ti braidwood station exelon generation company, llg

December 14, 2018 BW180119

ti

U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001

Braidwood Station, Units 1 and 2

Exelon Generation Company, LLG

Braidwood Station 35100 South Roule 53, Suite 84 Braceville, IL 60407-9619

www.exeloncorp.com

10 CFR 50.59(d)(2)

Renewed Facility Operating License Nos. NPF-72 and NPF-77 NRC Docket Nos. STN 50-456 and STN 50-457

Subject: 1 O CFR 50.59 Summary Report

Pursuant to the requirements of 1 O 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, 2016 through June 18, 2018 and consists of the 1 O 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. Francis Jordan, Regulatory Assurance Manager, at (815) 417-2800.

Respectfully,

(fi~!hwitu~ Marri Marchionda-Palmer Site Vice President Braidwood Station

Attachment: Braidwood Station 1 O CFR 50.59 Summary Report

cc: NRA 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. Rev Title

* Westinghouse Ovation Digital Upgrade for Rod Control Logic BRW-E-2015-92 1

Cabinet (N-1 Outage) BRW-E-2016-71 0 Unit 1 (2) AFW Diesel Enqine Air Intake Relocation BRW-E-2016-89 0 Westinqhouse Ovation Diqital Upqrade for DEH CN-1 Outaqe) BRW-E-2016-22 0 Revising AST Accident Dose Calculations

Temporarily Disable FW Water Hammer Prevention System BRW-E-2017-01 0 (WHPS) When 2FW009AJB/C/D Are Open (Above Approx.

25% Power) For Steam Generators 2A/2B/2C/2D Westinghouse Ovation Digital Upgrade for 7300 NSSS Cabinets 1 (2)PA05J, 1 (2)PA06J, 1 (2)PA07J, 1 (2)PA08J

BRW-E-2017-03 0 Westinghouse Ovation Digital Upgrade for 7300 BOP Cabinets 1 (2)PA20JA and 1 (2)PA20JB I Westinghouse Ovation Digital Upqrade for TDFWP Cabinets 1(2) FW36J and 1(2)FW37J

BRW-E-2017-25 0 1/2BwOA PRl-8 Essential Service Water Malfunction Unit 1/2 Disable FW Water Hammer Prevention System (WHPS) When

BRW-E-2017-26 0 2FW009A/B/C/D Are Open (Above Approx. 25% Power) For Steam Generators 2A/2B/2C/2D

BRW-E-2017-27 0 Aooendix J Local Leak Rate Scope Reduction Install/Remove Temporary Control Panel in Support of Ovation

BRW-E-2018-07 0 Digital Upgrade for 7300 NSSS and BOP Cabinets TCCP Temporary Change

The evaluation number was assigned in 2015. However, it was not completed/approved until 9/23/2016 which places it within the reporting period of June 19, 2016 through June 18,2018

50.59 REVIEW COVERSHEET FORM

Station/Unit(s): Braidwood Units I and 2

Page 1of36

LS-AA-104-1001 Revision 4 Page 1of3

Activity/Document Number: EC 400433 and EC 400435 I DRP 16-023 and DRP 16-035 Revision Number: 002 and 002 I Oand 0

Title: Westinghouse Ovation Digital Upgrade for Rod Control Logic Cabinet (N-1 Outage)

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 (EC 400433 for Unit 1 and EC 400435 for Unit 2) replaces components within the rod control logic cabinets (1RD07 J and 2RD07 J) as part of an overall phased project to upgrade key non-safety-related process control systems, integrating them into an Ovation-based distributed control system. The upgrade is being performed in accordance with the recommendations of the original equipment manufacturer (Westinghouse). Separate activities (EC 400918 and EC 400920) are installing the Ovation distributed control system "infrastructure" (hardware and software) necessary to support the upgrade of the rod control logic cabinet.

This activity will replace the existing logic cabinet internal hardware and components. The majority of the new components will be furnished as a pre-fabricated panel assembly sized according to the dimensions of the existing logic cabinet, which will be screwed to the existing rails inside the cabinet. New components will consist of redundant controllers, input/output (I/0) modules, media converters, power supplies, line filters, circuit breakers, surge suppressors, a power distribution module, a transition panel which allows connecting to additional branches ofl/O modules, a temperature and humidity transmitter, and terminal blocks.

The upgraded rod control logic cabinet is being designed by Westinghouse, the original designer and original equipment manufacturer, to reproduce the existing logic cabinet functions using new hardware and software. The rod speed and direction signals will be reproduced using new hardware and software. Instead of being adjustable via potentiometers at the cabinet, these speeds will be adjustable via the software server I engineering workstation. In auto mode, the rod speeds will be developed from the analog input using new hardware and software in the logic cabinet. Existing rod speeds and stop interlocks will be maintained. The potential for a single failure causing the rod speed to reach 77 steps/min will be eliminated, such that the maximum speed will be limited to 72 steps/min. The existing bank sequencing. bank overlap, and group alignment functions and the existing timing profile will be reproduced using new hardware and software. The bank overlap and timing profile will be adjustable via the software server I engineering workstation. As at present, overlap between successive banks will be adjustable between 0 and 50% of bank height. The upgraded rod control logic cabinet will rely on redundant controllers to perform functions that are at present performed by various cards. These functions include development of the timing profiles sent to the power cabinets, group sequencing, and bank overlap.

New internal cabinet cables will be installed that connect the Ovation I/0, bridge rectifier circuits and relays to the existing field terminal blocks located on the sides of the cabinet The pulse-to-analog (PIA) converters in the DC hold cabinets (IRD08J and 2RD08J) will be removed and replaced with analog output signals originating in the logic cabinet.

To provide maximum fault tolerance for loss of power conditions, the upgraded cabinet will utilize two independent AC-power feeds. The existing power feed provided to the cabinet from the motor-generator (MG) sets will be retained by this modification. A new second power feed will be provided from a spare circuit breaker in a local regular lighting cabinet in the miscellaneous electric equipment room (MEER). A new uninterruptible power supply will be installed in cabinets 1RD08J and 2RD08J to enhance the reliability of this power source. Auctioneering of the power sources will be provided by a new transfer switch, also installed in cabinets 1 RD08J and 2RD08J. In the event that one of these supplies fails, the other power source will continue to provide power to the Ovation system, assuring normal operation of the system.

In the control room, the existing I 00 VDC digital rod group step counters, fifteen per unit, will be replaced with new 24 VDC digital step counters of the same form, fit, and function. The rod control step counters will be mounted in the existing step counter housings in panels 1PM05J and 2PM05J. The existing rod in and rod out lamps will also be replaced with new LEDsnamps that are compatible with the 24 VDC circuit. Since the portion of the circuit that is being modified from 100 VDC to 24 VDC is also used by the loose parts monitoring system (LPMS), the LPMS circuitry will be modified to accommodate this change.

The rod bank position output signals to plant process computer (PPC) cabinet I CX02J and 2CX02J will be abandoned. This data will now be provided to the PPC instead via the Modbus interface from Ovation to the PPC, installed as part of a separate activity


Station/Unit(s): Braidwood Units I and 2

Page 2 of 36

LS-AA-104-1001 Revision 4 Page 2 of3

Activity/Document Number: EC 400433 and EC 400435 I DRP 16-023 and DRP 16-035 Revision Number: 002 and 002 I OandO

Title: Westinghouse Ovation Digital Upgrade for Rod Control Logic Cabinet CN-1 Outage)

(EC 400918 and EC 400920). The rod supervision PPC software will be revised by this activity to support rod bank position data being delivered to it via the Modbus interface to Ovation rather than the existing method of insert/withdraw pulses being read by the PPC 110 system. In addition, rod bank position information for group 1 as well as group 2 rod positions will be used in the program, whereas the existing program only uses group 1 data.

DRP 16-023 (Unit 1) and DRP 16-035 (Unit 2) will revise UFSAR Section 7.7.1.2.2 to incorporate changes associated with the Ovation-based control system upgrade.

Revision 1 of the 50.59 Screening, BRW-S-2015-091, has been performed to clarify tl1at this activity does not pose a departure from a UFSAR-described method of evaluation.

Revision 2 of the 50.59 Screening, BRW-S-2015-091, and Revision I of the 50.59 Evaluation, BRW-E-2015-092, have been · perfom1ed to further address the potential for common cause failures.

Reason for Activity: (Discuss why the proposed activity is being performed.)

The upgrade of the rod control logic cabinet is one of several upgrades of non-safety-related process control systems being performed because ilie existing 7300-series control systems have exhibited performance, maintenance and reliability issues. It has been determined that tl1e appropriate long term solution is to upgrade the existing 7300-series equipment to modern digital controls to enhance its reliability and overall lifespan.

Effect of Activity: (Discuss how the activity impacts plant operations, design bases, or safety analyses described in the UFSAR.)

The basic operation of the rod control system will not be changed. The upgrade of tlle logic cabinet will involve minor changes to the operator interface with the rod control system (e.g., slight differences in the appearance of the step counter displays and in-out lights).

The functional and performance characteristics of the existing rod control logic cabinet will be maintained. Therefore, the proposed activity does not impact 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.)

The proposed activity involves tlle replacement of components within the rod control logic cabinets (1 RD07J and 2RD07J) as part of an overall phased project to upgrade key non-safety-related process control systems, integrating them into an Ovation-based dist.Jibuted control system. The upgrade is being performed in accordance with the recommendations of the original equipment manufacturer (Westinghouse) to address equipment reliability and obsolescence issues.

Evaluations of the Ovation platform and its successful use in non-safety-related process control applications in the nuclear industry indicate that the system is expected to perform dependably. The application-specific software is being provided by Westinghouse - the original equipment supplier, who is employing their established product development processes used for the nuclear industry - to match the existing functional and performance characteristics of the rod control logic cabinet. The basic operation of the rod control system will not be changed, although the upgrade will involve minor changes to the operator interface with the rod control system. Therefore, the operator interface with the upgraded rod control system does not fundamentally change or adversely affect the methods used by the operator in performing or controlling a UFSAR-described design function. However, the upgraded rod control logic cabinet combines functions previously performed by multiple cards into redundant controllers, introducing the potential for failures to create new malfunctions which could adversely affect a UFSAR described design function. In addition, connecting the upgraded logic cabinet to a network linked to other plant control system introduces


Station/Unit(s): Braidwood Units l and 2

Page 3 of 36

LS-AA-104-1001 Revision4 Page 3 of3

Activity/Document Number: EC 400433 and EC 400435 I DRP 16-023 and DRP 16-035 Revision Number: 002 and 002 / OandO

Title: Westinghouse Ovation Digital Upgrade for Rod Control Logic Cabinet (N-1 Outage)

the potential for a software-based common cause failure that could adversely affect multiple systems. Therefore, 50.59 Evaluation BRW-E-2015-092 was performed to address these issues.

The 50.59 Evaluation determined that the redundant hardware and improved diagnostics combined with the reliability of the new equipment and software and the fact that single failures remained bounded by existing single failures (as determined in a failure modes and effects analysis) did not result in a more than minimal increase in the frequency of occurrence of an accident previously evaluated in the UFSAR or in a more than minimal increase in the likelihood of occurrence of a malfunction of an SSC important to safety previously evaluated in the UFSAR.

The 50.59 Evaluation also found that an analysis of the susceptibility of both the Ovation platform and the DCS applicationspecific design features to potential digital system failures - including software-based common cause failures - was performed using the guidance of EPRI Technical Report 3002005326 ("Methods for Assuring Safety and Dependability when Applying Digital Instrumentation and Control Systems"). In accordance with the EPRI guidance, the analysis considered the following potential sources of failure: random hardware failure, environmental disturbances, design defects, and operations or maintenance errors. Since individual device hardware failures had been previously addressed in a failure modes and effects analysis, the susceptibility analysis addressed hardware failures in shared resources. The analysis of design defects as a potential source of failure included a consideration of software-based common cause failures. The susceptibility analysis reviewed the design with respect to the various EPRl-identified preventive and limiting measures to determine whether these measure were sufficient to reduce the likelihood of a common cause failure to "Level 2", which is defined in the EPRI guidance as a failure likelihood at or below that of failures considered sufficiently unlikely that they would not typically be postulated and analyzed as part of a plant safety analysis report (e.g., random hardware failures or common cause failures of identical components caused by wear out and aging or by design or maintenance errors). The susceptibility analysis found that the design was adequate to reduce the likelihood of common cause failures in the Ovation platform or the specific DCS application to the desired level (i.e., EPRI Level 2}.

On the basis of the susceptibility analysis, the 50.59 Evaluation concluded that the increase in the likelihood of accidents or malfunctions due to a software common cause failure is no more than minimal, and that the activity does not create the possibility of an accident or malfunction of a different type. The radiological consequences of previously evaluated accidents or malfunctions were not affected.

The upgrade of the rod control logic cabinet in accordance with OEM recommendations does not involve a method of evaluation described, outlined, or summarized in the UFSAR that is used in the safety analyses or used to establish the design bases, nor does it involve a test or experiment not described in the UFSAR. The upgrade of the logic cabinet system does not affect the Technical Specifications or the Facility Operating License.

