WBL-23-045, 10 CFR 50.59 Summary Report

10 CFR 50.59 Summary Report

ML23311A044
Person / Time
Site:	Watts Bar
Issue date:	11/07/2023
From:	Anthony Williams Tennessee Valley Authority
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
WBL-23-045
Download: ML23311A044 (47)

Text

• ~ TENNESSEE VALLEY AUTHORITY Post Office Box 2000 Spring City, Tennessee 37381 WBL-23-045 November 7, 2023 10 CFR 50.59(d)(2)

U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Watts Bar Nuclear Plant, Units 1 and 2 Facility Operating License Nos. NPF-90 and NPF-96 NRC Docket Nos. 50-390 and 50-391

Subject:

Watts Bar Nuclear Plant Units 1 and 2-10 CFR 50.59 Summary Report

Reference:

TVA Letter to NRC, WBL-22-008 "Watts Bar Nuclear Plant Units 1 and 2-10 CFR 50.59 Summary Report," dated April 27, 2022 (ML20302A097)

Pursuant to Title 10, Code of Federal Regulations (10 CFR) 50.59(d)(2), Tennessee Valley Authority (TVA) is submitting a summary report of the changes, tests, and experiments implemented at the Watts Bar Nuclear Plant (WBN), Units 1 and 2 since the last 10 CFR 50.59 report was submitted on April 27, 2022 (Reference). The evaluations summarized in the enclosure cover the period from December 4, 2021, to May 12, 2023 and demonstrate that the described changes do not meet the criteria for license amendments as defined by 10 CFR 50.59(c)(2).

There are no new regulatory commitments in this letter. Should you have questions regarding this submittal, please contact Jonathan Johnson, Manager of Watts Bar Site Licensing, at jtjohnsonO@tva.gov.

Anthony L. Williams IV Site Vice President Watts Bar Nuclear Plant

U.S. Nuclear Regulatory Commission WBL-23-045 Page 2 November 7, 2023 Enclosure Watts Bar Nuclear Plant, Units 1 and 2 10 CFR 50.59 Summary Report cc: (Enclosure)

NRC Regional Administrator - Region II NRC Senior Resident Inspector - Watts Bar Nuclear Plant NRC Project Manager - Watts Bar Nuclear Plant

Enclosure Watts Bar Nuclear Plant, Units 1 and 2 10 CFR 50.59 Summary Report

1. Evaluation: 1-TI-44 Revision 3, Evaluation Revision 0

2. Evaluation: 2-TI-44 Revision 2, Evaluation Revision 0

3. Evaluation: 31814-EP-C-004 Revision 2, Evaluation Revision 1

4. Evaluation: Design Change Package (DCP) 56905A, Evaluation Revision 0

5. Evaluation: DCP 63651 Revision A, Evaluation Revision 1

6. Evaluation: DCP 66308 Revision A, Evaluation Revision 0

7. Evaluation: DCP 66476 Revision A, Evaluation Revision 1

8. Evaluation: DCP 66479 Revision A, Evaluation Revision 0

9. Evaluation: Design Change 100506 Revision 1, Evaluation Revision 1

10. Evaluation: Breaker Testing, Evaluation Revision 0

11. Evaluation: WBN-0-2020-031-00 Revision 4, Evaluation Revision 0

12. Evaluation: WBN-1-2022-030-001 Revision 0, Evaluation Revision 0

13. Evaluation: WBN-18-255, Evaluation Revision 0

14. Evaluation: WBN-19-352, Evaluation Revision 0

15. Evaluation: WBN-19-759, Evaluation Revision 1

16. Evaluation: WBN-19-760, Evaluation Revision 0

17. Evaluation: WBN-21-022 Revision 4, Evaluation Revision 1

18. Evaluation: WBN-21-047, Evaluation Revision 0

19. Evaluation: WBN-22-015, Evaluation Revision 0

20. Evaluation: WBN-22-060, Evaluation Revision 0

21. Evaluation: WBN-TS-22-09, Evaluation Revision 0 E1 OF E45 WBL-23-045

Enclosure Evaluation: 1-TI-44 Revision 3, Evaluation Revision 0 Activity

Description:

This activity is the change to the operator guidance provided for reactivity control during fuel cycle coastdown with a reduced RCS Tavg.

WBN Design Change WBN-21-022 analyzed for a reduced Hot Full Power (HFP) Tref range from 581.2 deg. F to 588.2 deg. F. This Design change includes a detailed 50.59 for the system changes necessary to provide this reduced temperature range. This 50.59 Screening is not meant to replicate that evaluation, but is focused only on the changes associated with guidance to operations during a reduced Tavg coastdown.

During a Reduced Tavg coastdown, the operator adjusts turbine load as necessary to allow Tavg to drift below Tref (but no lower than "Manual Tref'), thus providing positive reactivity to the core to compensate for the fact that Reactor Coolant System (RCS) boron dilution is no longer tenable at very low boron concentrations. The Manual Tref is a linear line as a function of thermal power between the no load Tref temperature of 557 deg. F and the HFP minimum Tref temperature of 581.2 deg. F as defined in WBN-21-022. This activity provides the detailed guidance to the operator for turbine control during the 3 stages of coastdown: 1) Temperature Coastdown, 2) Governor Valve Wide Open Coastdown, and 3) Power Coastdown. Maintaining Tavg between Tref and the Manual Tref keeps the RCS temperature within the bounds evaluated by WBN-21-022.

A reduced Tavg coastdown may result in Tavg deviating below Tref by greater than the Updated Final Safety Analysis Report (UFSAR) assumed 3.5 deg. F band. No automatic outward rod motion is initiated from the low Tavg - Tref mismatch because the control rods are above the C-11 setpoint. Should a plant upset occur, such as a turbine runback or other load reject, control rods remain available for automatic rod insertion. No steam dump actuation results from the Tavg-Tref mismatch because the Tavg is below Tref and because the steam dump system is not armed (Load Reject, Plant Trip, or Pressure Control) during steady state coastdown operation at power.

The reactor control unit (automatic rod control) is applicable to three design basis accidents: 1)

Dilution During Full Power Operation, 2) Single Rod Cluster Control Assembly Withdrawal at Full Power, and 3) Loss of External Electrical Load and/or Reactor Trip.

For the Dilution During Full Power Operation accident, with rod control in automatic, the dilution results in a temperature rise and control rods inserting, bringing in the rod insertion limit alarms, thus notifying the operators that a dilution event is in progress and preventing a reactor trip.

With rod control in manual and no operator intervention, the dilution results in a reactor trip on Over Temperature Delta Temperature (OTDT), but there are still 15 minutes available for operator action after the reactor trip to prevent a loss of shutdown margin.

For the Single Rod Cluster Control Assembly Withdrawal at Full Power, the event is assumed to be caused either by operator error or multiple electrical failures in the rod control system.

Both the automatic rod control system and the automatic steam dump system are applicable to the Loss of External Electrical Load and/or Reactor Trip design basis accident. The loss of load causes an increase in RCS temperature. With the automatic rod control system available, control rods would insert to mitigate the RCS temperature rise. With the automatic steam dumps E2 OF E45 WBL-23-045

Enclosure available, the dumps will prevent a significant heat up and will control RCS temperature close to Tref. With the automatic steam dumps not available, steam will still be dumped to the atmosphere. Even with the steam dumps, rod control, Steam Generator (SG) Power Operated Relief Valves (PORVs), Pressurizer (PZR) PORVs, PZR Spray, and direct reactor trip on turbine trip all unavailable, the SG and PZR Safety valves are sized to protect the RCS and SGs from overpressure. For this reason, no credit is taken for the operation of either the automatic rod control or the steam dumps in the analysis of this event.

Summary of Evaluation:

This 50.59 Evaluation concludes that this activity does not impact the frequency of likelihood of any assumed accident or malfunction of an Systems, Structures, and Components (SSC) because this activity does not affect the initiating mechanisms by which an accident is initiated or any SSC malfunctions. Similarly this activity does not create the possibility of an accident of a different type or a malfunction of an SSC important to safety with a different result than previously evaluated in the UFSAR. This activity does not contain any mechanism by which to negatively impact the operation of any SSC. This 50.59 Evaluation concludes that this activity does not increase the consequences of any assumed accident or malfunction of an SSC because the Tavg shall remain within the upper and lower limits assumed in the safety analyses. This 50.59 Evaluation concludes that this activity does not result in a design basis limit for a fission product barrier as described in the UFSAR being exceeded or altered and does not result in a departure from a method of evaluation described in the UFSAR used in establishing the design bases or in the safety analyses.

Based upon the results of this evaluation:

Implement the activity per plant procedures without obtaining a License Amendment.

