flow (underfeed) or an increase in flow (over feed). These results are bounded by the Loss of Normal Feedwater and Feedwater System Malfunction events in the UFSAR which consider an instantaneous loss offeedwater flow and a malfunction resulting in the maximum feedwater flow delivered to all four steam generators. The modified system will not interface with other plant systems and controls systems which could initiate other accidents. Therefore, the proposed activity does not create a possibility for an accident of a different type than any previously evaluated in the UFSAR.
The proposed does not introduce the possibility for a malfunction of an SSC with a different result because the activity does not introduce a failure mode that is not bounded by the Loss of Normal Feedwater or Feedwater System Malfunction events described in UFSAR. Any impact on a FWRV is bounded by these events. Therefore, the proposed activity does not create a possibility for a malfunction of an SSC important to safety with a different result than any previously evaluated in UFSAR.
The proposed change will not impact the feedwater isolation function provided by the FWRVs. The mass and energy release into containment associated with a main steamline break concurrent with a single failure of the l/2FW009A/B/C/D on the faulted steam generator will not change such that the containment fission product barrier is unchanged. No aspects of the proposed change impact the fuel cladding or Reactor Coolant System fission product barriers. Therefore, the proposed activity will not result in a design basis limit for a fission product barrier as described in the UFSAR being exceeded or altered.
As indicated in the summary for the 50.59 Screening, the proposed activity does not involve an adverse change to an element of a UFSAR described evaluation methodology, or use ofan alternative evaluation methodology, that is used in establishing the design bases or used in the safety analyses. In addition, no Technical Specification or Operating License changes are required for this activity.
50.59 REVIEW COVERSHEET FORM LS-AA-104-1001 Revision 4 Page 4 of 4 Station/Unit(s): Braidwood/Unit 1 and 2 Activity/Document Number: EC 638350 / EC 638351
Title:
Feedwater Regulating Valve Dual Positioner - Unit l / Unit 2 Revision N um her: 000/000 In summary, the proposed change can be implemented per plant procedures without obtaining a License Amendment.
Attachments:
Attach all 50.59 Review forms completed, as appropriate.
Forms Attached: (Check all that apply.)
[8J Applicability Review 50.59 Screening 50.59 Evaluation 50.59 Screening No.
50.59 Evaluation No.
BRW-S-2024-001 Rev.
0 -------
BRW-E-2024-001 Rev.
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.