Therefore, the proposed 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.

BRW-S-2015-091 Rev. 002 -------BRW-E-2015-092 Rev. 001 -------

See LS-AA- I 04, Section 5, Documentation, for record retention requirements for this and all other 50.59 forms associated with the Activity.

of 36

50.59 REVIEW COVERSHEET FORM LS-AA- I 04- l 00 I Revision 4 Page I of2

Activity/Docmnent Number: Modification EC 405740 and EC 405741 Revision Number: 0 I 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:

Modification EC 405740 and EC 405741 wHl relocate the combustion air intake of the Unit 1 and Unit 2 diesel dtiven Auxiliary Feedwater (AF) pumps (1/2AF01PB) from the 401' elevation of the turbine building to the auxiJiary building 364' elevation general area. The relocation will utilize the existing 'L' wall penetration at the 391' elevation and re-route the intake piping through a spare 'L' wall penetration at the 376' elevation. The existing intake piping routed to the 40 l' elevation of the turbine building will be removed.

Reason for

The original 1/2AF01PB diesel driven pump combustion air intake (l/2DOB lA-14) took suction from the turbine building. During the 2016 Braidwood NRC Component Design Basis Inspection (CDBI), the inspection team requested information supporting the acceptability of locating the combustion air intake for the B-train AF diesel engines in the turbine building. During the review of available·documentation related to the B-train AF diesel engine air intake, it was identified that the documentation did not suppo1t operation of the diesel with High Energy Line Break (HELB) environmental conditions postulated in the turbine building.

Effect of Activity:

The change will allow proper operation of the diesel ddven AF pumps during normal and accident conditions. However, the proposed activity will result in air from the 364' elevation general area of the auxiliary building being used for combustion air for the diesel engine and exhausted directly to the outside, bypassing the auxiliary building Ventilation (VA) exhaust flow paths, which are monitored effluent release pathways. The new release paths are unmonitored and unfiltered effluent release pathways, therefore, alternate monitoring capability is required to be implemented.

Summary of Conclusion for the Activity's 50.59 Review:

The proposed activity will maintain the design functional performance requirements assumed in the UFSAR for the diesel driven AF pumps. Although no physical changes are made to the VA exhaust filtration SSC's, the proposed activity will affect the UFSAR described design features of the VA exhaust system for precluding direct exfiltration of contaminated air from the auxiliary building following an accident which could result in abnormally high airborne radiation in the auxiliary building. Although it was determined that the existence of these unfiltered release pathways being created did not impact the radiological dose consequences following a Loss of Coolant Accident (LOCA), this was determined to be an adverse impact requiting a 10 CFR 50.59 Evaluation.

The 10 CPR 50.59 Evaluation concluded that the operation of the diesel driven AF pumps and the VA system were not degraded, the activity has no impact on the potential for or consequences of malfunctions of SSC's important to safety. Also, since the AF and YA systems are not accident initiators, there is no increase on the likelihood of an accident previously evaluated in the UFSAR from occuning. As indicated above, since the proposed activity was evaluated to have no impact on the radiological dose calculations, it was determined to have no impact on the consequences of an accident previously evaluated in the UFSAR. As failure scenarios for the AF and VA systems

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50.59 REVIEW COVERSHEET FORM LS .. AA-104-1001 Revision 4 Page 2 of2

Activity/Document Number: Modification EC 405740 and EC 405741 Revision Number: 0 I 0

are unaffected, the proposed activity does not create the potential for an accident of a different type than previously evaluated in the UFSAR. As the modified SSC's pe1form a passive function, no new malfunctions of SSC's important to safety with different results than any previously evaluated in the UFSAR were identified. The modified SSC's do not result in operation that can impact the containment, Reactor Coolant System. or fuel cladding, therefore, the proposed activity does not result in a fission product barrier being exceeded or altered. FinaJly, no methods of evaluation not described in the UFSAR were utilized to evaluate the proposed activity.

Based on this evaluation, the proposed activity requires a IO CFR 50.59 evaluation, however, the resulting evaluation determined that the proposed activity can be pe1formed without prior NRC permission per the applicable governing procedures.




50.59 Screening

50.59 Evaluation

50.59 Screening No.


BRW-S-2016-70 Rev. _o _____ _ BRW-E-2016-71 Rev. _o _____ _

See LS-AA- I 04, Section 5, Documentation, for record retention requirements for this and all other 50.59 fonns associated with the Activity.

of 36

50.59 REVIEW COVERSHEET FORM LS-AA-104-1001 Revision4 Page 1 of3

Station/Unit(s): Braidwood Unit 2

Activity/Document Number: =E~C_4~0~0~9~1~9 _______________ _ Revision Number: 002

Title: Westinghouse Ovation Digital Upgrade for DEH (N-1 Outage)



The proposed activity involves the upgrade of the Ovation-based digital electro-hydraulic (DEH) turbine control system in accordance with the recommendation of the original equipment manufacturer (Westinghouse). The existing OCR 161 controllers will be replaced vii.th OCR 1100 controllers, and the operator workstations in the Main Control Room will be replaced. This activity also introduces a Heater Isolation Runback "soft" button as part of tl1e DEH system user interface. The upgrade will also eliminate failure modes and effects in turbine stop and control valve positioning components that have been identified in the industry in earlier Ovation systems. Minor changes will be made to tlle power supplies in the generator cooling cabinet.

Revision l of the 50.59 Screening, BRW-S-2016-168, has been performed to fmther address the potential for common cause failures. Based on the results of that screening, a 50.59 Evaluation, BRW-E-2016-089, has been performed.


The turbine controls were previously upgraded to a DEH system which uses the Ovation platform. The proposed activity involves a second upgrade to make tlle DEH system compatible with upgrades of other plant control systems to a more recent Ovation platform (under separate activities). The proposed heater isolation runback feature has been requested by Operations, as part of tlle resolution to an INPO area for improvement (AFI), to improve the plant response to a "HI-2" level alarm in a 25 or 26 feedwater heater.


The proposed activity involves the replacement and reconfiguration of select components in the Ovation-based DEH system to make that system compatible with a later version of the Ovation platform that will be used for upgrading the DEH and other nonsafety-related process control systems and incorporating them into an Ovation-based distributed control system (DCS). The proposed activity also includes additional minor changes to the DEH system. The upgrade will involve slight changes to t11e operator interface with the DEH system (e.g., newer workstations). The new runback feature, which will ramp the turbine power down at 100 MW/Min until an operating level of 1050 MWe is reached, is intended to assist the operators in rapidly reducing power following receipt of a HI-2 level alarm for the 25 or 26 feedwater heaters in order to prevent a reactor trip due to the loss of feedwater heating. The net decrease from full power has been selected to bound the power reduction required once the cascading effects of heater shell-side isolation due to the HI-2 alarm actuation are taken into account, as specified in BwOP HD-6Tl. The new runback feature is similar to existing runback features for the loss of a main feed water or condensate/booster pump.

The functional and performance characteristics of tl1e existing turbine control system will be maintained. The proposed activity does not impact the design bases or safety analyses described in the UFSAR.


The proposed activity involves the replacement and reconfiguration of select components in the Ovation-based DEH system, in accordance with the recommendations of the OEM (Westinghouse), in order to incorporate that system into an Ovation-based distributed control system (DCS) that will be used for tlle DEH and other non-safety-related process control systems. The overall Ovation-based DEH control system architecture and its capabilities are being maintained, and the upgrade of the DEH system will reproduce the existing turbine control system functional and performance characteristics using the new controllers, workstations, and associated software.

50.59 REVIEW FORlVI


Page 7 of 36


Activity/Document Number: =E=C,_4~0~0""9-"1~9 _______________ _ Revision Number: 002

Title: Westinghouse Ovation Digital Degrade for DEH CN-1 Outage)

Evaluations of the Ovation platform and its successful use in non-safety-related process control applications in the nuclear industry indicate that the minor upgrades to the DEH system will not affect its dependability. The application-specific software is being provided by Westinghouse - the original equipment supplier, who is employing their established product development processes used for the nuclear industry- to match the existing functional and performance characteristics of the DEH system. As a result, the modified Ovation platform and the upgraded DEH controls are expected to perform dependably. However, connecting the DEH system to a network linked to other plant control systems introduces the potential for a software-based common cause failure that could adversely affect multiple systems. This was considered a change to an SSC that adversely affects a UFSAR-described design function; therefore, a 50.59 evaluation was performed.

The 50.59 Evaluation found that an analysis of the susceptibility of both the Ovation platform and the DCS application-specific design features to potential digital system failures - including software-based common cause failures - was performed using the guidance of EPRl Teclmical Report 3002005326 ("Methods for Assuring Safety and Dependability when Applying Digital Instrumentation and Control Systems"). In accordance with the EPRI guidance, the analysis considered the following potential sources of failure: random hardware failure, environmental disturbances, design defects, and operations or maintenance errnrs. Since individual device hardware failures had been previously addressed in a failure modes and effects analysis, the susceptibility analysis addressed hardware failures in shared resources. The analysis of design defects as a potential source of failure included a consideration of software-based common cause failures. The susceptibility analysis reviewed the design with respect to the various EPRI-identified preventive and limiting measures to determine whether these measures were sufficient to reduce the likelihood of a common cause failure to "Level 2". Level 2 is defined in the EPRI guidance as a failure likelihood at or below that of failures considered sufficiently unlikely that they would not typically be postulated and analyzed as part of a plant safety analysis report (e.g., random hardware failures or common cause failures of identical components caused by wear out and aging or by design or maintenance errors). The susceptibility analysis found that the design was adequate to reduce the likelihood of conunon cause failures in the Ovation platform or the specific DCS application to the desired level (i.e., EPRl Level 2).

On the basis of the susceptibility analysis, the 50.59 Evaluation concluded that the increase in the likelihood of accidents or malfunctions due to a software common cause failure is no more than minimal, and that the activity does not create the possibility of an accident or malfunction of a different type. The radiological consequences of previously evaluated accidents or malfunctions were not affected.

The minor changes to the operator interface with the upgraded DEH system do not involve a change to a procedure that adversely affects how UFSAR-described SSC design functions are performed or controlled. The new heater isolation runback feature is similar to existing turbine runback features, which allow for a rapid reduction in turbine and reactor power to prevent an upset condition from leading to a reactor trip, while remaining well within the load rejection capability of the reactor control and steam dump control systems. Therefore, this aspect of the proposed activity does not adversely affect a UFSAR-described design function. The runback on high-high heater level simplifies and consolidates existing actions by providing a single button with a pre-programmed ramp rate and power level to bound the power reduction required once the cascading effects of heater shell-side isolation are taken into account. The new runback feature will remain a manual action but will reduce the operator burden, with no new potential failure modes in the interaction of operators with the plant systems. Therefore, in accordance with the guidance provided in Section 5.2.2.2 of LS-AA-104-1000 (50.59 Resource Manual), this aspect of the proposed activity does not involve a change to a procedure that adversely affects the method of performing or controlling a UFSAR-desciibed design function.

The upgrade of the DEH system in accordance with OEM recommendations and the introduction of the new nmback feature do not involve a method of evaluation described, outlined, or summarized in the UFSAR that is used in the safety analyses or used to establish tl1e design bases, nor do they involve a test or experiment not desciibed in the UFSAR. The upgrade of the DEH system does not affect the Technical Specifications or the Facility Operating License.

Therefore, NRC approval is not required for the proposed activity. The activity may be implemented per the applicable governing procedure.




Page 8 of 36

LS-AA-104-1001 Revision 4 Page 3 of 3

Activity/Document Number: =E=C_4~0~09~1=9 ______________ _ Revision Number: 002

Title: Westinghouse Ovation Digital Upgrade for DEH (N-1 Outage)



50.59 Screening

50.59 Evaluation

50.59 Screening No.


BRW-S-2016-168

BRW-E-2016-089

Rev. 001 -------Rev. 000 -------

See LS-AA-104, Section 5, Documentation, for record retention requirements for this and all other 50.59 forms associated with the Activity.


Page 9 of 36

L..<i·AA- I04- IOO I Revision 4 Page I of4

Activity/Document. Number: EC 400277/EC 399174/DRP 16-012 Revision Number: 000 I 000 /000

Title: Revising AST Accident Dose CalculaLions

NOTE: For 50.59 Evaluations, informalion 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).


This activity will implement Engineering Changes (EC) 400277 (BYR), EC 399174 (BRW) and UFSAR change DRP 16-012. The above ECs incorporate the four revised design analyses listed below. The analysis determines the radiological consequences using Alternative Source Term (AST) methodology. The potential accidents for which Control Room (CR), Exclusion Area Boundary (EAB) and Low Population Zone (LPZ) doses have been recalculated for Byron and Braidwood arc as follow:

a) BYR04-051/BRW-04-0038-M. Rev. 5. "Re-analysis of Loss Of Coolant Accident (LOCA) Using Alternative Source Terms".

b) BYR04-045 & BRW-04-0039-M, Rev. 4, "Re-analysis of Control Rod Ejection Accident (CREA) Using Alternative Source Tenns".

c) BYR04-047/BRW-04-0041-M. Rev. 4. "Re-analysis of Fuel Handling Accident (FHA) Using Alternative Source Term".

d) BYR04-049/BRW-04-0043-M. Rev. 4, "Re-analysis of Locked Rotor Accident (LRA) Using Alternative Source Term".