E3 OF E45 WBL-23-045

Enclosure Evaluation 2-TI-44 Revision 2, Evaluation Revision 0 Activity

Description:

This activity is the change to the operator guidance provided for reactivity control during fuel cycle coastdown with a reduced RCS Tavg.

WBN Design Change WBN-21-022 analyzed for a reduced Hot Full Power (HFP) Tref range from 581.2 deg. F to 588.2 deg. F. This Design change includes a detailed 50.59 for the system changes necessary to provide this reduced temperature range. This 50.59 Screening is not meant to replicate that evaluation, but is focused only on the changes associated with guidance to operations during a reduced Tavg coastdown.

During a Reduced Tavg coastdown, the operator adjusts turbine load as necessary to allow Tavg to drift below Tref (but no lower than "Manual Tref'), thus providing positive reactivity to the core to compensate for the fact that RCS boron dilution is no longer tenable at very low boron concentrations. The Manual Tref is a linear line as a function of thermal power between the no load Tref temperature of 557 deg. F and the HFP minimum Tref temperature of 581.2 deg. F as defined in WBN-21-022. This activity provides the detailed guidance to the operator for turbine control during the 3 stages of coastdown: 1) Temperature Coastdown, 2) Governor Valve Wide Open Coastdown, and 3) Power Coastdown. Maintaining Tavg between Tref and the Manual Tref keeps the RCS temperature within the bounds evaluated by WBN-21-022.

A reduced Tavg coastdown may result in Tavg deviating below Tref by greater than the UFSAR assumed 3.5 deg. F band. No automatic outward rod motion is initiated from the low Tavg - Tref mismatch because the control rods are above the C-11 setpoint. Should a plant upset occur, such as a turbine runback or other load reject, control rods remain available for automatic rod insertion. No steam dump actuation results from the Tavg-Tref mismatch because the Tavg is below Tref and because the steam dump system is not armed (Load Reject, Plant Trip, or Pressure Control) during steady state coastdown operation at power.

The reactor control unit (automatic rod control) is applicable to three design basis accidents: 1)

Dilution During Full Power Operation, 2) Single Rod Cluster Control Assembly Withdrawal at Full Power, and 3) Loss of External Electrical Load and/or Reactor Trip.

For the Dilution During Full Power Operation accident, with rod control in automatic, the dilution results in a temperature rise and control rods inserting, bringing in the rod insertion limit alarms, thus notifying the operators that a dilution event is in progress and preventing a reactor trip.

With rod control in manual and no operator intervention, the dilution results in a reactor trip on OTDT, but there are still 15 minutes available for operator action after the reactor trip to prevent a loss of shutdown margin.

For the Single Rod Cluster Control Assembly Withdrawal at Full Power, the event is assumed to be caused either by operator error or multiple electrical failures in the rod control system.

Both the automatic rod control system and the automatic steam dump system are applicable to the Loss of External Electrical Load and/or Reactor Trip design basis accident. The loss of load causes an increase in RCS temperature. With the automatic rod control system available, control rods would insert to mitigate the RCS temperature rise. With the automatic steam dumps available, the dumps will prevent a significant heat up and will control RCS temperature close to Tref. With the automatic steam dumps not available, steam will still be dumped to the E4 OF E45 WBL-23-045

Enclosure atmosphere. Even with the steam dumps, rod control, SG PORVs, PZR PORVs, PZR Spray, and direct reactor trip on turbine trip all unavailable, the SG and PZR Safety valves are sized to protect the RCS and SGs from overpressure. For this reason, no credit is taken for the operation of either the automatic rod control or the steam dumps in the analysis of this event.

Summary of Evaluation:

This 50.59 Evaluation concludes that this activity does not impact the frequency of likelihood of any assumed accident or malfunction of an SSC because this activity does not affect the initiating mechanisms by which an accident is initiated or any SSC malfunctions. Similarly this activity does not create the possibility of an accident of a different type or a malfunction of an SSC important to safety with a different result than previously evaluated in the UFSAR. This activity does not contain any mechanism by which to negatively impact the operation of any SSC. This 50.59 Evaluation concludes that this activity does not increase the consequences of any assumed accident or malfunction of an SSC because the Tavg shall remain within the upper and lower limits assumed in the safety analyses. This 50.59 Evaluation concludes that this activity does not result in a design basis limit for a fission product barrier as described in the UFSAR being exceeded or altered and does not result in a departure from a method of evaluation described in the UFSAR used in establishing the design bases or in the safety analyses.

Based upon the results of this evaluation:

Implement the activity per plant procedures without obtaining a License Amendment.

E5 OF E45 WBL-23-045

Enclosure 31814-EP-C-004, Rev. 2 Evaluation Revision 1 Activity Description Engineering Package 31814-EP-C-004 addresses the rigging and transport activities that are necessary to perform the construction activities associated with the replacement of the steam Generators (SGs) for Watts Bar Nuclear Plant Unit 2 (WBN-2). Included within the scope of this engineering package are Old Steam Generator (OSG) and Replacement Steam Generator (RSG) rigging and transport, and rigging of other RSG Project related items (e.g., Shield Building dome removed concrete sections, Steel Containment Vessel (SCV) plate sections, and SG compartment roof concrete sections), that involve use of a large capacity Terex CC8800 crawler crane (termed the Outside Lift System (OLS)). A mobile M2250 crane will be used in assembling/disassembling the OLS and will also be used to make adjustments to the OLS during OLS performance of certain heavy lifts (e.g., placement/removal of counterweights).

Consistent with the requirements of NUREG-0612, rigging activities associated with movement of heavy loads during OLS assembly, post-assembly load testing, operation, and disassembly will not be performed over spent fuel or while fuel is in the Unit 2 reactor core.

Although NUREG-0612 applies to permanently installed plant cranes, the use of the construction aid OLS and activities performed to assemble, load test, and disassemble the OLS are procedurally controlled and utilize safe load paths that meet the intent of NUREG-0612. OSG/RSG transport and other transport activities performed by Engineering Package 31814-EP-C-004 ensure no adverse effects are experienced by underground commodities/piping evaluated for the haul routes.

Summary of Evaluation RSG Project rigging activities by the large capacity Terex CC8800 crawler crane outside lift system (OLS) as addressed by Engineering Package 31814-EP-C-004 do not result in more than a minimal increase in the frequency of occurrence of an accident or likelihood of occurrence of a malfunction of an important-to- safety SSC previously evaluated in the Updated Final Safety Analysis Report (UFSAR). Radiological consequences of a drop of an OSG are evaluated to remain below the 10 CFR 100 and 10 CFR 50, GDC 19 limits and are bounded by the UFSAR Chapter 15 Waste Gas Decay Tank (WGDT) release. Therefore, the consequences of an OSG drop are bounded by the analyses in the UFSAR and do not increase the consequences of an accident or malfunction previously evaluated in the UFSAR. For the Unit 2 defueled condition when major Unit 2 RSG Project lifts by the OLS are performed, safe load paths for individually handling the OSGs, RSGs, Shield Building dome removed concrete sections, removed Steel Vessel Containment (SCV) plate sections, and SG compartment roof sections are designed to rig and lift these items in paths away from the vulnerable areas in the yard where underground Essential Raw Cooling Water (ERCW) Supply Headers 2A and 2B, and Class 1E emergency power duct banks are located. Instead, with the maneuverability obtained by using a large capacity mobile Terex CC8800 crane, only the ERCW Discharge Headers are potentially affected. This ensures that in the unlikely event of a postulated drop of any OLS major lifted load, any effects upon the ERCW Discharge Header piping located north of the Condensate Demineralizer Waste Evaporator (CDWE) Building will maintain both trains of ERCW System supply header piping (Supply Headers 1A, 1B, 2A, and 2B) unaffected and capable of providing heat removal from required plant equipment supporting Unit 1 operation and spent fuel storage pool cooling and water level. Plant procedures and installed instruments are in place and E6 OF E45 WBL-23-045

Enclosure available for detecting any leakage in ERCW piping and to successfully respond to indications where ERCW pipe damage is suspected. As a result, a postulated drop of an OLS major lifted load remains bounded by accidents evaluated in the UFSAR. Therefore, a load drop from the OLS will not result in a different type of accident or a malfunction with a different result.

Rigging and transport activities implemented by Engineering Package 31814-EP-C-004 have no impact on a design basis limit for a fission product barrier (DBLFPB) as described in the UFSAR (e.g., Steel Containment Vessel (SCV) Containment Pressure, Reactor Coolant System design pressure, centerline melting temperature of the nuclear fuel). Accordingly, no DBLFPB as described in the UFSAR is exceeded or altered. The rigging and transport activities implemented by Engineering Package 31814-EP-C-004 meet the intent of NUREG-0612 principles for controlled load handling performed by temporary equipment, and therefore do not involve a departure from a method of evaluation described in the UFSAR used in establishing the design bases or in the safety analyses.

Results Implement the activity per plant procedures without obtaining a License Amendment.

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Enclosure DCN 56905A Evaluation Revision 0 Activity Description DCN 56905A replaces analog control with digital control with the Distributed Control System (DCS) on Watts Bar Nuclear (WBN) Unit 1 for control of: Main Steam Generator (S/G) Pressure (Atmospheric Dump Valves Control), Steam Dump Pressure, Hotwell Level Dumpback and Makeup, Steam Generator Slowdown Flow, Generator Hydrogen Temperature, Main Turbine Oil Tank Temperature, Cold Leg Accumulator Nitrogen Supply Valve, Residual Heat Removal (RHR) Heat Exchangers flow, RHR Letdown, and Chemical and Volume Control System (CVCS) (Excess Letdown flow, Letdown Heat Exchanger outlet temperature and pressure, and Boric Acid Tanks A and C recirculation flow). The new control systems address obsolescence of existing control system equipment and remove many single points of failures which will improve the reliability of the involved control systems on WBN 1.

Summary of Evaluation The new digital DCS system replaces function-for-function existing analog components for S/Gs (1 through 4) Atmospheric Dumps valves, Steam Dump pressure and temperature, Cold Leg Accumulator (CLA) Nitrogen Supply valve, Residual Heat Removal (RHR) Heat Exchanger (A, B, and Bypass) flow, Boric Acid Tanks (A and C) recirculation flow, Excess Letdown Flow, Letdown Heat Exchanger outlet temperature and pressure, S/G Slowdown flow, Hotwell level (Dumpback and makeup), Generator Hydrogen temperature, and Main Turbine oil tank temperature with many reliability improvements. The new system provides redundant inputs, redundant processors, redundant networks, redundant power supplies, etc. The new system is designated as "Quality Related" and is designed to meet Quality Related requirements; the reliability of the new system is superior to the old analog system. The modification does not negatively impact any SSC that is important to safety nor does it impact the consequences or the frequency of malfunction. The new Distributed Control System (DCS)S does not create a new type of malfunction or accident. The new (DCS) reduces the likelihood of failures and their consequences by providing a more reliable and redundant control system. In addition, this modification provides the capability to reduce manual operator actions, thereby, allowing greater operator opportunity for assessment, monitoring and response.

The upgrade to DCS results in overall improvement in the plant and the ability to function with individual devices out of service as:

DCS provides for use of additional input signals (as available) for control. The DCS will continue to maintain function with the loss of a single input for controls with multiple inputs. In the case of a single input the last good value will be used prior to the failure.

The DCS will provide an alarm on the DCS Visual Display Unit (VDU) for loss of an input, "DCS Trouble Alarm", and in some cases initiate the "DCS Critical Loop Alarm" on 1-M-4.

The DCS is powered from redundant power sources thus for loss of any single power supply or power source the DCS will continue to maintain control.

The signal output to plant control devices, such as valves, use redundant Field Bus Modules (FBM) such that should one FBM fail the other FBM maintains the control of the device.

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Enclosure For important functions such as SG Atmospheric Dump Valves (ADV)s and Steam Dump to Condenser each SGADV is on a separate DCS processor pair and the Steam Dump to Condenser is on a processor pair separate from the ADVs.

Results Implement the activity per plant procedures without obtaining a License Amendment.

E9 OF E45 WBL-23-045

Enclosure DCN 63651A, Evaluation Revision 1 Activity Description In the past when a Tritium Producing Burnable Absorber Rod (TPBAR) consolidation effort was completed, all of the TPBAR canisters. whether filled or partially filled, were shipped off-site to the processing facility. A filled TPBAR canister will hold up to 300 TPBARs. In order to save shipping cost associated with sending partially filled TPBAR canisters to the processing facility, consideration is now being given to storing any partially filled TPBAR canister in the SFP until the next refueling outage when it can be filled to its limit and then shipped to the processing facility. Due to concerns of potential TPBAR damage from foreign material entering the canister while stored in the SFP, a Foreign Material Exclusion (FME) cover may be installed over a tritium canister occupied spent fuel pool rack cell to protect the TPBARs.

WBN is licensed to burn up to 2304 TPBARs in a given cycle. In the interim between removing TPBARS from fuel and shipping the full TPBAR canisters to the processing facility, WBN may elect to cover the full TPBAR canisters with an FME cover.

The FME cover will protect a TPBAR canister from the entry of large foreign material (FM) such as badges, lanyards, tools. etc. while being stored in a Spent Fuel Pool (SFP) cell. It is not the goal or function of this cover to prevent very small FM from entering the TPBAR canister. The cover will be constructed from stainless steel and designed to rest on the top of the SFP cell which the TPBAR canister is contained. The cover when installed will protrude less than 1 inch above the racks and will not protrude into adjacent SFP cells. This will ensure the cover will not impact fuel movement in the SFP including the adjacent SFP cells.

The FME cover is perforated to allow cooling water flow through the TPBAR canister and will prevent large FM from falling into the TPBAR canister. Use of the FME cover is to be integrated with the existing TPBAR procedures and utilizes existing consolidation tools. The FME cover has been designed to interface with any of the cells of either the 7x8 or the 7x9 rack assemblies.

SSCs Impacted by the change:

Spent Fuel Pool Cooling and Cleaning System:

Past practice has been to ship off all TPBARS obtained from the previous cycle to the processing facility. Extended storage of TPBARS in the spent fuel pool raises concerns with Spent Fuel Pool FME and TPBAR heat load, both within the canister and relative to the spent fuel pool cooling and cleaning system.

Spent Fuel Racks The spent fuel racks were designed for storage of up to 1386 fuel assemblies. The spent fuel racks are designed as free standing and are qualified as seismic Category I. The racks can withstand the drop of a fuel assembly from its maximum supported height and the drop of tools used in the pool. The canisters are designed for storage in the spent fuel racks.

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Enclosure Consolidation Equipment The consolidation equipment is provided to allow consolidation of TPBARs in a manner that insures existing fuel handling equipment functional and design requirements are met.

Consolidation equipment must meet seismic qualification, environmental compatibility, materials compatibility. NUREG-0612, and As Low As Reasonably Achievable (ALARA) requirements.

Watts Bar Nuclear Plant is to design and utilize FME covers to keep foreign material out of TPBAR canisters stored in the Spent Fuel Pool.

Summary of Evaluation FME covers were evaluated for effects on the Spent Fuel Racks, the TPBARs. and long term (more than 18 month) storage of TPBARs. They are acceptable and do not require a change to the Watts Bar Operating License or Technical Specifications.

Results Implement the activity per plant procedures without obtaining a License Amendment E11 OF E45 WBL-23-045

Enclosure DCN 66308, Rev. A, Evaluation Rev. 0 Activity Description In performing the Watts Bar Unit 2 Replacement Steam Generator (RSG) Project, the scope of Design Change Notice (DCN) 66308, Reactor Building Structural Modifications, addresses the modifications to the Shield Building concrete dome, the Steel Containment Vessel (SCV), and the steam generator compartment concrete roofs that are necessary to support removal of the Old Steam Generators (OSGs) and installation of the RSGs. To facilitate removal of the OSGs and installation of the RSGs, Engineering Package 31814-EP-C-002, Temporary Structures and Commodities for the Removal and Restoration of the Shield Building Concrete Dome and Steel Containment Vessel, will cut openings in the concrete Shield Building dome, the SCV, and the steam generator compartment roofs. Following completion of steam generator rigging and handling that is performed by Engineering Package 31814-EP-C-004, Rigging and Transport, the openings in the Shield Building, SCV, and steam generator compartment roofs will be restored by implementation of DCN 66308.

The four openings in the steam generator compartments roofs will be restored by reinstalling the cut sections of the roof in their respective roof openings and using through-bolted connection frames to hold the concrete sections in place.

The two openings in the SCV will be restored by welding the cut steel pieces back in place.

The activities implemented by DCN 66308 will restore the openings in the Shield Building concrete by splicing new reinforcing bars to the existing reinforcing bars using mechanical rebar couplers, cadwelds, and/or welding, and pouring new concrete to close the openings.

Summary of Evaluation DCN 66308 restores the openings in the Unit 2 Shield Building concrete dome, the SCV, and the steam generator compartment concrete roofs to an acceptable design condition to meet the functional requirements performed by these structures. Restoration of the openings in the Unit 2 Shield Building concrete dome is performed by splicing new reinforcing bars to the existing reinforcing bars using mechanical rebar couplers, cadwelds, and/or welding, and pouring new concrete to close the openings. The two openings in the SCV will be restored by welding the cut steel pieces back in place. The four openings in the steam generator compartments roofs will be restored by reinstalling the cut sections of the roof in their respective holes and using through-bolted connection frames to hold the concrete sections in place. In reestablishing the functional qualifications of these structures by the implementing the DCN 66308 restoration design for these openings, the proposed modifications do not increase the frequency or likelihood of accidents or malfunctions, increase the consequences of an accident or malfunction, or create a new type of accident or malfunction. The SCV and Shield Building are fission product barriers, but they will be restored to meet their design basis limits, and therefore no design basis limits will be altered or exceeded. Implementation of DCN 66308 does not involve a departure from a method of evaluation described in the UFSAR used in establishing the design bases or in the safety analyses.