The proposed activity also involves UFSAR change (DRP 16-012) to incorporate the revised doses.


RADTRAD is the computer code used for the Analysis Of Record (AOR) listed above. An error was discovered in the version of RADTRAD that was used in the AOR for the LOCA dose described in UFSAR 15.6.5 (IR 01320861 ). The correction to this error required reanalysis which resulted in use of the latest calculated Off site Atmospheric Dispersion Factors (X/Q) due to the commitment made to NRC (UFSAR 2.3.6.3). Based on this commitment, the

X/Q values needed to be re-evaluated based on the finer wind speed categories provided in the latest Regulatory Guidance (RG) 1.23, Revision 1, the next time ca1cu1ations associated with the dose consequences for the LOCA, MSLB. CREA, LRA, SGTR and FHA were revised. In support of the License Amendment Request (LAR) for Measurement Uncertainty Recapture (MUR) Power Uprate, Exe1on performed a re-evaluation of the offsite XIQ values for the radiological dose analyses for the Main Steam Line Break (MSLB) and Steam

Generator Tube Rupture (SGTR) accidents. Hence, the new Offsite xJQ values needed to be incorporated in the accident dose analyses for the LOCA, CREA, LRA and FHA due to the commitment made to the NRC.


This activity does not involve any physical change, nor has any impact on plant operations, nor does it impact the initiation or progression of any accidents. It only changes the calculated dose following postulation of the accidents discussed above.

An en·or was discovered in the version of RADTRAD that has been used in the AOR described in UFSAR 15.6.5. The correction to this error resulted in the analyzed dose to be higher (conservative) than those

of 36

50.59 REVIEW COVERSHEET FORM LS-AA-104-1001 Revision 4

Page 2 of4 Stution/Unit(s): Jl.YIQH anti Braidwootjf J,J.11illi 1&.L

Activity/Document Number: EC 4Q0277 /E~J.99 I 7 4/.DRP 16-012 Revision Number: 000 I 000 /000

previously calculated. The correction in the RADTRAD code is lo a design clement. Per NEI 96-07 Revision L "In general, licensees can make changes to elements of a methodology without first obtaining a license amendment if the results are essentially the same as, or more conservative than, previous results." As shown in Lable below, the results are conservative and remain within the limits of IO CFR 50.67(b)(2) and RG 1.183.

The resultant doses from the existing and the re-analyzed Accident Radiological Dose Analyses and the limits of I 0 CPR 50.67(b)(2) and RG l. l 83 are listed in the table below:

Rem TEDE

Accident Dose CR EAB LPZ

Existi!:~ t-4.78 2.99 -·· -

New 'l(. ¥* 4.76

a. LOCA

Limit I 5 25

Existing ·~ _4.s~sL 4.647 1.983

..

New 4.53sl * 5.358 * 2.278

Limit 5 6.3 6.3

b. CREA

Existing .~:~.tl ... 4.24 0.356

New __ 4.2~L ___ *. 4.s9 * 0.87 ·- ~-·- . " --·· ---- -

Limit 51 6.3 6.3

c. FHA

d. LRA Existing 2.79 1.456 0.525

New 2.79 * 1.679 * 0.602

limit sl 2.5 2.5

NOTEs: "*" The dose increase is due to use of new Off site Atmospheric Dispersion factor (XIQ). In accordance with this commitment, the 'X/Q calculation was revised to re-evaluate the offsite 'X/Q values, as related to the use of the PAV AN computer model based on finer wind speed categories provided in the latest NRC RG 1.23, "Meteorological Monitoring Programs for Nuclear Power Plants;" Revision I, March 2007. As argued in the question number 3 of the screening criteria, the change is considered a change to an element of methodology. As specified in Reference 3, the NRC staff performed a qualitative review of the inputs and assumptions used in the licensee's PAVAN computer calculations and of the resulting x/Q values. The staff calculated comparative xtQ values, and found the results to be similar to the EAB and LPZ x/Q values calculated by the licensee. On the basis of this review, the NRC staff determined that the resulting offsite EAB and LPZ XIQ values for the MSLB and SGTR generated by the licensee and presented in Table 3.3.1-6 of Reference 3 SE are acceptable for use in making DBA dose assessments.

"¥" The dose increase is due to change in an element of methodology to correct RADTRAD error in modeling (IR 01320861). The RADTRAD enor applies to the LOCA analysis, because this analysis uses modeling method that combines the release into two compartments, representing the sprayed and unsprayed regions within the structure in single run. The LOCA re-analysis uses a separate RADTRAD model for each containment compartments leakage dose contributions (sprayed and unsprayed regions), based on the software issuer suggested modeling method to compensate for the code internal en-or. The dose results from the two tuns were added to determine the total calculated doses associated with the accident.


Smtion/Unit(s}: Byron and Braidwood I Units I &=' ~2 ____________ _

Page 11 of 36

LS-AA-I04-IOOI Revision 4 Page 3 of4

Activity/Document Number: EC 400277/EC 399 I 74/DR.~P~l~6-~0~l 2=------ Revision Number: 000 I 000 1000

Title: Revising AST Accident Dose Calculations

Arkansas Nuclear One performed the LOCA dose reanalysis using lhe exacl methodology as above lo compensate for the RADTRAD en-or. As noted in the December 24, 2013 letter (ML! 3326A502-Arkansas Nuclear One, Units l and 2, Safety Evaluation Related to RADTRAD E1rnr) the NRC staff reviewed the methods, parameters, and assumptions that the licensee used in its LOCA radiological dose consequence analyses for ANO-I and AN0-2 and concludes that they are consistent with the conservative guidance provided in RG 1.183. The NRC staff used the RADTRAD computer code to perform an independent confirmatory dose evaluation lo ensure an understanding of the licensee's methods. The NRC staff concludes that summing the results of multiple runs, as described by the software issuer to cotTcct the error. is an acceptable approach for correcting the RADTRAD error because it will account for all the nuclides available for release. The NRC staff evaluated the radiological consequences resulling from the postulated LOCA using the AST and concluded that the radiological consequences at the EAB and LPZ and in the CR are within the dose crite1ia specified in 10 CFR 50.67. Therefore. this change is acceptable with respect to the radiological consequences of DBAs.

References: I. Letter from Craig Lambert (Exelon Generation Company, LLC) to U. S. NRC, "Request for License Amendment Regarding Measurement Uncertainty Recapture Power Uprate," dated June 23, 2011 (ML! 11790030).

2. Letter from N. J. DiFrancesco (U.S. NRC) to M. J. Paci1io (Exelon Generation Company, LLC), "Braidwood Station, Unit Nos. 1 and 2 and Byron Station, Unit Nos. 1 and 2 - Supplemental Information Needed for Acceptance of Licensing Action Regarding Measurement Uncertainty Recapture Power Uprate (TAC Nos. ME6587. ME6588. ME6589. and ME6590)," dated August 22. 2011 (MU 12150563)

3. Letter from Joel. S. Wiebe (U.S. NRC) to M. J. Pacilio (Exelon Generation Company, LLC), "Braidwood Station, Unit Nos. 1 and 2 and Byron Station, Unit Nos. 1 and 2 - Issuance of Amendments Regarding Measurement Uncertainty Recapture Power Uprate (TAC NOS. MF2418, MF2419, MF2420, AND MF242l)," dated February 7, 2014 (ML13281AOOO).

4. NRC Approval letter dated December 24, 2013-Arkansas Nuclear One, Units l And 2-Safety Evaluation Related to Revised Dose Consequences Based on Alternate Source Term (MLl 3326A502).

Summary of Conclusion:

(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 will implement Engineering Changes (EC) 400277 (BYR), EC 399174 (BRW) and UFSAR change DRP 16-012. The above ECs incorporate four (4) revised design analyses determining the radiological consequences using Alternative Source Term (AST) methodology. The potential accidents for which Control Room (CR), Exclusion Area Boundary (EAB) and Low Population Zone (LPZ) doses have been recalculated for Byron and Braidwood are as followed: (a) BYR04-05l/BRW-04-0038-M - Re-analysis of Loss Of Coolant Accident (LOCA) Using Alternative Source Terms, Rev. 5. (b) BYR04-045 & BRW-04-0039-M, "Re-analysis of Control Rod Ejection Accident (CREA) Using Alternative Source Terms", Rev. 4. (c) BYR04-047/BRW-04-0041-M, "Re-Analysis of Fuel Handling Accident (FHA) Using Alternative Source Term", Rev. 4. (d) BYR04-049/BRW-04-0043-M, "Re-Analysis of Locked Rotor Accident (LRA) Using Alternative Source Term", Rev. 4.


Page 12 of 36


Activity/Document Number: EC 400277/!;i(:_J99174/DRP !6-0~-- __ _ Revision Number: 000 LQOQJ.900 _

The revised calculated Total Effective Dose Equivalent (TEDE) is within the limits specified in I OCFR 50.67. and RO 1.183 (as shown in the table above).

The e1rnr discovered in the version of RADTRAD that has been used in the AOR described in the applicable sections of the chapter 15 in the UFSAR. The correction to this error resulted in the analyzed dose to be higher (conservative) than those previously calculated. The correction in the RADTRAD code is to a design element. Per NEI 96-07 Revision I. "In general. licensees can make changes to elements of a methodology without first obtaining a license amendment if the results are essentially the same as, or more conservative than. previous results." As shown in table above, the results are conservative and remained within the limits of IO CFR 50.67(b)(2) and RO 1.183.

There are no changes by the Activity to the manner in which the plant is operated or controlled. This change is strictly analytical in nature. thus il does not constitute a test or experiment. There arc no changes to the Technical Specifications or the Operating License required. Because this activity results in higher doses calculation for the affected accidents due to changes in elements of the methodology the Screening question number 3 is answered "Yes". The remaining Screening questions are answered "No" the activity screens in and a full 50.59 Evaluation is required.

Attachments:

Attach all 50.59 Review forms completed, as appropriate.


D Applicability Review

50.59 Screening

50.59 Evaluation

50.59 Screening No.


(BYR) 6E-15-043/ BRW-S-2015-74 (BYR) 6G-I 6-002/ BRW-E-2016-22

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 this activity.



Page 13 of 36


Activity/Document Number: EC 617642 Revision Number: Q

Title: Temporarily Disable FW Water Hammer Prevention System (WBPS) When 2FW009A/B/C/D Are Open (Above Approx. 25% Power) For Steam Generators 2A/2B/2C/2D

Including EC 617642 Supporting Procedure Revisions: 2BwGP 100-1 Rev. 34; 2BwGP 100-3 Rev. 71: 2BwGP 100-4 Rev. 42; 2BwGP 100-4T4 Rev. 01; 2BwGP 100-lTl Rev. 16; 2BwGP 100-4Tl Rev. 18 and 2BwGP 100-4T5 Rev 1


Description of Activity: (Provide a brief, concise description of what the proposed activity involves.) This Temporary Configuration Change (TCC) Engineering Change (EC) 617642 will disable the Water Hammer Prevention System (WHPS) low Steam Generator (SG) level and low SG pressure feedwater isolation signals on Unit 2 after Feed water (FW) flow is transferred from the FW isolation valve bypass valves, 2FW043A/B/C/D, to the main feedwater isolation valves, 2FW009A/B/C/D, (when power is above approximately 25% power).

The proposed change to disable the WHPS FW isolation involves installing an electric switchable jumper between two terminal block points for each of the 2FW009A/B/C/D and 2FW039A/B/C/D valves, and repositioning the control switch from "Auto" to "Open" for each of the 2FW035A/B/C/D valves. The WHPS FW isolation signals to the 2FW043A/B/C/D valves are not disabled.

The following changes will be made to plant operating procedures:

• 2BwGP 100-1, PLANT HEATUP, will be revised to verify that WHPS FW isolation has been enabled prior to starting any FW pumps.

• 2BwGP 100-3, POWER ASCENSION, will be revised to disable WHPS FW isolation after the 2FW009A/B/C/D valves are open.

• 2BwGP 100-4, POWER DESCENSION, will be revised to enable WHPS FW isolation for the 2FW035A/B/C/D and 2FW039A/B/C/D valves, close the 2FW009A/B/C/D valves, and then perform a follow-up action to enable WHPS FW isolation on the 2FW009A/B/C/D valves.

• 2BwGP 100-4T4, REACTOR TRIP POST RESPONSE GUIDELINE, will be revised to verify that the 2FW009A/B/C/D, 2FW043A/B/C/D, and 2FW039A/B/C/D valves are closed prior to starting the Startup Feedwater Pump. Action will also be taken to enable WHPS FW isolation later in the post trip response.

• Associated Flowchart procedures 2BwGP 100-lTl, 2BwGP 100-4Tl and 2BwGP 100-4T5 will be revised as necessary to reflect updated step numbers.