Results Implement the activity per plant procedures without obtaining a License Amendment.

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Enclosure DCN 66476, Rev. A, Evaluation Rev. 1 Activity Description In support of the Watts Bar Unit 2 Replacement Steam Generator (RSG) Project, Design Change Notice (DCN) 66476, Watts Bar Unit 2 Replacement Steam Generator (RSG) - SG Vessel Impacts, documents effects to the plant from replacing the existing Watts Bar Unit 2 Westinghouse Model D3 Steam Generators (termed Old Steam Generators (OSGs) with Westinghouse Model 68AXP Steam Generators.

SG Vessel Design and Operation The Nuclear Steam Supply System (NSSS) performance parameters, design transients, systems, components, accidents, and nuclear fuel areas were either analyzed or evaluated to demonstrate that the applicable licensing criteria and requirements are satisfied in support of the Unit 2 RSG Project. In anticipation of Unit 2 Measurement Uncertainty Recapture (MUR) power uprate from original core power of 3411 MWt to 3459 MWt, these analyses and evaluations consider the effects of the Unit 2 RSGs at a core power level of 3459 MWt (Nuclear Steam Supply System (NSSS) power level of 3475 MWt (which includes 3459 MWt reactor core power+ 16 MWt of Reactor Coolant Pump generated heat). NRC approval of the Unit 2 Margin Uprate (MUR) power update is documented in ML20226A444, and the MUR power uprate was implemented in Unit 2 during the Fall 2020 refueling outage (U2R3). A detailed description of the analyses and evaluations performed for the Unit 2 RSGs can be found in WCAP-18167-P, Watts Bar Unit 2 Replacement Steam Generator Program NSSS Engineering Report. One of the analyses that were re-performed due to the Unit 2 RSGs was the Steam Generator Tube Rupture (SGTR) margin to overfill the steam generator.

Tavg Reduction and Associated Considerations The evaluation of a 2 degree F reduction in Tavg was required, since Watts Bar Unit 2 can potentially be operating at a core power level of 3459 MWt after the Unit 2 RSGs are installed.

Due to the larger heat transfer surface area of the Unit 2 RSGs, there is an increase in both the Unit 2 RSG steam temperature and steam pressure. Watts Bar Unit 2 is currently designed for a nominal secondary side pressure of 1200 psia, which is protected by the lowest Main Steam Safety Valve (MSSV) setpoint (minus tolerance). The current licensed Tavg is 588.2 degree F.

Evaluations indicated that the combination of the Unit 2 RSG design and expected operating conditions (core power 3459 MWt) could lead to a high secondary side steam pressure, which could be high enough to potentially challenge the MSSV setpoints of 1200 psia minus tolerance on a reactor trip.

To offset the potential high secondary side steam pressures, a revised set of NSSS parameters were generated with a 2 degree F reduction in reactor coolant vessel average temperature to 586.2 degree F Tavg* Both the reactor coolant vessel outlet temperature (THot) and the vessel inlet temperature (TCold) were reduced such that the impact on delta T (THot - T Cold) was only

+0.2 degree F. The purpose of the evaluations performed in WCAP-18167-P was to allow Watts Bar Unit 2 to operate at the lower Tavg of 586.2 degree F.

One of the evaluations performed to support a 2 degree F reduction in reactor coolant vessel average temperature to 586.2 degree F was an evaluation of the reactor internals system structural response and integrity.

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Enclosure Summary of Evaluation SG Vessel Design and Operation The results of the Westinghouse NSSS analyses and evaluations demonstrate that the applicable licensing criteria and requirements are satisfied for the NSSS performance parameters, design transients, systems, components, accidents, and nuclear fuel at a bounding NSSS power level of 3475 MWt (in-process MUR) power uprate from reactor core power of 3411 MWt to 3459 MWt, NSSS power (3475 MWt) is 3459 MWt + 16 MWt of Reactor Coolant Pump generated heat). In performing these analyses, it was identified that the revised Steam Generator Tube Rupture (SGTR) margin to overfill analysis resulted in a reduced margin to overfilling a Unit 2 RSG experiencing an SGTR Event.

The SGTR margin to overfill analysis was reanalyzed for the Unit 2 RSGs. The Unit 2 RSGs have less of a secondary side unfilled volume than the current Unit 2 steam generators, which reduces the margin to overfilling the SG following an SGTR. The reduction in secondary side unfilled volume required a faster operator response time to isolate Auxiliary Feedwater (AFW) to prevent overfilling the SG following an SGTR. Based on similar evaluations performed for the Unit 1 RSGs, TVA performed two sets of simulator exercises to ensure that faster operator response time to isolate AFW could be performed prior to overfilling the SG. The margin to SG overfill was reduced in the SGTR reanalysis, however the reanalysis demonstrated that the SGs will not be overfilled following an SGTR and thereby prevent water from being relieved from a Main Steam Safety Valve (MSSV), which is the acceptance criterion for the analysis. The margin to overfill the SG is approximately 122 cubic feet. Therefore, the decrease in the margin to overfill the SG following an SGTR is acceptable for the Unit 2 RSGs.

Steam Generating Team (SGT) re-performed the SGTR offsite and Control Room dose analysis based on primary and secondary side steam releases provided by Westinghouse. For the Main Control Room (MCR) thyroid dose the pre-accident iodine spike (14 µCi/gm max peak) case, the resultant accident radiation dose increase is greater than 10 percent of the difference between the current SGTR dose analysis value and the regulatory limit. A License Amendment was obtained from the NRC to approve the greater than 10 percent increase in dose from an SGTR (ML21334A389). License amendments were also received from the NRC to address changes in SG secondary side water level (ML21260A210) and to address changes in SG inspection requirements (ML21306A287).

Tavg Reduction and Associated Considerations The results of the Westinghouse NSSS analyses and evaluations demonstrate that the applicable licensing criteria and requirements are satisfied for the NSSS performance parameters, design transients, systems, components, accidents, and nuclear fuel at the bounding NSSS power level of 3475 MWT (reactor core power of 3459 MWt + 16 MWt Reactor Coolant Pump generated heat with a 2°F reduction in reactor coolant vessel average temperature to 586.2 degree F. However, the decrease in vessel/core fluid temperature due to the Tavg reduction will result in the hydraulic lift forces increasing by less than 1.0 percent at Mechanical Design Flow (MDF). There is sufficient margin available to offset the increase in the lower internals hydraulic lift forces due to the decrease in vessel/core inlet temperature.

Therefore, the lower internals will remain seated and stable for the 2 degree F Tavg reduction, and capable of performing their design function.

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Enclosure Based on this evaluation, and the amendments received, this activity may be performed without additional NRC approval.

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Enclosure DCN 66479, Rev. A, Evaluation Rev. 0 Activity Description The Watts Bar Unit 2 Replacement Steam Generator (RSG) Project replaces the Unit 2 old (Westinghouse Model D3) steam generators (old steam generators (OSGs)) with Westinghouse Model 68AXP RSGs that possess greater heat transfer area and improved material construction, which are evaluated for installed operation by Design Change Notice (DCN) 66476, Watts Bar Unit 2 (WBN2) Replacement Steam Generator (RSG) - SG Vessel Impacts.

DCN 66479, WBN2 Replacement Steam Generator- Balance of Plant Impacts, evaluates the effects that the increases in secondary side pressure and temperature due to the larger heat transfer area of the RSGs have upon the Main and Reheat Steam system and the remaining Balance of Plant (BOP) systems, structures, and components (SSCs). The 10 CFR 50.59 Screening Review prepared for DCN 66479 identified the following design functions as being adversely affected and requiring further review by performing a 10 CFR 50.59 Evaluation. All other BOP aspects of RSG installed operation screened out.

With the RSGs installed, the operating pressures in non-safety related portions of the Main Feed Water (MFW) system piping at full load operation will be greater than the current design pressure for this piping (1185 psig). To address this concern, DCN 66479 increases the design pressure from the MFW Pump outlet piping through the FW heaters to 1250 psig, and increases the design pressure to 1230 psig for the piping from the FW heater outlet to the FW piping class break.

The Extraction Steam System Number 4 Extraction Steam Line is expected to operate above its current design temperature of 360 degree F. The design temperature of this line will be raised to 450 degree F to accommodate operation with the RSGs.

With the RSGs installed, the Main Feed Pump Turbine (MFPT) Condenser Drain Tank is expected to operate at a temperature slightly above its 150 degree F design temperature. Allowance for operating above the 150 degree F design temperature is addressed in this 10 CFR 50.59 Evaluation.

Summary of Evaluation DCN 66479 evaluations of the BOP effects from installed operation of the Unit 2 Replacement Steam Generators (RSGs) identified the following for review by performing a 10 CFR 50.59 Evaluation: 1) increasing the design pressures of non-safety related portions of Main Feedwater (MFW) system piping to accommodate expected higher operating pressures of this piping for effective flow of feedwater into the RSGs at their expected higher Main Steam pressure, 2) increase the design temperature of the Extraction Steam System Number 4 Extraction Steam Line to reflect RSG operation, and 3) permit operating the Main Feed Pump Turbine (MFPT)

Condenser Drain Tank above its 150 degree F design temperature for expected RSG operation.

These activities are evaluated to demonstrate compliance of these BOP SSCs with their respective design codes and to conclude that the reliabilities of these SSCs to perform their intended functions remain acceptable. Therefore, these design pressure and temperature changes do not increase the frequency or likelihood of accidents or malfunctions, increase the consequences of an accident or malfunction, or create a new type of accident or malfunction.

The design pressure of the non-safety related portions of the MFW system piping, and the E16 OF E45 WBL-23-045

Enclosure design temperatures of the number 4 extraction steam line and MFPT Condenser Drain Tank do not constitute fission product barriers. The analyses performed by DCN 66479 do not involve a departure from a method of evaluation described in the UFSAR used in establishing the design bases or in the safety analyses.

Results Implement the activity per plant procedures without obtaining a License Amendment.

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Enclosure WBN DC-100506 Rev. 1, Evaluation Rev. 1 Activity Description The purpose of this Design Change (DC) is to correct the Degraded Non-Conforming Condition identified in CR 1316395. The CR explains that during development of the Proto-Flo hydraulic model of the Essential Raw Cooling Water (ERCW) system, a design basis issue was identified concerning the design and analysis of the Component Cooling System (CCS) Heat Exchanger (HX) ERCW discharge header.

The limiting ERCW design basis accident assumed one Unit in Hot/Cold Standby on Residual Heat Removal (RHR) cooling when the other unit experiences a Loss of Coolant Accident (LOCA) concurrent with a complete loss of power of an entire shutdown train. However, after further evaluation, it was determined that due to the current layout of the CCS HX ERCW discharge header (CCS HX A and C both discharging to the B ERCW discharge header), the loss of an entire shutdown train was not limiting, and instead, the loss of a single active component on the CCS or ERCW Train B supply header could present a more limiting scenario.

In this scenario, ERCW Train B supply flow could continue to be discharged to the B ERCW discharge header through CCS HX C creating unacceptable back-pressure on the B ERCW discharge header and preventing adequate flow through Train A CCS HX A.

In response to this design basis issue and the inadequacy of the current ERCW configuration to ensure minimum required flow to all cooling loads, a Prompt Determination of Operability (PDO) was developed (attached to CR 1316395) which justifies continued operability with lower required ERCW flowrates through numerous ERCW supplied heat exchangers.

This DC resolves the design basis problem from CR 1316395 described above by:

1. Changing the position of CCS Heat Exchanger C outlet flow control valve 0-FCV-067-0152-B (152-B valve) by moving it upstream of its existing location and prior to the CCS HX ERCW discharge header cross-tie.

2. Cross-tying the A and B ERCW discharge headers downstream of the CCS HXs to balance flow across the discharge headers and prevent excessive backpressure on ERCW discharge header B.

3. Rebalancing ERCW header flow to ensure adequate cooling to all users under all design basis scenarios.

4. Taking credit for a modified operator action to start a 3rd ERCW pump on Train A ERCW supply regardless of the status of Train B emergency power during the limiting design basis accident.

Power and control for the relocated 152-B valve will be provided by extending the existing power and control cables such that the 152-B valve performs the same design functions as it currently does. Electrical analysis shows that addition of a 14.2 sec. time delay in the 0-FCV 152-B valve open logic is needed to ensure adequate voltage to the 152-B valve after receipt of an Safety injection (SI) signal.

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Enclosure The Screening Review determined that the addition of the 14.2 sec. time delay and the modification of the credited operator action to start a 3rd ERCW pump have an adverse effect.

Therefore, this evaluation will be limited to those items. This evaluation will not address the relocation of 0-FCV-067-0152-B, cross-tying of the discharge headers, or the ERCW flow balance changes.

Summary of Evaluation The addition of the 14.2 sec. time delay relay within the electrical circuit of 0-FCV-067-0152-B does not require prior NRC approval. The changes to the credited manual action (1-E-1 and 2-E-1, step 23.b) do not require prior NRC approval.

Results Implement the activity per plant procedures without obtaining a License Amendment.

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Enclosure TRM and FSAR Breaker Testing, Evaluation Rev. 0 Activity Description Per 10 CFR 50.59 Screening review for Technical Requirements Manual (TRM)/FSAR Revisions for WBN Testing Requirements, only Question 3 for revising UFSAR described Methodology was marked Yes. This requires only Question 8 of the 50.59 Evaluation to be answered.

UFSAR Appendix 8C, Probability/Reliability Analysis of Protection Device Schemes for Associated and Non-Class 1E Cables, provides analysis to verify that the reliability of (1) a circuit breaker and a fuse in series, or (2) two circuit breakers in series (without periodic testing) is essentially equal to the reliability of a single circuit breaker periodically tested.

Although no revisions are required to the analysis which will affect failure probability of the breaker/fuse or breaker/breaker combinations, the assumption for a single breaker tested every 18 months will be revised to provide clarification of testing and how breaker styles will be covered under the individual Molded Case Circuit Breakers (MCCB) Surveillance Instruction (SI)/ Preventive Maintenance (PM) program. The assumptions are made to provide a standard of comparison for the analysis.

The assumptions for the single breaker testing used as the standard for comparison for the Breaker/Breaker and Breaker/Fuse combinations will be revised from "Each circuit breaker is tested every 18 months" to" Each circuit breaker style (Mechanical, Magnetic, Thermal magnetic, etc.) is tested approximately every 18 Months."

The revisions to the assumptions will not effect specific reliability numbers calculated using IEEE 500-1977. The reliability numbers for breaker/breaker and breaker/fuse will remain the same. In addition, the standard reliability calculated for a single breaker will remain the same.

These numbers were calculated using IEE 500-1977 failure rates based on industry trends and analysis. The assumption for single breaker testing will be enhanced to better describe the program testing. Currently, only schemes which utilize a single breaker as isolation are tested.

The change does not affect breaker/breaker or breaker/fuse combinations, as those combinations are not included in the testing program. So revising the assumption for Single Breaker testing does not change the conclusions of the analysis that breaker/breaker or breaker/fuse combinations have high reliability and are not required to be tested.

The current sampling program tests 10% of each breaker type every 18 months (Not each circuit breaker every 18 months as described in the assumption of Appendix 8C). Breakers are selected such that a 100% of the required test population is completed over a 10 test cycle interval. This means the longest historical interval for an individual breaker to be tested (including margin) is approximately 18 years. This establishes a basis for a nominal 18 year test frequency for individual breakers. The proposed UFSAR and TRM revisions will replace the 10 percent sampling frequency at which functional testing and maintenance is performed for MCCBs and metal-clad circuit breakers with a statement that the frequency will be in accordance with 0-Tl-109, nominal 18 years for each breaker. This change does not remove the requirement to periodically test breakers.

Procedure 0-Tl-109 is the WBN Breaker Testing and maintenance program and serves to document the technical basis for the program. The procedure will be revised to include the TRM and FSAR breaker requirements for testing and maintenance. The current TRM and FSAR E20 OF E45 WBL-23-045

Enclosure requirements (10% Sampling) were based on manufacturer and industry guidelines that were available at the time and accepted by the NRC during issuance of WBN's originating operating license (Safety Evaluation Report- NUREG-0847).

Since that time the industry has developed programs for managing the aging of Important equipment which includes input from the Original Equipment Manufacturer (OEM). The revisions are intended to provide the plant with more flexibility to consider maintenance history data and industry best practices to help perform the right maintenance on the right equipment at the right time. In addition, individual breaker PMs/Sls in Maximo will provide a mechanism to more efficiently and accurately to track completion of individual breaker testing.

The revisions will remove the specific sampling program requirements that exist for circuit breaker testing. This sampling program has been in place for WBN's full history, therefore, the population for sampling now includes various breaker types. The historical site data along with manufacturer and industry guidance will be used to develop a technical basis fora fixed frequency testing program utilizing the WBN Surveillance process per NPG- SPP 6.9.2. The technical basis for the WBN Breaker Program and frequencies will be documented in the Breaker Testing and Maintenance Program procedure 0-Tl-109.