Reason for Activity: (Discuss why the proposed activity is being performed.) The steam generator water hammer prevention relays are interlocked with the FWIV "Close" circuit<;. The relays are normally energized for the low SG level and low SG pressure permissive circuit on each FWIV. On a loss of power to the relays, the result is closure of the associated FWIV and a reactor trip will follow. Also note that the WHPS relays are powered by NonSafety Related (NSR) power. The proposed change will allow the SG water hammer prevention actuation circuit to be disabled once feedwater isolation valves have been opened.

The changes to the plant operating procedures are made to administratively control when the WHPS FW isolation function is disabled and enabled.


The safety function of the Feedwater Isolation Valves is to close, isolating main feedwater flow upon a safety injection signal, feedwater isolation signal and high-2 SG level. As described in UFSAR section 10.4.7.3, "Several water hammer prevention features have been designed into the feedwater system. These features are provided to minimize the possibility of various water



Page 14 of 36



Title: Temporarily Disable FW Water Hammer Prevention System (WHPS) When 2FW009A/B/C/D Are Open (Above Approx. 25% Power) For Steam Generators 2A/2B/2C/2D

Including EC 617642 Supporting Procedure Revisions: 2BwGP 100-1 Rev. 34; 2BwGP 100-3 Rev. 71; 2BwGP 100-4 Rev. 42; 2BwGP 100-4T4 Rev. 01; 2BwGP 100-lTl Rev. 16; 2BwGP 100-4Tl Rev. 18 and 2BwGP 100-4T5 Rev 1

hammer phenomena in the steam generator preheater, steam generator main feedwater inlet piping, and the steam generator upper nozzle feedwater piping." Steam generator (two-out-of-three logic) low level trips are provided to close all feedwater isolation valves, feedwater isolation bypass valves and feedwater preheater bypass valves. Steam generator (two-out-of-three logic) low pressure trips are provided to close all feedwater isolation valves, feedwater isolation bypass valves, feedwater preheater bypass valves, and the feedwater bypass tempering valves.

The UFSAR described water hammer prevention features were included in the NRC's basis for approval of (1) the designs of the Steam Generator and the FW/AFW systems, and the measures to address Unresolved Safety Issue (USI) A-I, Steam Generator Water Hammer; and (2) the removal of pipe restraints resulting from the elimination of arbitrary intermediate breaks in the FW system piping system design.

The purpose of the WHPS FW isolation is to prevent water hammer by precluding the admission of cool FW into a steamvoided preheater section of the steam generator (when SG narrow range level is less than 5% or when SG pressure is less than 600 psig). When the WHPS FW isolation is disabled at power the water hammer prevention function will be fulfilled by the safety related ESFAS signals function to close the feedwater isolation valves on low steam line pressure safety injection or steam generator water level low-low reactor trip coincident with Tave less than 564° F. Note that the 2FW039 valves close on reactor trip and no coincident Tave is required. At low power levels or during startup the WHPS FW isolation features will be enabled. The design bases function to preclude the admission of cool FW into a steam-voided preheater section of the steam generator is not adversely impacted.

Additionally, the other UFSAR described WHPS features will not be altered by this EC. Operating procedures associated with only providing feedwater to the upper nozzle of the steam generator during startup and low load conditions are unchanged. Temperature monitoring and alarms to detect possible back leakage of steam from the steam generators into the feedwater piping are unchanged. Forward flushing and temperature permissive interlocks associated with opening the FWIV are also unchanged. This EC does not alter split feed water flow operation.

Implementation of EC 617642 does not impact the safety related circuits described in the UFSAR and Technical Specifications that generate an independent Feedwater Isolation Signal, including Safety Injection (SI), SG Level HI-2 Trip (P-14), and Reactor Trip (P-4) coincident with RCS Average Temperature Low at 564°F.

The change will require new operator actions to appropriately enable/disable WHPS FW isolation in the plant startup, power ascension, power descension, and plant uip response procedures. Operators will reposition the control board hand switch for the 2FW035 AIB/CfD valves and reposition the jumper switch for the 2FW009NB/CfD and 2FW039A/B/CfD valves.

• For startup WHPS FW isolation will be verified to be enabled prior to starting any FW pumps. Once the 2FW009 valves are open, action will be taken to disable WHPS FW isolation for the 2FW009, 2FW035, and 2FW039 valves.

• For a normal shutdown or low power operation action will be taken to re-enable WHPS FW isolation for the 2FW035 and 2FW039 valves prior to closing the FW009 valves. WHPS FW isolation for the 2FW009 valves will be enabled prior to defeating the ESFAS FW isolation signals.

• On a plant trip action will be taken to verify that the 2FW009, 2FW039, and 2FW043 valves are closed prior to starting the Startup FW pump. The 2FW035 valve flow path is used to direct FW to the upper nozzle. The UFSAR described main FW flow and temperature permissive interlocks will prevent inadvertent introduction of cold FW to the lower nozzle until action is taken to enable WHPS FW isolation.



Page 15 of 36



Title: Temporarily Disable FW Water Hammer Prevention System (WHPS) When 2FW009A/R/C/D Are Open (Above Approx. 25% Power) For Steam Generators 2A/2B/2C/2D

Including EC 617642 Supporting Procedure Revisions; 2BwGP 100-1 Rev. 34; 2BwGP 100-3 Rev. 71; 2BwGP 100-4 Rev. 42:].BwGPJQQ~4T4 Rey. 01; 2BwGP 100-lTl.Rev. 16: 2BwQP 1.00-4Tl Rev. 18 and 2BwGP 100-4T5 Rev 1

There is no impact on any emergency operating procedures. The proposed operator actions are not time critical and do not impose a significant new burden on the operating crew.


Disabling the WHPS low SG level and low SG pressure FW isolation auto close signals above 25% power does not introduce the possibility of a change in the frequency of an accident because WHPS is not an initiator of any accident and no new failure modes are introduced. Disabling the WHPS low SG level and low SG pressure FW isolation auto close signals above 25% power reduces the probability of a loss of feedwater event due to a spurious WHPS actuation.

In the plant operating conditions under which the WHPS automatic FW isolation features may be procedurally disabled, the safety related ESFAS signals function to close the feed water isolation valves on a low steam line pressure safety injection ( < 640 psig) or steam generator water level low-low reactor trip($ 36.3% NR) coincident with low Tave (less than 564° F). Note that the 2FW039A/B/C/D valves close on reactor trip and no coincident Tave is required. A review of the postulated transients that could create conditions for a water hammer event when power is above 25% was performed (Refer to EC 617642). The review determined that the ESFAS FW isolation signals (Steam line pressure SI or the Reactor trip+ Lo Tave) will function to prevent water hammer by precluding the admission of cool FW into a steam-voided preheater section of the steam generator during power operating conditions. Therefore disabling the WHPS low SG level and low SG pressure FW isolation auto close signals above 25 % power does not change the possibility of water hammer in the SG preheater and the main FW inlet piping. Thus the proposed change does not result in more than a minimal increase in the frequency of occurrence of a FW line break or steam generator tube rupture.

The defeated WHPS FW isolation function is not used or credited to mitigate any design basis event, analysis, or Emergency Operating Procedure Operator Action. The credited automatic function is the ESFAS FW Isolation which is a completely separate circuit. Thus ESFAS FW Isolation is not affected by this change.

The change adds new operator actions to disable and enable the WHPS low SG level and low SG pressure FW isolation auto close signals. Failure to enable the WHPS low SG level and low SG pressure auto close signals for strutup, low load, or increasing load conditions could potentially result in a malfunction of equipment important to safety, specifically prevention of water hammer in the steam generator preheater and/or steam generator FW inlet piping. The likelihood of occurrence of a malfunction associated with the new operator actions is not more than minimal based on the following:

• The actions to disable/enable the WHPS low SG level and low SG pressure FW isolation auto close signals has been added to the appropriate plant procedures and operators will be trained on the revised procedures.

• The actions to disable/enable the WHPS low SG level and low SO pressure FW isolation auto close signals do not have a required time for completion. Above 25% power there is no specific time requirement for disabling the WHPS low SG level and low SG pressure FW isolation auto close signals. There is no impact on any emergency operating procedures. The proposed operator actions for WHPS enable/disable during power ascension, power descension, and reactor trip are not time critical and do not impose a significant new burden on the operating crew.

• This change does not introduce credible errors during performance of the operator actions required for this change. Three actions are required to block or enable WHPS following this design change. The 2FW009A/B/C/D and



Page 16 of 36

LS-AA-104-1001 Revision 4 Page4 of6

Activity/Document Number: EC 617642 Revision Number: !!


Including EC 617642 Supporting Procedure Revisions: 2BwGP 100-1 Rev. 34; 2BwGP 100-3 Rev. 71; 2BwGP 100-4 Rev. 42; 2BwGP 100-4T4 Rev. 01; 2BwGP 100-ITl Rev. 16; 2BwGP 100-4Tl Rev. 18 and 2BwGP 100-4T5 Rev 1

2FW039A/B/C/D valve water hammer prevention feature will be enabled or blocked with a switchable jumper on each valve specific actuation device. This jumper will be installed at the original water hammer prevention actuation relay contacts, thereby allowing blocking or enabling the original contacts. This switch will be clearly labeled. The 2FW035A/B/C/D valve control switch bypasses the water hammer prevention circuit when it is placed in open (as opposed to auto). This is a routine method of configuration control by operators.

Operation of the blocking switch for the 2FW009's and 2FW039's, and operation of the permanent switch for the 2FW035's will be controlled as part of a formal procedure. The Switchable jumpers to be installed and used, and hand switches to be manipulated per this activity will be clearly flagged and labelled per the temporary Engineering Change to allow for clear and easy operator identification. Maintenance procedure MA-AA-1070, which maintenance has trained on, exists to provide guidance for lifting and landing leads and installing jumpers and will be utilized in support of this activity. These actions do not initiate a water hammer prevention actuation; therefore, tl1ere would be time to recover from unlikely actions such as blocking or enabling the WHPS out of the expected sequence.

"' The system function to preclude the admission of cool FW into a steam-voided preheater section of the steam generator is not adversely impacted.

When the WHPS FW isolation closure signals are disabled, the UFSAR described automatic action to isolate feedwater to the preheater region of the Unit 2 steam generators on low SG level or low SG pressure will be fulfilled by tl1e safety related ESFAS FW isolation signals. Appropriate administrative controls will be put in place to enable the automatic WHPS FW isolation function during plant startup, power descension, and after a plant trip when the ESFAS FW isolation is not available. Thus the proposed change does not permanently substitute manual action for automatic action for performing a UFSAR-described design function.

The WHPS low SG level and low SG pressure FW isolation auto close signals are classified as non-safety related. The WHPS FW isolation auto close signal for each individual generator is based on a two-out-of-three logic. The safety related ESFAS FW isolation signal on low steam line pressure is based on two-out-of-three logic in any one of the four steam lines. A reactor trip is actuated on two-out-of-four low-low water level signals occuning in any steam generator. A FW isolation signal is generated with the reactor trip coincident with a one-out-of-two Lo Tave signals. The safety related ESFAS FW isolation signals provide the same level of redundancy as the WHPS FW isolation signals. The safety related ESFAS components are considered to be more reliable than the non-safety related WHPS components. The highest risk for water hammer events is during startup and low power operation when the ESFAS FW isolation signals may be bypassed and only the WHPS FW isolation signals would be available for water hammer protection. This same level of redundancy will be in place with only the ESFAS FW isolation signals providing protection when the WHPS low SG level and low SG pressure auto close FW isolation signals are disabled.

The switchable jumpers used to defeat water hammer protection on the 2FW009A/B/C/D and 2FW39A/B/C/D valves are built using a rocker switch to facilitate simple operation by the Operator. The installed devices meet or exceed the quality requirements of the circuits they are installed in and are installed in accordance with approved quality controlled procedures to insure the appropriate level of reliability. The jumpers for the 2FW009A/B/C/D valves are installed in series with the applicable contacts with lugs on the terminal block and are considered to be as reliable as the existing circuit. The jumpers on the 2FW039A/B/C/D valves are installed in parallel with tl1e applicable contacts using existing "banana" jacks. Failure of the 2FW039A/B/C/D jumper would conservatively re-institute the WHPS logic for the 2FW039A/B/C/D valves. The control switches for the 2FW035 vales are designed to allow the valves to be held open. Thus the proposed change does not change the likelihood of occurrence of a malfunction of the 2FW035 valves.



Page 17 of 36




Including EC 617642 Supporting Procedure Revisions: 2BwGP 100-1 Rev. 34; 2BwGP 100-3 Rev. 71: 2BwGP 100-4 Rev. 42; 2BwGP 100-4T4 Rev. 01; 2BwGP 100-lTl Rev. 16; 2BwGP 100-4Tl Rev. 18 and 2BwGP 100-4T5 Rev 1

The proposed change does not impact the other UFSAR described WHPS features. When the WHPS FW isolation is disabled at power the water hammer prevention function to preclude injection of cool water to the Steam Generators will be fulfilled by the safety related ESFAS signals to close the feedwater isolation valves on low steam line pressure safety injection or steam generator water level low-low reactor trip coincident with Tave less than 564° F. At low power levels or during startup the WHPS FW isolation features will be enabled. Additionally, the proposed change does not revise the operating procedure steps associated with switching feedwater flow from the top feed nozzle to the lower main feed nozzle. As described in the response to FSAR Question QOI0.51 initial plant startup testing was performed to confirm that no damaging water hammer occurs when FW delivery is transferred from the top SG nozzle to the main feed nozzle. Since the transfer procedure steps are unchanged, there is no change in the frequency of a water hammer when FW delivery is transferred. Therefore the proposed change wiII not result in more than a minimal increase in the likelihood of a water hammer induced malfunction.