The current sampling program tests 10% of each breaker type every 18 months. Breakers are selected such that a 100% of the required test population is completed over a 10 test cycle interval. This means the longest historical interval fora individual breaker to be tested is 18 years. This establishes a basis for a nominal 18 year test frequency for individual breakers.

Because 10% of each breaker type was tested every 18 months, the base dates are distributed chronologically such that if failure trends of a particular model were to develop they would be observed and the testing frequency could be increased, insuring that common mode failures would be detected. Thus, future testing with due dates based on the previous test program individual last test date or program initial inclusion date will ensure breakers are spread out to get a representative sample to ensure reliability over the new test program life time.

Individual testing SI's/PM's will be generated for each breaker in the current lest population. A nominal frequency of 18 years will be utilized for testing purposes. The due dates for each breaker will be established to consider the base dates of past breaker testing or inclusion under the current sampling program. Proper application of the preventive maintenance process will ensure that there is no decrease in the reliability of the TRM breakers. The technical basis for the WBN breaker preventive maintenance program will be documented in the Breaker Testing and Maintenance Program procedure 0-Tl-109. The changes to the TRM and FSAR will reference the 0-Tl-109 document and program for frequency and testing of breakers.

The proposed change does not change the intent of the breaker test program, only how it is tracked and implemented. Since the future individual breaker SI's will have a due date based off of either the last test date or inclusion into the program, breaker testing will be spread out over the 18 year testing period. This ensures breakers are periodically tested to capture any adverse trends. With each individual breaker having a SI that is tracked by Maximo, this gives more assurance that every breaker is captured into the program, rather than depending on a sampling population where a single person chooses breakers to test every 18 months. The functional test performed by the individual SI is not changed and will be performed in the same manner as already discussed in the TRM and FSAR A review of MCCBs tested under the current program has been performed for reliability of the breakers and test results. The review included the past 10 years of Condition Reports (CR)

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Enclosure related to breaker test deficiencies based on a Maximo search for Unique Identifiers (UNIDs) with "BKR". The total test population for Watts Bar includes approximately 2700 breakers, so a ten year interval or approximately 5 test cycles is sufficient for trending. Based on the review breakers have been reliable and no adverse trends identified.

Based on the sizable population of breakers in the test population, and the limited number of consequential test deficiencies noted, reliability of Watts Bar breakers have been proven and support current maintenance and testing practices. This reliability supports transition to the individual testing frequency for breakers. In addition to satisfactory test results, the Watts Bar practice over the past test performances is typically to remove the breaker to be tested and replace with a new breaker. Use of the removed breaker for subsequent breaker swaps/testing is allowed, but only after Breaker Engineer review to ensure individual breaker is not exhibiting a degraded trend in performance. In a vast majority of past breaker swaps, the removed breaker was tested and discarded. Thus, the breaker population at Watts Bar is relatively new, with new grease and technology not present with old breakers, which were available upon initial Watts Bar licensing or been reviewed to ensure no degrading condition exists. Thus, breakers installed would be even more reliable. These practices provide further justification of moving to a fixed, individual program versus the sampling program currently in place.

The revision does not adversely affect how the design functions of any SSCs are performed or controlled. Removing the percentage sample time based frequency will allow Watts Bar to schedule circuit breaker maintenance to more effectively utilize maintenance and plant resources. Testing results and maintenance will continue to be monitored for trends and the testing frequencies will be adjusted as necessary based on the data utilizing existing preventative maintenance, corrective action, and engineering programs which have been established in accordance with 10 CFR50.65 requirements. Testing results and frequencies will be monitored and adjusted based on periodic reviews of CRs and the breaker program guidance of 0-Tl-109.

Summary of Evaluation Conclusion is to implement changes to the TRM/FSAR Breaker testing program without obtaining a License Amendment The proposed UFSAR and TRM revisions will replace the 10 percent sampling frequency at which functional testing and maintenance is performed for MCCBs with a statement that the frequency will be in accordance with 0-Tl-109, nominal 18 years for each breaker. This change does not remove the requirement to periodically test breakers. Individual testing SI's/PM's will be generated for each breaker in the current test population. The sampling program described in the UFSAR of 10 percent Sampling over 18 months will be essentially the same as testing Individual breaker on a specific frequency, as the current test program has been in place since initial Licensing. Breakers in the program were evenly distributed over the testing cycles in the past. Future individual breaker testing SI's will have due dates based on their last performance. Thus, breakers will be evenly distributed over the testing timeframe. Thus, using the corrective action program, trends will be noted and capture any actions to address adverse trends in breaker issues. This will ensure coverage of various breaker styles, mechanical, magnetic trip, thermal magnetic trip, which incorporate similar technologies and ensure various types are tested approximately every 18 months. With the calculated reliability rates remaining the same for a breaker/breaker and fuse/breaker scheme, the enhanced wording to the assumptions does not change or affect the actual intent of precluding the breaker/breaker or breaker/fuse schemes from the testing program. Combined with the new individual breaker SI's which will be put in place to ensure reliability of singe E22 OF E45 WBL-23-045

Enclosure breaker schemes, the change does not result in a departure from the method of evaluation described in the UFSAR.

Results Implement the activity per plant procedures without obtaining a License Amendment.

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Enclosure Evaluation: WBN-0-2020-031-00 Revision 4, Evaluation Revision 0 Activity

Description:

This Temporary Modification (T-Mod) will provide a temporary chilled water supply to the B-Train Shutdown Board Room (SDBR) air handling units (AHUs), to be utilized during the time period with the permanent B-Train SDBR chiller out of service. This will occur during replacement of the B-Train SDBR chiller as part of DC WBN-19-479-04.

During the periods with the B-Train SDBR chiller out of service and once this T-Mod is installed, this temporary chilled water supply is planned to be used as the primary source for cooling the areas served by these air handling units, with the permanently-installed A-Train chiller and AHUs maintained in a standby mode to be available if the temporary chilled water supply becomes unavailable. The automatic starting of the A-Train SDBR air conditioning system upon low differential pressure across the B-Train SDBR chilled water pump is disabled with this T-Mod, as the permanently-installed chilled water circuit is not the supply of chilled water to the B-Train SDBR AHUs while this T-Mod will be in service and the B-Train chiller and chilled water pump are to be out of service with their breakers controlled via a clearance. This configuration is an activity in support of maintenance, but a 50.59 Screening and Evaluation has been completed in advance in the event of any delays relating to the installation of the new B-Train SDBR chiller. This will not be a permanent configuration.

Normal operation involves one of either the A-Train or B-Train SDBR air conditioning systems supplying cooling to the areas served by the SDBR chillers, with the opposite train serving as a backup. The normal configuration includes provisions for automatic swap-over in the event of either system becoming unavailable. The systems include 2 A-Train and 2 B-Train AHUs, with 1 AHU of each train supplying the respective Unit 1 side or Unit 2 side ductwork from a respective shared mechanical equipment room. The ductwork downstream of the AHUs is shared by A-Train and B- Train as shown on drawing 0-47W866-3. The temporary chilled water supply provided by this T-Mod will be connected to each respective B-Train AHU.

The A-Train SDBR air conditioning system will remain in standby and will be placed into service automatically in the event that either B-Train SDBR AHU fan experiences a failure or high intake air temperature. Also, the A-Train SDBR air conditioning system can be manually placed into service if any issues with providing the chilled water supply to the B- Train AHUs are detected prior to any conditions triggering an automatic start.

Summary of Evaluation:

Accidents evaluated in Chapter 15 of the UFSAR for Watts Bar will not increase more than minimally in likelihood as a result of this T-Mod, due to considerations taken in the design for the implementation of this T-Mod. Malfunctions to SSCs important to safety will not see a more than minimal increase as operator rounds will monitor temperatures and the automatic startup of the A-Train SDBR air conditioning system is retained for B-Train SDBR AHU low flow or high air intake temperature. Radiological consequences from neither accidents nor from malfunctions of SSCs important to safety will increase more than minimally as a result of the installation and use of this T-Mod. Barriers protecting against radiological doses will not be impacted and no missions outside of the main control room will be necessary to operate the SDBR air conditioning system in the manner presented by installation and use of this T-Mod. The design considerations included as part of this T-Mod, in conjunction with the plants existing design considerations, preclude any new accident types or malfunctions to SSCs important to safety E24 OF E45 WBL-23-045

Enclosure from being introduced by the installation and use of this T-Mod. No fission product barriers are impacted by the installation or use of this T-Mod, and no methods of evaluation are impacted by this T-Mod, as determined in question 3 of the 50.59 Screening for this T-Mod.

Based upon the results of this evaluation:

Implement the activity per plant procedures without obtaining a License Amendment.

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Enclosure WBN-1-2022-030-001, Evaluation Rev. 0 Activity Description Control Rod Drive Mechanism (CRDM) Cooler 1D-B tripped unexpectedly on 2/7/22.

Troubleshooting identified a hard ground on 1-MTR- 30-80/1-B as the cause of the trip. This motor is located inside the Polar Crane Wall and is inaccessible during plant operation and requires an outage to replace. Furthermore, Appendix R requires the CRDM Coolers to be available to perform a required Fire Safe Shutdown (FSSD) function. This motor failure has caused Operations to enter OR 14.10.

T-MOD WBN-1-2022-030-001 will temporarily disable the failed motor of CRDM Cooler 1D-B to allow CRDM Cooler Fan 1D-B Fan #2 to be operated without Fan #1. This will be done by removing the six primary disconnect finger clusters from the rear of 1-BKR-30-80/1 and insulating the line and load side stabs. This will allow the controls of CRDM 1D-B to operate as designed but will prevent 1-MTR-30-80/1-B from being energized. Disabling a fan stage will require this cooler to be maintained in Supplemental Lower Compartment Cooling ("bypass")

mode, which is the required alignment for an Appendix R event. Implementation of this Temporary Modification (TMOD) will restore FSSD functionality to CRDM Cooler 10- B and allow Operations to exit OR 14.10 Summary of Evaluation The Design function of the CRDM Coolers is to maintain acceptable temperature with the CRDM shroud for the protection of equipment and controls during normal reactor operation and normal shutdown. The CRDM Cooling system is designed to operate with the Lower Compartment Coolers (LCCs) to maintain a maximum air temperature with the upper reactor cavity of 120 degrees F and to route all of the reactor well air through the CRDM shroud to maintain a maximum air temperature of 185 degrees F. Air drawn through the CRDM shroud is cooled by the active fan-coil assemblies to approximately 120 degrees F and discharged into the lower compartment of the Reactor Building. When additional cooling in the lower compartment is required, the arrangement of dampers allows either or both standby CRDM fan-coil assemblies to recirculate air in the lower compartment general spaces and supplement the LCC system capacity.

The four CRDM air cooling fan-coil assemblies are located in the main lower compartment space at floor Elevation 702.78. Each assembly consists of a plenum, three air cooling coils, two vane-axial fans in series, air operated dampers, instruments, and controls. The four CRDM coolers are divided into two pairs. Cooler 1D-B is paired with 1A-A, and 1C-A is paired with 1B-B. One cooler in each pair is required to provide adequate cooling to the CRDM shroud during normal operation. Each fan motor has its own breaker, such that two breakers operate together for each CRDM Cooler. The control circuit for Fan 1 contains all the control logic and operates the breaker for Fan 1. There are no additional controls in the circuit for Fan 2, as its breaker is just "slaved" off the status of the Fan 1 breaker, such that the two breakers open and close together to control the double fan/motor assemblies of each cooler.

The CRDM Coolers and associated duct/dampers are not safety-related and are not required to perform a primary nuclear safety function. However, the CRDM Cooler along with the LCCs are required for safe shutdown, per 10 CFR 50 Appendix R to keep containment temperatures from exceeding operability limits on safe shutdown equipment inside containment. The CRDM coolers in conjunction with the LCCs also maintain the normal weighted average lower E26 OF E45 WBL-23-045

Enclosure compartment air temperature within upper limits specified in TS 3.6.5 in order to support initial conditions assumed in the Westinghouse accident analyses. 1-BKR-30-80/1 is safety related to protect 480V Shutdown Board 1B2-B.

This TMOD will disable CRDM Cooler 1D-B fan/motor set #1 by removing the six primary disconnect finger clusters from the rear of 1-BKR-30-80/1 and then insulating the breaker and/or switchgear side line side and load side contacts ("stabs"). These finger clusters are the components that connect the breaker to the stationary line and load side contacts in the switchgear itself. With these removed, there will be a gap between all primary contacts on the breaker and the switchgear. This will allow the breaker for Motor 1 to be racked in and operated without energizing its load (failed Motor 1). The breaker for Motor 2 will continue to follow Breaker 1 and will energize Motor 2 and allow it to be operated without energizing Motor 1. This configuration will still provide all the normal controls, indications and interlocks for CRDM Cooler 1D-B to function. Since only 1 motor is able to run, CRDM Cooler 1D-B will only be operating at reduced capacity and will only be available for supplemental cooling mode. This TMOD will maintain CRDM Cooler 1D-B aligned in supplemental cooling mode, which is the required alignment for the Appendix R event, by administratively controlling 1-TCO-30-81-B closed and 1-TCO-30-82-B open. Its paired cooler, 1A-A, will be available for normal operation at full capacity to provide cooling to the CRDM shroud.

The normal controls, interlocks and indications for CRDM Cooler 1D-B will remain unchanged.

The control circuit shown on drawing 1-45W760-30-8 will be unaltered. 1-BKR-30-80/1 will operate as normal but will no longer be connected to any load since the primary disconnect finger clusters have been removed. All status light indications, interlocks, alarms, and manual &

automatic control functions are unaltered and will continue to operate as designed. Annunciator 1-XA-55-5C-102A "CRDM COOLER FLOW LO" may alarm when CRDM Cooler 1D-B is operated in the bypass mode due to less than normal negative pressure at 1-FS-30-80A/B.

During normal operation, the required 2,600,000 BTU/hr of cooling is achieved by two CRDM Coolers operating together and aligned to the CRDM Shroud. This is unaffected by this TMOD.

CRDM Cooler 1A-A plus either 1B-B or 1C-A may be aligned to the CRDM shroud to provide the required cooling during normal operation. In addition, it is assumed other two CRDM coolers are set to bypass mode. Based on analysis found in NDQ0009992020000888 one of the two CRDM coolers has a failed vane-axial fan.

Calculation EDQ00099920090012 specifies the equipment required for 10CFR50 Appendix R Fire Safe Shutdown. Logic diagram K-37J-1 requires either 3 of 4 LCCs or 2 of 4 LCCs plus 2 of 4 CRDM Coolers in bypass mode to provide the required containment cooling to support FSSD.

Calculation WBN-APS2-070 determines the temperature response of Containment following an Appendix R fire. Two base cases are analyzed in this calculation along with 4 sensitivity cases which consider off-normal airflow rates of the various containment coolers. The two base cases match logic diagram K-37J-1. Base Case 1 considers 2 LCCs plus 2 CRDM Coolers (at design flow rates). Base Case 2 considers 3 LCCs (at design flow rates) and no CRDM Coolers.

Sensitivity Cases 2, 3, & 4 consider potential CRDM Fan stage failures. Sensitivity Case 2 considers 2 LCCs plus 1 CRDM Cooler at design flow and 1 CRDM Cooler at 9,000 CFM. This was done to bound conditions when one of the two in-series CRDM Cooler fans fails and results in a potential operating point on the fan curve of approximately 9,000 CFM. Sensitivity Case 3 considers 2 LCCs plus 2 CRDM Coolers each at 9,000 CFM. Sensitivity Case 4 considers 2 LCCs plus a single CRDM Cooler at 9,000 CFM. The computed temperature response for each case is compared to its applicable Environmental Qualification (EQ)profile and results show that E27 OF E45 WBL-23-045

Enclosure with five exceptions, all areas remain bounded by the EQ curve. The five exceptions are the lower reactor cavity, the upper reactor cavity, the upper containment, the lower containment, and the SG enclosures. The excursions were determined to be minor (less than 2% of event duration) or the demonstrated temperature profile for the EQ equipment located in these specific areas bounds the Appendix R event profile. This provides justification that these areas are also acceptable.

The normal operation sensitivity studies are analyzed in calculation NDQ0009992020000888.

The minimal CRDM fan case based on Base Case 2 evaluates the condition where one of the two vane-axial CRDM fans operating in series fails, resulting in a minimum flow rate of 9,000 acfm. This is analogous to the flow rate used in Appendix R WBNAPS2070 Appendix D Case 3 model (Ref. 22). The flow rate for the modified CRDM in normal operation is 33,925 acfm (30,953 scfm) and one of the bypass mode CRDMs is set to 9,000 acfm.

Therefore, the reduction in heat removal capability made by this TMOD remains bounded by this most limiting original sensitivity studies and this TMOD is acceptable for normal and Appendix R FSSD scenarios. Additionally, the PMT for this TMOD will operate CRDM Cooler 1D-B motor 2 in bypass mode and verify the motor amps are within the expected range.

Results Based on the results of this evaluation, this activity may be implemented without obtaining a license amendment.