Disabling the WHPS low SG level and low SG pressure FW isolation auto close signals above 25% power does not introduce the possibility of a change in the consequences of an accident because WHPS is not an initiator of any accident and no new failure modes are introduced. WHPS FW isolation is not credited with preventing or mitigating any accident previously evaluated in the UFSAR. The ESFAS FW isolation signals (Steam line pressure SI or the Reactor trip+ Lo Tave) will continue to close the feedwater isolation valves as designed for accident mitigation. Thus the proposed activity will not result in a more than minimal increase in the consequences of an accident previously evaluated in the UFSAR.

Disabling the WHPS low SG level and low SG pressure FW isolation auto close signals above 25% power does not introduce the possibility of a change in the consequences of a malfunction of an SSC important to safety previously evaluated in the UFSAR because the proposed change does not create any new malfunctions of the FW isolation valves and no new failure modes are introduced. The proposed change does not impact the safety related ESFAS FW isolation circuitry. Thus there is no change to the UFSAR described failure modes and effects for the ESF actuation system. The proposed changes to the WHPS FW isolation auto close signals does not introduce any new failure mode for the system. WHPS FW isolation is not credited with preventing or mitigating any malfunction of an SSC important to safety previously evaluated in the UFSAR.

Disabling the WHPS low SG level and low SG pressure FW isolation auto close signals above 25% power does not introduce the possibility of a new accident because the proposed change is not the initiator of any accident and no new failure modes are introduced. When the WHPS FW isolation is disabled at power the water hammer prevention function will be fulfilled by the safety related ESFAS signals. Administrative controls will be put in place to ensure that the WHPS FW isolation signals are enabled during startup and low power operation. The likelihood of operator error leading to an accident of a different type is not credible because the actions are reflected in plant procedures and operator training programs, the actions can be completed in the required time, the actions are simple and adequate time is available to recover from any credible error. The design bases function to preclude the admission of cool FW into a steam-voided preheater section of the steam generator is not adversely impacted, thus no new accident types are introduced.

Disabling the WHPS low SG level and low SG pressure FW isolation auto close signals above 25% power does not introduce the possibility of a malfunction of a different result because the activity does not introduce a failure result. When the WHPS FW isolation is disabled at power the water hammer prevention function will be fulfilled by the safety related ESFAS signals. Administrative controls will be put in place to ensure that the WHPs FW isolation signals are enabled during startup and low power operation.

This temporary change to the Unit 2 WHPS low level and low pressure FW isolation auto close signals does not result in a change that would cause any system parameter to change. WHPS is not credited for mitigation in any accident analysis, thus there is no impact on any UFSAR described fission product barrier limit. Therefore, the proposed activity does not result in a DBLFPB as described in the UFSAR being exceeded or altered.

of 36

50.59 REVIEW COVERSHEET FORM LS-AA-104-1001 Revision 4 Page 6 of6




Including EC 617642 Supporting Procedure Revisions: 2BwGP 100-1 Rev. 34; 2BwGP 100-3 Rev. 71; 2BwGP 100-4 Rev. 42; 2BwGP 100-4T4 Rev. 01; 2BwGP 100-lTl Rev. 16; 2BwGP 100-4Tl Rev. 18 and 2BwGP 100-4T5 Rev 1

The proposed change does not involve a method of evaluation as defined in LS-AA-104-1000. 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 analysis.

Based on the results of the 50.59 Evaluation the activity may be implemented per plant procedures without obtaining a License Amendment.



D D C8l


50.59 Screening

50.59 Evaluation

50.59 Screening No.

50.59 Evaluation No. BRW-E-2017-1

Rev.

Rev. 0 -------



Station/Unit(s): Braidwood Units 1 & 2


Activity/Document Number: EC 400434 (Ul) and EC 400436 (U2) I EC 404358 (Ul) and EC 404362 (U2) I EC 404360 (Ul) and EC 404363 (U2) I DRP 17-003 (Ul) and DRP 17-024 (U2) Revision Number: 002 and 001 I 004 and 001I002 and 001 /NA forDRPs

Title: Westinghouse Ovation Digital Upgrade for 7300 NSSS Cabinets 1(2)PA05J, 1(2)PA06J, 1(2)PA07J, 1(2)PA08J I Westinghouse Ovation Digital Upgrade for 7300 BOP Cabinets 1(2)PA20JA and 1(2)PA20JB I Westinghouse Ovation Digital Upgrade for TDFWP Cabinets 1(2) FW36J and 1(2)FW37J

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).


The proposed activity upgrades nuclear steam supply system (NSSS), balance of plant (BOP), and turbine-driven feedwater pump (TDFWP) control systems by modifying individual control systems and incorporating those individual control systems into a plant-wide distributed control system (DCS). The activity includes various control system changes to incorporate improvements and lessons learned.

The activity encompasses the following engineering changes:

• EC 400434 and EC 400436 replace the 7300-series equipment in NSSS cabinets PA05J, PA06J, PA07J, and PA08J. This EC also includes changes to the main control board (MCB) to support the modifications to these NSSS cabinets.

• EC 404358 and EC 404362 replace the 7300-series equipment in BOP cabinets PA20JA and PA20JB. This EC also includes changes to the MCB to support the modifications to these BOP cabinets.

• EC 404360 and EC 404363 eliminate the 7300-series equipment in TDFWP cabinets PA36J and PA37J and install new local cabinets FW36J and FW37J. This EC also includes both changes to the main control board (MCB) to support the modifications to these cabinets and changes to the TDFWP turbine control and protection system hardware.

The proposed activity involves the instrument loops for the following systems, with a wide range in the level of complexity:

1. Distributed Control System (DCS) 2. Reactor Coolant System (RC/RY) 3. Nuclear Instrumentation System (NR) 4. Chemical and Volume Control System (CV) 5. Residual Heat Removal (RH) 6. Safety Injection (SI) 7. Containment Spray (CS) 8. Component Cooling (CC) 9. Auxiliary Feedwater (AF) 10. Main Steam (MS) 11. Main Feedwater (FW) 12. Heater Drain (HD) 13. Condensate I Condensate Booster (CD/CB) 14. Primary Water (PW) 15. Main Generator (HY/WS) 16. Main Turbine (MT) 17. Essential Service Water (SX) 18. Circulating Water (CW) 19. Liquid Radwaste (RF/WX)

The specific changes to these systems are described in the 50.59 screening.



LS-AA-104-1001 Revision 4

Page 2 of3

Activity/Document Number: EC 400434 (Ul) and EC 400436 (U2) I EC 404358 (Ul) and EC 404362 (U2) I EC 404360 (Ul) and EC 404363 (U2) I DRP 17-003 (Ul) and DRP 17-024 (U2) Revision Number: 002 and 001 I 004 and 001I002 and 001 /NA forDRPs

Title: Westinghouse Ovation Digital Upgrade for 7300 NSSS Cabinets 1(2)PA05J, 1(2)PA06J, 1(2)PA07J, 1(2)P A08J I Westinghouse Ovation Digital Upgrade for 7300 BOP Cabinets 1(2)PA20JA and 1(2)PA20JB I Westinghouse Ovation Digital Upgrade for TDFWP Cabinets 1(2) FW36J and 1(2)FW37J


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. The activity also includes various control system changes to incorporate improvements and lessons learned, based on operating experience with the existing control systems.


The activity involves numerous changes to the operator interface with the affected systems. The specific changes are described in the 50.59 screening.

The upgrade and changes to the control systems do not affect the design bases. The changes were reviewed with respect to the safety analyses described in the UFSAR and it was concluded that no changes to the safety analyses are required and the existing safety analyses remain bounding.


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 01-01, these aspects of the proposed activity were "screened in" for further evaluation under the 50.59 process. In addition, certain functional changes to the affected control systems were screened in:

11 elimination of: auctioneered-high signals (Tavg, auctioneered-high nuclear power, and auctioneered-high delta-T) as inputs to major plant control systems, and the feedwater flow input to the heater drain tank level control circuit

11 installation of automatic features in place of manual actions involving: RCS cooldown when Tavg is below 550°F, control of pressurizer pressure below 1700 psig, and several aspects of the main feed water system

The rigorous process used in developing the digital hardware and software and the integrated testing of the major control systems using a plant-specific model were credited with ensuring the modified systems would perform as required. The involvement of the plant operations staff in the development of the human-machine interface, the work with Idaho National Laboratory in reviewing operator interaction with the new equipment at their simulation laboratory, the simulator testing at both Westinghouse and Exelon, and the Braidwood operator training process were credited with ensuring a successful operator interface with the new system. As a result, the proposed activity does 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 existing safety analyses for accidents that could be initiated by failures in the control systems remain bounding, no radiological consequences beyond the current consequences for such events would occur. Therefore, the proposed activity does not result in more than a minimal increase in the consequences of accidents or malfunctions and does not result in a design basis




Activity/Document Number: EC 400434 (Ul) and EC 400436 (U2) I EC 404358 (Ul) and EC 404362 (U2) I EC 404360 (Ul) and EC 404363 (U2) I DRP 17-003 (Ul) and DRP 17-024 (U2) Revision Number: 002 and 001I004 and 001 I 002 and 001 /NA forDRPs

Title: Westinghouse Ovation Digital Upgrade for 7300 NSSS Cabinets 1(2)PA05J, 1(2)PA06J, 1(2)PA07J, 1(2)PA08J I Westinghouse Ovation Digital Upgrade for 7300 BOP Cabinets 1(2)PA20JA and 1(2)PA20JB I Westinghouse Ovation Digital Upgrade for TDFWP Cabinets 1(2) FW36J and 1(2)FW37J

limit for a fission product barrier as described in the UFSAR being exceeded or altered. The proposed activity does not involve a method of evaluation described in the UFSAR.

A review of the failure modes and effects analysis for the activity indicates that that the upgrade of the control systems has eliminated many single point vulnerabilities and those few single failures which could lead to a significant power transient or other serious effect on the plant are present in the existing control systems. A software hazards analysis also evaluated controller malfunctions with respect to existing UFSAR Chapter 15 events to determine whether transient and accident analyses might be affected. The system-level failure analysis did not identify malfunctions (software hazards) that affect the transient and accident analysis in Chapter 15 of the UFSAR. The likelihood of a common cause failure is considered sufficiently low based on the design attributes of the system (e.g., preventive, limiting, and likelihood-reduction measures), the quality of the design processes employed, and operating experience in similar applications. Therefore, the evaluations of single failures and of the potential for common cause failures provide an adequate basis for concluding that the proposed activity does not create a possibility for an accident of a different type or a malfunction with a different result than any previously evaluated in the UFSAR.

Design basis limits for fission product barriers are not affected by the proposed activity. There is no adverse change to an element ofa UFSAR-described evaluation methodology, or use of an alternative methodology, that is used in establishing the design bases or used in the safety analyses.




50.59 Screening

50.59 Evaluation

50.59 Screening No.


BRW-S-2017-2 Rev. 0 -------BRW-E-2017-3 Rev. 0 -------


of 36

50.59 REVIEW COVERSHEET FORM LS-AA-104-1001 Revision 4 Pagel of2

Activity/Document Number: ... 11:..::2"'B'""w"-O~A,_,P,__,RI=--"'-8 _____________ _ Revision Number: 107/108

Title: l/2Bw0A PRI-8 Essential Service Water Malfunction Unit 112



The activity is to revise 1/2BwOA PRI-8, "Essential Service Water Malfunction Unit 1/2'', to allow the use of the non-safety related piping connection (l/2 FX304B-8) between the unit 1/2 'B' Essential Service Water(SX) pumps and the 'B' Auxiliary Feed Water (AFW) Shaft Driven Cooling Water Pumps (l/2 SX04P) in the beyond-design-basis event where there is a complete loss of ultimate heat sink cooling capability. This flow path was installed under EC 394153 to support the requirements of the Flexible and Diverse Coping Strategies (FLEX) Implementation Guide, NEI 12-06 and NRC Order EA-12-049, Order to Modify Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design Basis External Events.


Currently Guideline 1/2 BwFSG-2 "Alternate AFW/EFW Suction Source" for Unit I and Unit 2 is approved to allow the use of the FLEX piping FX304B-8 to connect the discharge of the 'B' SX pumps to the suction of the 'B' AFW pump following a loss of all AC. In order to enable the probabilistic risk assessment (PRA) evaluations for Braidwood Station to credit this connection during a beyond-design-basis event of a loss of all SX, a station procedure is required to implement the actions to connect the SX alternate flow path to the AF Shaft Driven Cooling Water Pump. The Operating Abnormal Procedure l/2Bw0A PRI-8 currently exists to address a loss of Essential Service Water due to a loss of all SX pumps; however, it does not currently credit the use of the FLEX alternate flow path connection piping FX304B-8.

Currently, upon a loss of all SX pumps the 'B' AFW Shaft Driven Cooling Water pump takes suction from the SX system, pumps it through the 'B' AFW pump room cooler, engine cooler, and gear coolers to maintain operation of the 'B' AFW pump. However the discharge of the Shaft Driven Cooling Water pump cycles back into the suction of the cooling pump due to the hydraulic pressures in the SX system. This creates a 'cycling' effect where the cooling water being used by the 'B' AFW pump is being heated by the waste heat from the pump. The current design documents demonstrate the 'B' Af<W pump can operate for a minimum of 2 hours in this condition. For extended operation of the 'B' AFW pump following this beyond-design basis event (a complete loss of all SX pumps) a new cooling source is needed. This activity, revising l/2Bw0A PRI-8, provides guidance on aligning the FLEX alternate path to provide a new cooling source. When the FLEX alternate path is aligned, the Shaft Driven Cooling pump will take suction for the Ultimate Heat Sink and discharge back to the Ultimate Heat Sink, thus ensuring long term access to sufficient cooling water to support the extended operation of the 'B' AFW pump.


The effect of this activity is to allow alignment the non-safety related FLEX alternate flow path between the SX pump discharge and the 'B' AFW Shaft Driven Cooling water pump suction following a complete loss of SX pumps. This allows for extended, greater than 2 hour, run times for the 'B' AFW pump in the condition where there is no forced SX flow on either Unit. The AFW system has a safety related design function to operate during a loss of all AC - a condition which also results in a loss of forced SX flow. The time period for 'B' AFW pump operation during a loss of all AC is described in the UFSAR as being a minimum of2 hours.

Connecting the non-safety related FLEX piping to the safety-related AFW system following a complete loss of SX pumps during this 2 hour period adversely affects the AFW system. The non-safety related FLEX piping has been designed (as described in EC 394153) as Augmented Quality, seismic category I and is designed to remain functional after a Design Basis SSE Earthquake. There the piping is considered not likely to fail during the maximum 2 hour operation it will be required to support the AFW pumps design requirement. The Operator will make the determination that the condition where there is no SX forced flow will exist for longer than an hour, and if it will then the FLEX piping will be used, thus ensuring the safety related piping remains available to support pump operations for as long as practical.

of 36

50.59 REVIEW COVERSHEET FORM LS-AA-104-1001 Revision 4 Page 2 of2

Station/Unit(s): Braidwood Unit l!Unit 2

Activity/Document Number: .:.1/'""2""B....,w'""'O=A~P-=-RI=-"""8'--------------- Revision Number: 107/108

Title: 1/2BwOA PRI-8 Essential Service Water Malfunction Unit 1/2


The design requirement of the AFW system, for the loss of all AC event, is to provide a safety related injection to the steam generators for a period of at least 2 hours. This activity permits the use of a non-safety related piping connection to the AFW system following a beyond-design-basis event (loss of all SX pumps due to an SX system malfunction) to allow extending the operation time of the AFW system beyond the 2 hour requirement, however this piping system may be connected prior to the 2 hour operation time being exceeded.

The piping has been designed to withstand a design basis seismic event and continue to function. The water source for the Shaft Driven Cooling pump will be changed from the outlet of the SX strainer, to the inlet of the strainer, and the effects of this nonstrained water has been evaluated to not degrade the AFW pump performance. The use of procedure 1/2Bw0A PRI-8 to direct the evolution to connect the FLEX alternate piping has been reviewed and determined to be within the allowed use for an Operating Abnormal Procedure as described in UFSAR section 13.5 'Plant Procedures'. However, the AF system currently is designed to start, initiate and provide injection flow to the Steam Generators without operator intervention. The change to 1/2BwOA PRI-8 requires operator intervention, albeit an optional intervention, that has been determined to be adverse in the screening.

In addition allowing the connection of the non-safety related FLEX piping to the AFW system during the time period where the AFW system is described as performing a safety related function was determined to be adverse in the screening, and was reviewed in the 50.59 Evaluation.

The attached Evaluation has demonstrated that the effect of using the non-safety FLEX piping does not present a more then minimal increase in the likelihood of a malfunction of an SSC important to safety previously evaluated in the UFSAR. In addition the operator intervention provided by the procedure has been demonstrated to not be more than a minimal increase in the likelihood of a malfunction of an SSC due to the option to use operator action to extend the operational time for the AFW system during a loss of all forced SX flow.

Therefore, based on this Screening and Evaluation the activity may be implemented under the governing procedures.




50.59 Screening

50.59 Evaluation

50.59 Screening No.


BRW-S-2017-24 Rev. 0 -------BRW-S-2017-25 Rev. _o ______ _

See LS-AA-I 04, Section 5, Documentation, for record retention requirements for this and all other 50.59 fo1ms associated with the Activity.



Page 24 of36


Activity/Document Number: EC 618295 /UFSAR DRP 17-010 Revision Number: 0 I 0

Title: Disable FW Water Hammer Prevention System (WHPS) When 2FW009A/B/C/D Are Open (Above Approx. 25% Power) For Steam Generators 2A/2B/2C/2D

Including EC 618295 Supporting Procedure Revisions: 2BwGP 100-1Rev.35; 2BwGP 100-3 Rev, 72; 2BwGP 100-4 Rev. 43: 2BwGP 100-4T4 Rev. 02; 2BwGP 100-lTl Rev. 17; 2BwGP 100-4Tl Rev. 18 and 2BwGP 100-4T5 Rev l; BwAR 2-15-ElZ, Rev. 52


Description of Activity: (Provide a brief, concise description of what the proposed activity involves.) Modification Engineering Change (EC) 618295 provides the means to disable the Water Hammer Prevention System (WHPS) low Steam Generator (SG) level and low SG pressure feed water isolation signals on Unit 2 after Feed water (FW) flow is transferred from the FW isolation valve bypass valves, 2FW043A/B/C/D, to the main feedwater isolation valves, 2FW009A/B/C/D, (when power is above approximately 25% power).

The proposed change to disable the WHPS FW isolation involves installing a new switch in the circuit between two terminal block points for each of the 2FW009A/B/C/D and 2FW039A/B/C/D valves, and repositioning the control switch from "Auto" to "Open" for each of the 2FW035NB/C/D valves. The associated annunciator circuit and annunciator tile for 2FW035A/B/C/D position is modified to reflect the new nonnal position for power operation (above 25%). The WHPS FW isolation signals to the 2FW043A/B/C/D valves are not disabled.

The following procedures implement manual enabling/disabling of the WHPS:

• 2BwGP 100-1, PLANT HEATUP, will be revised to verify that WHPS FW isolation has been enabled prior to starting any FW pumps.

e 2BwGP 100-3, POWER ASCENSION, will be revised to disable WHPS FW isolation after the 2FW009A/B/C/D valves are open.

e 2BwGP 100-4, POWER DESCENSION, will be revised to enable WHPS FW isolation for the 2FW035A/B/C/D and 2fW039A/B/C/D valves, close the 2FW009A/B/C/D valves, and then perform a follow-up action to enable WHPS FW isolation on the 2FW009A/B/C/D valves.

.. 2BwGP 100-4T4, REACTOR TRIP POST RESPONSE GUIDELINE, will be revised to verify that the 2FW009A/B/C/D, 2FW043A/B/C/D, and 2FW039A/B/C/D valves are closed prior to starting the Startup Feedwater Pump. Action will also be taken to enable WHPS FW isolation later in the post trip response.

.. Associated Flowchart procedures 2BwGP 100-lTl, 2BwGP 100-4Tl and 2BwGP 100-4T5 will be revised as necessary to reflect updated step numbers.

UFSAR Section 10.4.7.3 is being revised to describe the disabling of WHPS when the ESFAS Feedwater Isolation feature is active.

Reason for Activity: (Discuss why the proposed activity is being performed.) The steam generator water hammer prevention relays are interlocked with the FWIV "Close" circuits. The relays are normally energized for the low SG level and low SG pressure permissive circuit on each FWIV. On a loss of power to the relays, the result is closure of the associated FWIV and a reactor trip will follow. Also note that the WHPS relays are powered by NonSafety Related (NSR) power. The proposed change will allow the SG water hammer prevention actuation circuit to be disabled once feedwater isolation valves have been opened.

The changes to the plant operating procedures are made to administratively control when the WHPS FW isolation function is disabled and enabled.

of 36



LS-AA-!04-1001 Revision 4 Page 2 of 6


Title: Disable FW Water Hammer Prevention System (WHPS) When 2FW009A/B/C/D Are Open (Above Amnox. 25% Power) For Steam Generators 2A/2B/2C/2D

IncludingEC618295 Supporting Procedure Revisions: 2BwGP 100-1Rev.35; 2BwOP 100-3 Rev. 72; 2BwGP 100-4 Rev. 43: 2BwGP I00-4T4 Rev. 02: 2BwGP !00-!Tl Rev. 17; 2BwGP I00-4TI Rev. 18 and 2BwGP I00-4T5 Rev l: BwAR 2-l5-El2, Rev. 52

Effect of Activity: (Discuss how the activity impacts plant operations, design bases, or safety analyses desc1ibed in the UFSAR.)

The safety function of the Feed water Isolation Valves is to close, isolating main feedwater flow upon a safety injection signal, feedwater isolation signal and high-2 SO level. As described in UFSAR section 10.4.7.3, "Several water hammer prevention features have been designed into the feedwater system. These features are provided to minimize the possibility of various water hammer phenomena in the steam generator preheater, steam generator main feedwater inlet piping, and the steam generator upper nozzle feedwater piping." Steam generator (two-out-of-three logic) low level trips are provided to close all feedwater isolation valves (2FW009A/B/C/D), feedwater isolation bypass valves (2FW043A/B/C/D) and feedwater preheater bypass valves (2FW039A/B/C/D). Steam generator (two-out-of-three logic) low pressure trips are provided to close all feedwater isolation valves, feedwater isolation bypass valves, feedwater preheater bypass valves, and the feedwater bypass tempering valves (2FW035A/B/C/D).

The UFSAR described water hammer prevention features were included in the NRC's basis for approval of (1) the designs of the Stearn Generator and the FW/AFW systems, and the measures to address Unresolved Safety Issue (USI) A-1, Steam Generator Water Hammer; and (2) the removal of pipe restraints resulting from the elimination of arbitrary intermediate breaks in the FW system piping system design.

The purpose of the WHPS FW isolation is to prevent water hammer by precluding the admission of cold FW into a steamvoided preheater section of the steam generator (when SG narrow range level is less than 5% or when SG pressure is less than 600 psig}. When the WHPS FW isolation is disabled at power the water hammer prevention function will be fulfilled by the safety related ESFAS signals function to close the feedwater isolation valves on low steam line pressure safety injection or steam generator water level low-low reactor trip coincident with Tave less than 564° F. Note that the 2FW039 valves close on reactor trip and no coincident Tave is required. At low power levels or during startup the WHPS FW isolation features will be enabled. The design bases function to preclude the admission of cold FW into a steam-voided preheater section of the steam generator is maintained.

Additionally, the other UFSAR described WHPS features will not be altered by this EC. Operating procedures associated with only providing feedwater to the upper nozzle of the steam generator during startup and low load conditions are unchanged. Temperature monitoring and alarms to detect possible back leakage of steam from the steam generators into the feedwater piping are unchanged. Forward flushing and temperature permissive interlocks associated with opening the FWIV are also unchanged. This EC does not alter split feedwater flow operation.

Implementation of EC 618295 does not impact the safety related circuits desc1ibed in the UFSAR and Technical Specifications that generate an independent Feedwater Isolation Signal, including Safety Injection (SI), SG Level HI-2 Trip (P-14), and Reactor Trip (P-4) coincident with RCS Average Temperature Low at 564°F.

The change will require new operator actions to appropriately enable/disable WHPS FW isolation in the plant startup, power ascension, power descension, and plant trip response procedures. Operators will reposition the control board hand switch for the 2FW035 A/B/C/D valves and reposition the switches for the 2FW009A/B/C/D and 2FW039A/B/C/D valve circuits.

• For startup WHPS FW isolation will be verified to be enabled prior to starting any FW pumps. Once the 2FW009 valves are open, action will be taken to disable WHPS FW isolation for the 2FW009, 2FW035, and 2FW039 valves.



Page 26 of 36

LS-AA- I 04- J 00 I Revision 4 Page 3 of6


Title: Disable FW Water Hammer Prevention System (WHPS) When 2FW009A/B/C/D Are Open (Above Approx. 25% Power} For Steam Generators 2A/2B/2C/2D

Inclmling EC 618295 Supporting Procedure Revisions: 2BwGP 100-1 Rev. 35; 2BwGP 100-3 Rev. 72; 2BwGP 100-4 Rev. 43; 2BwGP I00-4T4 Rev. 02; 2BwGP 100-ITl Rev. 17: 2BwGP 100-4T1Rev.18 and 2BwGP I00-4T5 Rev I: BwAR 2-15-El2, Rev. 52

.. For a normal shutdown or low power operation action will be taken Lo re-enable WHPS FW isolation for the 2FW035 and 2FW039 valves prior to closing the FW009 valves. WHPS FW isolation for the 2FW009 valves will be enabled prior to defeating the ESFAS FW isolation signals.

"' On a plant trip action will be taken to verify that the 2FW009, 2FW039, and 2FW043 valves are closed prior to starting the Startup FW pump. The 2FW035 valve flow path is used to direct FW to the upper nozzle. TI1e UFSAR described main FW flow and temperature permissive interlocks will prevent inadvertent introduction of cold FW to the lower nozzle until action is taken to enable WHPS FW isolation.

There is no impact on any emergency operating procedures. The proposed operator actions are not time critical and do not impose a significant new burden on the operating crew.


Disabling the WHPS low SG level and low SG pressure FW isolation auto close signals above 25% power does not introduce the possibility of a change in the frequency of an accident because WHPS is not an initiator of any accident and no new failure modes are introduced. Disabling tbe WHPS low SO level and low SG pressure FW isolation auto close signals above 25 % power reduces the probability of a loss of feed water event due to a spurious WHPS actuation.

In the plant operating conditions under which the \VHPS automatic FW isolation features may be procedurally disabled, the safety related ESFAS signals function to close tbe feedwater isolation valves on a low steam line pressure safety injection(< 640 psig) or steam generator water level low-low reactor trip($ 36.3% NR) coincident with low Tave (less tban 564" F). Note that the 2FW039NB/C/D valves close on reactor trip and no coincident Tave is required. A review of the postulated transients tbat could create conditions for a water hammer event when power is above 25% was performed (Refer to EC 618295). The review determined that the ESFAS FW isolation signals (Steam line pressure SI or the Reactor trip+ Lo Tave) will function to prevent water hammer by precluding the admission of cold FW into a steam-voided preheater section of tbe steam generator during power operating conditions. Therefore disabling the WHPS low SG level and low SG pressure FW isolation auto close signals above 25% power does not change the possibility of water hammer in the SG preheater and the main FW inlet piping. Thus the proposed change does not result in more than a minimal increase in the frequency of occurrence of a FW line break or steam generator tube rupture.

The defeated WHPS FW isolation function is not used or credited to mitigate any design basis event, analysis, or Emergency Operating Procedure Operator Action. The credited automatic function is the ESFAS FW Isolation which is a completely separate circuit. Thus ESFAS FW Isolation is not affected by this change.

The change adds new operator actions to disable and enable the WHPS low SG level and low SG pressure FW isolation auto close signals. Failure to enable the WHPS low SG level and low SG pressure auto close signals for startup, low load, or increasing load conditions could potentially result in a malfunction of equipment important to safety, specifically prevention of water hammer in the steam generator preheater and/or steam generator FW inlet piping. The likelihood of occurrence of a malfunction associated with tbe new operator actions is not more than minimal based on the following:



Page 27 of36

LS-AA-104-1001 Revision 4 Page 4 of 6

Activity/Docmnent Number: EC 618295 I UFSAR DRP 17-010 Revision Number: 0 I 0

Title: Disable FW Water Hammer Prevention System (WHPS) When 2FW009A/B/C/D Are Open (Above Apnrox. 25 % Power) For Steam Generators 2A/2B/2C/2D

Inclmling EC 618295 Supporting Procedure.Revisions: 2BwGP 100-1 Rev. 35; 2BwGP 100-3 Rev. 72: 2BwGP 100-4 Rev. 43: 2BwGP I00-4T4 Rev. 02: 2BwGP 100-ITl Rev. 17: 2BwGP I00-4Tl Rev. 18 and 2BwGP I00-4T5 Rev I; BwAR 2-15-El2, Rev.52

GI The actions to disable/enable the WHPS low SG level and low SG pressure FW isolation auto close signals has been added to the appropriate plant procedures and operators will be trained on the revised procedures.

GI The actions to disable/enable the WHPS low SO level and low SG pressure FW isolation auto close signals do not have a required time for completion. Above 25% power there is no specific time requirement for disabling the WHPS low SG level and low SG pressure FW isolation auto close signals. There is no impact on any emergency operating procedures. The proposed operator actions for WHPS enable/disable during power ascension, power descension, and reactor trip are not time critical and do not impose a significant new burden on the operating crew.

111 This change does not introduce credible errors during performance of the operator actions required for this change. Three actions are required to block or enable WHPS following this design change. The 2FW009A/B/C/D valve water hammer prevention feature will be enabled or blocked with a switch in series with each valve specific actuation device circuit. The 2FW039A/BJC/D valve water hammer prevention feature wiH be enabled or blocked with a switch on each valve specific actuation device circuit. This switch will be installed at the original water hammer prevention actuation relay contacts, thereby allowing blocking or enabling the original contacts. The new switches will be clearly labeled. The 2FW035A/B/C/D valve control switch bypasses the water hammer prevention circuit when it is placed in open (as opposed to auto). This is a routine method of configuration control by operators.

Operation of the switches for the 2FW009's and 2FW039's, and operation of the control switch for the 2FW035's will be controlled as part of an approved procedure. The switches to be installed and used, and hand switches to be manipulated per this activity will be clearly labelled per the Engineering Change to allow for clear and easy operator identification. Operation of the switches does not initiate a water hammer prevention actuation; therefore, there would be time to recover from unlikely actions such as blocking or enabling the WHPS out of the expected sequence.

GI The function to preclude the admission of cold FW into a steam-voided preheater section of the steam generator is maintained.

When the WHPS FW isolation closure signals are disabled, the UFSAR described automatic action to isolate feedwater to the preheater region of the Unit 2 steam generators on low SG level or low SG pressure will be fulfiIJed by the safety related ESFAS FW isolation signals. Appropriate administrative controls will be put in place to enable the automatic WHPS FW isolation function during plant startup, power descension, and after a plant trip when the ESFAS FW isolation is not available. Thus the proposed change does not substitute manual action for automatic action for performing a UFSARdescribed design function.

The WHPS low SG level and low SO pressure FW isolation auto close signals are classified as non-safety related. The WHPS FW isolation auto close signal for each individual generator is based on a two-out-of-three logic. The safety related ES FAS FW isolation signal on low steam line pressure is based on two-out-of-three logic in any one of the four steam lines. A reactor trip is actuated on two-out-of-four low-low water level signals occuning in any steam generator. A FW isolation signal is generated with the reactor llip coincident with a two-out-of-four Lo Tave signals. The safety related ESFAS FW isolation signals provide the same level of redundancy as the WHPS FW isolation signals. The safety related ESFAS components are considered to be more reliable than the non-safety related WHPS components. The highest risk for water hammer events is during startup and low power operation when the ESFAS FW isolation signals may be bypassed and only the WHPS FW isolation signals would be available for water hammer protection. This same level of redundancy will be in


Station/Unit(s): :Braidwood Unit 2

Page 28 of 36

LS-AA-104-JOOl Revision 4 Page 5 of 6

Activity/Document Number: EC 618295 / UFSAR DRP 17-010 Revision Number: 0 I 0

Title: Disable FW Water Hammer Prevention System IWHPS) When 2FW009A/B/C/D Are Open (Above Approx. 25% Power) For Steam Generators 2A/2B/2C/2D

Including EC 618295 Supporting Procedure Revisions: 2BwGP 100-1 Rev. 35; 2BwGP 100-3 Rev. 72; 2BwGP I00-4 Rev. 43: 2BwGP l00-4T4 Rev. 02; 2BwGP 100-!Tl Rev. 17; 2BwGP l00-4Tl Rev. 18 and 2BwGP I00-4T5 Rev I; BwAR 2-15-El2, Rev. 52

place with only the ESFAS FW isolation signals providing protection when the WHPS low SG level and low SG pressure auto close f<W isolation signals are disabled.

The switches used to defeat water hammer protection on the 2FW009A/B/C/D and 2FW39A/B/C/D valves meet or exceed the quality requirements of the circuits they are fostalled in and are installed in accordance wit11 approved quality controlled procedures to ensure the appropriate level of reliability. The switches for the 2FW009A/B/C/D valves are installed in series with the applicable contacts with lugs on the terminal block and are considered to be as reliable as the existing circuit The switches on the 2FW039A/B/C/D valves are installed in parallel with the applicable contacts with lugs on the terminal block and are considered to be as reliable as the existing circuit.

The control switches for the 2FW035 vales are designed to allow the valves to be held open. Thus the proposed change does not change the likelihood of occurrence of a malfunction of the 2FW035 valves.

The proposed change does not impact the other UFSAR described WHPS features. When the WHPS FW isolation is disabled at power the water hammer prevention function to preclude injection of cold water to ilie Steam Generators will be fulfilled by the safety related ESFAS signals to close the feedwater isolation valves on low steam line pressure safety injection or steam generator water level low-low reactor trip coincident wiili Tave less than 564° F. At low power levels or during startup the WHPS FW isolation features will be enabled. Additionally, the proposed change does not revise the operating procedure steps associated with switching feedwater flow from the top feed nozzle to the lower main feed nozzle. As described in the response to FSAR Question QOI0.51 initial plant startup testing was perfonned to confirm that no damaging water hammer occurs when FW delivery is transferred from the top SG nozzle to the main feed nozzle. Since ilie transfer procedure steps are unchanged, iliere is no change in the frequency of a water hammer when FW delivery is transferred. Therefore ilie proposed change will not result in more than a minimal increase in the likelihood of a water hammer induced malfunction.

Disabling the WHPS low SO level and low SG pressure FW isolation auto close signals above 25% power does not introduce the possibility of a change in the consequences of an accident because WHPS is not an initiator of any accident and no new failure modes are introduced. WHPS FW isolation is not credited wiili preventing or mitigating any accident previously evaluated in ilie UFSAR. The ESFAS FW isolation signals (Steam line pressure SI or ilie Reactor trip+ Lo Tave) will continue to close the feedwater isolation valves as designed for accident mitigation. Thus ilie proposed activity will not result in a more ilian minimal increase in the consequences of an accident previously evaluated in the UFSAR.

Disabling the WHPS low SG level and low SG pressure FW isolation auto close signals above 25% power does not introduce the possibility of a change in the consequences of a malfunction of an SSC important to safety previously evaluated in ilie UFSAR because the proposed change does not create any new malfunctions of ilie FW isolation valves and no new failure modes are introduced. The proposed change does not impact the safety related ESFAS FW isolation circuitry. Thus there is no change to the UFSAR desclibed failure modes and effects for the ESF actuation system. The proposed changes to the WHPS FW isolation auto close signals does not introduce any new failure mode for the system. WHPS FW isolation is not credited with preventing or mitigating any malfunction of an SSC important to safety previously evaluated in the UFSAR.

Disabling the WHPS low SG level and low SG pressure FW isolation auto close signals above 25% power does not introduce the possibility of a new accident because the proposed change is not ilie initiator of any accident and no new failure modes are introduced. When the WHPS FW isolation is disabled at power the water hammer prevention function will be fulfilled by ilie safety related ESFAS signals. Administrative controls will be put in place to ensure that the WHPS FW isolation signals are enabled during startup and low power operation. The likelihood of operator error leading to an accident of a different type is not credible because the actions are reflected in plant procedures and operator training programs, the actions can be completed in the required time, the actions are simple and adequate time is available to recover from any credible eITor. The


Station/Unit(s): :Braidwood Unit 2

Page 29 of 36


Activity/Document Number: EC 618295 I UFSAR DRP 17-010 Revision Number: 0 I 0

Title: Disable FW Water Hammer Prevention System (WHPS) When 2FW009A/B/C/D Are Open (Above Approx. 25% Power) For Steam Generators 2A/2B/2C/2D

Including EC 618295 Supporting Procedure Revisions: 2BwGP 100-1Rev.35; 2BwGP 100-3 Rev. 72; 2BwGP 100-4 Rev. 43: 2BwGP l00-4T4 Rev. 02; 2BwGP 100-ITI Rev. 17; 2BwGP I00-4Tl Rev. 18 and 2BwGP I00-4T5 Rev 1: BwAR 2-!5-E12, Rev. 52

design bases function lo preclude the admission of cold FW into a steam-voided preheater section of the steam generator is maintained, thus no new accident types are introduced.

Disabling the WHPS low SG level and low SG pressure FW isolation auto close signals above 25% power does not introduce the possibility of a malfunction of a different result because the activity does not introduce a failure result. When the WHPS FW isolation is disabled at power the water hammer prevention function will be fulfilled by the safety related ESFAS signals . . Administrative controls will be put in place to ensure that the WHPS FW isolation signals are enabled during startup and low power operation.

This change to the Unit 2 WHPS low level and low pressure FW isolation auto dose signals does not result in a change that would cause any system parameter to change. WHPS is not credited for mitigation in any accident analysis, thus there is no impact on any UFSAR described fission product barrier limit. Therefore, the proposed activity does not result in a DBLFPB as described in the UFSAR being exceeded or altered.

The proposed change does not involve a method of evaluation as defined in LS-AA-I 04-1000. 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 analysis.

Based on the results of the 50.59 Evaluation the activity may be implemented per plant procedures without obtaining a License Amendment.



D D [gJ

Applicabitity Review

50.59 Screening

50.59 Evaluation

50.59 Screening No.


Rev.

BRW-E-2017-26 Rev. 0 -------


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50.59 REVIEW COVERSHEET FORM LS-AA-104-1001 Revision4 Page 1 of4

Station/Unit(s): Braidwood Station/Units 1 and 2

Activity/Document Number: DRP 17-009ffSB Change 17-001 Revision Number: 0/0

Title: Appendix J Local Leak Rate Scope Reduction



The proposed activity entails the exclusion from IO CFR 50 Appendix J, Option B, Type C local leak rate testing (LLRT) of a number of Containment Isolation Valves. The isolation valves and the corresponding containment penetrations are given below:

Penetration Valve HPN Descriution Chemical and Volume Control System fCVl

1(2)CV8100 MOV U-1(U-2) Reactor Coolant Pumps Seal Leak/Off Header Outside Containment Isolation Valve

1(2)CV8112 MOVU-1(U-2) Reactor CoolantPmnps Seal Leak/Off Header Inside Containmentlsolation Valve

1(2)CV8113 U-1(U-2) RCP Seal Leak/Off Header Isolation Valve 1£2lCV8112 Bypass Header Check Valve

UFSAR Section 6.2.4.2.7 will be revised to document the exclusion from 10 CFR 50 Appendix J, Option B, Type C local leak rate testing for the penetrations and valves above. UFSAR Table 6.2-58 will be revised to indicate that Type C testing is not applicable for these penetration and valves.

Technical Specifications Bases Table B3.6.3-1 will be revised to annotate that the penetrations and valves above are excluded from 10 CFR 50 Appendix J, Option B, Type C local leak rate testing.

The procedures below are impacted by the exclusion from 10 CFR 50 Appendix J, Option B, Type C local leak rate testing. The current revision levels for the affected procedures are listed below:

Unitl • lBwOSR 3.6.1.1-9 Rev. 14, Primary Containment Type C Local Leakage Rate Tests of Chemical and

Volume Control System • IBwVSR 3.6.1.1-25 Rev. 8, Summation of Type Band C Tests for Acceptance Criteria

Unit2 • 2BwOSR 3.6.1.1-9 Rev. 11, Primary Containment Type C Local Leakage Rate Tests of Chemical and

Volume Control System • 2BwVSR 3.6.1.1-25 Rev. 9, Summation of Type Band C Tests for Acceptance Criteria


The elimination of local leak rate testing for valves which meet the exclusion criteria will reduce personnel radiation exposure and will decrease work scope for future refueling outages.


StationfUnit{s): Braidwood Station/Units 1 and 2

Page 31 of 36

LS-AA-104-1001 Revision4 Page2of4

Activity/Document Number: DRP 17-009/I'SB Change 17-001 Revision Number: 0/0

Title: Appendix .I Local Leak Rate Scope Reduction


This activity revises the UFSAR, Technical Specification Bases, and applicable procedures to reflect the exclusion from 10CFR50, Appendix J, Option B, Type C local leak rate testing (LLRT) for the containment isolation valves listed in the Description of Activity section above. EC 406445 and supporting documents justify this exclusion based on the current licensing basis for Braidwood Station. The basis for this exclusion is given below.

Technical Specification 5.5.16, Containment Leak Rate Testing Program states that the program shall be in accordance with the guidelines contained in Regulatory Guide (RG) 1.163, "Performance-Based Containment LeakTesting Program," dated September 1995 with exceptions as noted in the TS. RG 1.163 was issued by the NRC to endorse the use of NEI 94-01.

In NEI 94-01 Revision 0, Section 6.0, General Requirements, components that may be excluded from the LLRT requirements of 10CFR50 Appendix J are described below:

An LLR.T is a test performed on Type Band Type C components. An LLR.T is not required for the following cases: • Primary containment bozmdaries that do not constitute potential primary containment atmospheric pathways

during and following a Design Basis Accident (DBA); • Boundaries sealed with a qualified seal system; or, • Test connection vents and drains between primary containment isolation valves which are one inch or less in

size, administratively secured closed and consist of a double barrier.

EC 406445 concludes that the CV valves in the Description of Activity section above satisfy the first criterion above for exclusion from Type C tests as the affected penetrations do not represent a potential atmospheric release pathway. The valves will continue to be tested in accordance with Inservice Testing (IST) Program required by 10 CFR 50.55a. The affected valves will also remain classified as containment isolation valves. No physical work is required to the valves or the system as a result of the proposed activity.

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 µte 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 LLRT scope reduction and the document changes under EC 406445 are considered a change to a procedure (LLRT program implementation procedures) that may adversely affect how a UFSAR described design function is performed or controlled. This activity reduces the scope of Type C LLRT testing, the effect of the proposed changes is considered to be adverse. Therefore, a 50.59 Evaluation has been completed.

The completed 50.59 Evaluation concluded that that the changes can be implemented per plant procedures without obtaining a License Amendment. The basis for this conclusion is presented below.

The proposed activity does not physically alter any equipment, system performance, or operator actions that could affect the accidents and transients in Chapters 6 and 15 of the UFSAR or their frequency of occurrence. Therefore, the changes related to implementing the LLRT scope reduction under EC 406445 do not introduce the possibility of a change in the frequency of occurrence of an accident previously evaluated in the UFSAR because the leakage of these valves is not an initiator of any accident and no new failure modes are introduced.


Page 32 of 36

LS-AA-104-1001 Revision4 Page3 of4


ActivityIDocument Number: DRP 17-009ffSB Change 17-001 Revision Number: 0/0


The proposed changes do not physically alter any equipment, system performance, or operator actions that could affect the performance of the design function of the containment isolation valves associated with these changes. The current UFSAR analyses remain bounding. Therefore, the documentation changes implemented under this activity do not result in more than a minimal increase in the likelihood of occurrence of a malfunction of an SSC important to safety previously evaluated in the UFSAR. The valves will continue to be tested in accordance with the Inservice Testing Program and station surveillance testing procedures. Inservice Testing can include stroke timing, exercising, and/or leakage checks. This testing remains unchanged as it is required by 10 CFR 50.55a, and is intended to ensure that a malfunction of the affected valves does not occur.

The proposed activity does not result in any changes to the operation of the affected system (CV) that could impact the diversity and redundancy of the system in a manner that could reduce the ability to perform its intended design functions. The affected penetration does not represent a potential primary containment pathway for atmospheric release. Leakage through the isolation valves for penetration P-28 will be contained within the CV pumps' suction header which is a closed system outside containment (Reference NUREG-0876, Byron Safety Evaluation Report, also applicable to Braidwood) and meets the applicable regulatory guidance (ASME Class 2, Safety Category I, not communicate with the outside atmosphere, withstand containment temperature and pressure design conditions, protected against HELB and missiles, be capable of being leak tested). This activity does not result in more than a minimal increase in the consequences of an accident previously evaluated in the UFSAR.

The affected CV penetration and valves are excluded from Type C LLRT on the basis that they do not constitute a potential primary containment atmospheric pathway during a design basis accident, therefore, this activity does not result in more than a minimal increase in the consequences of a malfunction of an SSC important to safety previously evaluated in the UFSAR.

The documentation changes implemented under this activity do not physically alter any equipment, system performance, or operator actions that could affect the performance of the design function of the containment isolation valves associated with these changes. The current UFSAR analyses remain bounding. Therefore, the proposed activity does not create a possibility for an accident of a different type than any previously evaluated in the UFSAR.

Removing Type C (LLRT) testing requirements for the affected CV penetration and containment isolation valves in accordance with approved regulatory documentation does not introduce the possibility for a malfunction of an SSC with a different result because the activity does not introduce any new failure modes or failure results. 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 the UFSAR.

This activity does not affect the design function of plant systems. Therefore, the maximum calculated containment pressure, Pa. as given in Technical Specification Section 5.5-16 is not affected. Since Pa is used in plant procedures to calculate the value of the allowed maximum containment leakage, La, in Standard Cubic Feet per Hour (SCFH), the calculated value of La is not impacted. The changes related to implementing the LLRT scope reduction under EC 406445 do not alter the containment leakage criteria defined in Technical Specifications 5.5.16. 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.

The changes to the UFSAR, Technical Specification Bases, and procedures related to implementing the LLRT scope reduction under EC 406445 do not involve a method of evaluation as described in the UFSAR. La is not being changed as a result of this activity.

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50.59 REVIEW COVERSHEET FORM LS-AA-104-1001 Revision4 Page4 of4


Activity/Document Number: DRP 17-009/l'SB Change 17-001 Revision Number: 0/0


Therefore, the activity does not result in a departure from a method of evaluation. This activity does not impact, revise or replace the methods used to determine primary containment leakage.



D D [81


50.59 Screening

50.59 Evaluation

50.59 Screening No.

50.59 Evaluation No. BRW-E-2017-27

Rev.

Rev.


50.59 REVIEW COVERSHEET JrORM

Stntion/Unit(s): Braidwood Unit I

Activity/Document Number: EC 404449

LS-AA-104-1001 Rcvision 4 Page I of3

Revision Number: 002

Title: Install/Remove Temporary Con!rol Panel in Support or Ovation Digital Upgrade for 7300 NSSS and BOP Cabinets TCCP Temporary Change

NOTE: For 50.59 Evaluations. information on this form will provide the hasis for preparing the biennial summary report suhmiucd to the NRC in accordance with rhe requiremenrs of lO CFR 50.59(d)(2).

Dcscri1>tion 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 performed in support of separate activities which arc upgrading multiple nuclear steam supply syslem (NSSS) and balance of plant (BOP) control systems and trnnsforring them from the existing 7300-serics system to an Ovation-based distributed control system (DCS) under EC 400434 and EC 404358.

The TCC involves the transfer of instrument loops which are to be upgraded but which are also required and/or desired to be available during the trnnsil1on 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 input/output (110) modules o temporary cables from the existing NSSS (PA05J, PA06J, PA07J, PA08J) and BOP (PA20JA, PA20JB) cabinets to the

TCP for the affected instrument loops • two temporary workstations in the control room and one temporary workstation in the radwaste control room • connections of the temporary workstations and TCP lo 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 lhe EC 404449 implementing refueling outage. 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 wi JI then he a part of the permanent! y upgraded NSSS and BOP control systems.

The proposed activity involves the following systems:

l. 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. Primary Water (PW) 9. Essential Service Water (SX) 10. Circulating Water (CW) I I. Liquid Radwaste (RW)

The proposed activity does not involve the major NSSS or BOP control systems (i.e., control of Tavg, pressurizer pressure I 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 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.


Station/Unit(s): Braidwood Unit I

LS-AA-104-JOOl Revision 4 Page2of3

Activity/Documcnl Number: EC 404449 Revision Number: QQ1

Title: Install/Remove Temporary Control Panel in Suppo11 or Ovation Digital Upgrade for 7300 NSSS and BOP Cabinets TCCP Temporary Change


The activity is part of an overall phased project lo upgrade key process control systems integrating them imo a DCS and eliminating the existing Westinghouse 7300-serics 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 i mplernenting refueling outages. The TCC allows work required for the permanent upgrade of the NSSS and BOP control systems to !he Ovation-based DCS to be initimed prior to commencement of the implementing refueling outages.


The TCC will install a temporary con!rol panel in the AEER and two temporary workstations in the control room. Most of the affected instrnment loops are used to moni!or the status of equipment and provide indication and alarm functions only. These instrument loops will be transferred lo 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 instrnment loops to the TCP will result in the loss of certain main control board indication, control, or almm (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 aJarm 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 lo 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. Jn accordance with the guidance provided in NEI 01-0 I, these aspects of tl1e 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 1igorous 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 humanmachine interface and the Braidwood operator training process were credited with ensuring a successful operator interface with 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 maJfunction 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. No radiological

59 REVIEW COVERSHEET FORM

Station/Unit(s): Braidwood Uni! I

Activity/Document Number: EC 404449

LS,AA-104-1001 Revision 4 Page 3 or 3

Revision Number: QQ£

Title: Install/Remove Temporary Control Panel in Support of Ovation Digital Ungrade for 7300 NSSS and BOP Cabinets TCCP Temporary Change

consequences would occur, and no design basis limil for a fission product burlier as described in the UFSAR is exceeded or altered. Since administrative controls lo isolate the RCS from potential sources of unboratecl water will be implemented in Modes 3, 4, and 5 lo ensure that an inadvertent dilution cannot occur, the possibility for a malfunction with a different resul! is not created. The proposed activity docs not involve a method of evaluation described in the UFSAR.




50.59 Screening

50.59 Evaluation

50.59 Screening No.


BRW-S-20 l 8-6

BRW-E-2018-7

Rev.

Rev. 0 -------

See LS-AA-104, Section 5, Documentation, for record retention requirements for this and all olher 50.59 forms associated with the Activity.

ti braidwood station exelon generation company, llg

Documents