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Enclosure WBN-18-255, Evaluation Rev. 0 This design change enhances security features at the Intake Pumping Station (IPS). LightLOC Smart Grate welded steel barriers are installed on the four IPS intake bays to detect and assess any intrusion attempts through the bays. Smart Grate is a lattice of steel tubing laced with optical fiber. The fiber is present in all components of the grating providing complete coverage in the event of an attempted breach Accessing the secure area protected by the grating requires compromising the grate and, thus. the detection fiber. resulting in an alarm. The Smart Grates are installed in the existing stop log slots. The Smart Grates are loosely fit into the slots without being fastened to the structure. Detection fibers from each grate system are routed from a water-tight junction box at the top of each grate to an enclosure above each bay. Fiber is routed in conduit through the IPS to an electrical cabinet containing LightLOC Express 2 modules. The modules are connected to the security system through a new Local Intelligence Unit (LIU).

Additionally, this design change installs seven thermal security cameras for assessment. Four of the cameras are located on elevation 741 feet of the IPS near the traveling water screens. The other three cameras are mounted outside the IPS on camera towers 26B1/26B1A, 26B2. and PTZ-8. Cables and fiber are routed in conduit from the cameras to camera electrical cabinets. A spare fiber is used to connect the camera feeds to the Central Alarm Station (CAS) and Secondary Alarm Station (SAS).

Summary of Evaluation Addition of the LightLOC Smart Grates to the four intake bays of the IPS have been evaluated because they impede flow to the pumps housed in the IPS. Effects of the grates on the design basis functions of ERCW Pumps, RCW Pumps, and High Pressure Fire Protection Pumps have been evaluated and determined that they can be implemented without prior NRC approval.

Results Implement the activity per plant procedures without obtaining a License Amendment.

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Enclosure WBN-19-352, Evaluation Rev. 0 Activity Description WBN-19-352 Rev. 0 installs equipment which TVA defines as Category C (Distributed Control System Equipment & Valve Controllers) and Category E (Transmitters) digital equipment, based on TVA's procedure SS-E18.15.01, "Requirements for Digital Systems (Real-Time Data Acquisition and Control Computer Systems)". WBN-19-352 involves a digital upgrade to the level control for the WBN Unit 2 Heater Drain Tanks (HDTs) and Feedwater Heaters (FWHs) to eliminate and harden Single Point Vulnerabilities (SPVs). The field devices (level transmitters) for the HDTs and FWHs will provide input to the local Foxboro I/A Distributed Control System (DCS) remote 1/0 field cabinets. The Foxboro I/A DCS will use these inputs to send an analog control output to each digital valve controller (DVC) to modulate valve position, discrete contact output signals to control pumps, use Highway Addressable Remote Transducer (HART) communication for valve position feedback and diagnostic information, and output a discrete contact closure for annunciation. Throughout this 50.59 the terminology DCS is used referencing the Foxboro I/A system and/or the replacement system as a whole, pending the application described within the statement.

This modification will require several tasks which include:

HDT Scope o Adding level switches for HDT-3 & HDT-7 (software level switches) and removal of existing physical HDT level switches.

o Software upgrade of:

Alarm Indication/Annunciation - Group 7 HDT Group 8 HDT Group 9 HDT-3, HDT-7, FWH-1, and FWH-2 Pressure and Flow Indication - Group 10/11 o Replacement of Field Control Processors (FPC) FCP270s to FCP280s in Cabinet 2-R-21 and 2-R-26 (Groups 8 & 9).

o Upgrading existing level modifiers (LMs) to Fisher DVC6200 positioners o Adding additional Input/Output (I/O) to existing cabinets 2-L-900 and 2-L-901 for the new HDT functions.

o Adding relays for the HDT pumps start/stop permissive.

o Evaluating a generic substitution of the Rosemount 3051C Differential Pressure (DP) Transmitters with Rosemount 3051S FWH Scope o Replacing level indicating controllers and level modifiers with new field equipment tied into the Foxboro DCS.

Adding Redundant Guided Wave Radar level monitoring to Feedwater Heaters.

Upgrading existing level modifiers to Fisher DVC6200 positioners o Adding new Foxboro field cabinets to integrate the new input output (I/O) for the FWH DCS components o Integrate the FWH Pressure and Flow Monitoring into DCS HMI Displays Common FWH/HDT Scope o Adding 15KVA (480:208/120V) transformers to power the Field Control Processors (FCPs) in cabinet 2-R-197 in the Auxiliary Instrument Room (AIR)

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Enclosure and the new Foxboro I/O cabinets (2- L-986, 2-L-987, and 2-L-988) in the Turbine Building.

o Power the new and existing Foxboro Virtualization cabinet 2-R-197 power supplies from Unit 2 15kVA transformers (which include power to Group 19 -

Turbine Electrohydraulic Control (EHC) HMI Interface.

The following components for the HDTs and FWHs are impacted and considered as SPVs:

Level Indicating Controller (LIC), Positioner, Control Air Pressure Regulator, Level Gage Float Chamber (LG), Tubing/ Sense Lines, Level I/P Transducer (LM), Level Switch (LS), Power Supply, Relay (Power Transfer), and Fuse.

Field instrumentation could potentially fail and cause the loss of indication of that instrument in the DCS. The loss of these devices, and the impact of this loss, are discussed in the Single Point Failure (SPF) Analysis SPF-WBN-2019- 0003, Rev. 000, "Watts Bar, Unit 2, FWH/HDT Level & DCS Upgrade Single Point Failure Analysis, Review, and Acceptance."

This DC eliminates/hardens the SPV(s) associated with Level Indicating Controllers (LIC), Level I/P Transducer (LM), and Level Switches (LS) through:

The functionality of the LIC is now being provided by a combination of level transmitters, DCS, and the DVC Positioners. The use of multiple (triple redundant) transmitters as input to level switches (for pump protection) and Proportional Integral Derivative (PID) controllers in DCS removes single failure vulnerabilities associated with the LICs and improves the reliability and robustness of the control system.

The function of the removed level switches is performed by the DCS software and hardware. Outputs to each pump interlock are now separate outputs, segregated to separate FBMs.

The installation of the DVCs for the main and bypass valves for HDTs and FWHs provides significant hardening for each control loop. All analog DCS control is provided through redundant output channels.

o The FWHs have pneumatic level controllers which are comprised of several component SPVs including the bellows, input connections, output connections, supply air, internal tubing, nozzles, mechanical linkages, and other components.

The replacement of the pneumatic level controllers and positioners with DVCs does not eliminate the SPVs; however, DVCs have fewer components and are more reliable.

o HDT-3 and HDT-7 currently use LMs with pneumatic positioners to control the level control valves from DCS. These will be replaced with DVCs for system hardening and to eliminate Unit discrepancies. In addition, HDT-3 Bypass LCVs will be independently controlled by DCS.

o All AC and DC power to the DCS components and instrumentation is redundant, eliminating a SPV failure mechanism associated with power to existing instrumentation.

Local Foxboro I/A DCS remote I/O field cabinets are installed in the Turbine Building on EL.

729.0' and additional modules are added to existing cabinets on 708.0 feet. The new cabinets will contain the necessary field bus modules (FBMs) and field communications modules (FCMs) for accommodating the process level control of FWHs and will interface with the FCPs and DCS Server Platform located in the Auxiliary Instrument Room (AIR). Existing FBMs and baseplates E31 OF E45 WBL-23-045

Enclosure within I/O cabinets 2-L-900 and 2-L-901 will interface with the upgraded HDT-3 and HDT-7 FCPs. As appropriate, the system architecture will be configured to address SPV design issues ensuring that a loss of a single controller, transmitter, power supply, etc. will not result in a system failure, which could cause a plant trip or runback.

The existing Foxboro DCS consists of two operator stations (thin clients) in the Unit 2 Main Control Room (MCR), one engineering station in the Unit 2 AIR, and multiple field 1/0 cabinets.

The existing workstation computers supporting the operator and engineering stations are updated to include the changed HDT displays and new FWH displays. A Human Factors Evaluation is performed as a part of WBN-19-352 to evaluate the acceptability of the displays.

Monitoring and control of the HDTs and FWHs are performed in the Main Control Room (MCR) via the operator stations. The Engineering Workstation is in the AIR for FCP programming and configuration.

This modification expands the functionality of the existing displays, however, the method of using operator software displays for level control, using a triple redundant transmitter architecture, is already in use in other systems controlled through DCS. Alarms, status displays, control interfaces, and other interfaces should therefore be familiar to Operators, Engineers, and Maintenance, minimizing the possibility of failure due to human error.

Summary of Evaluation The digital DCS system replaces existing analog controls in the existing balance of plant (BOP) systems with digital control systems and reduces many single point failure vulnerabilities with reliability improvements. The impact of SPVs are reduced by eliminating single components that can fail and cause a transient of greater than 5 percent in the system. The remaining SPVs are hardened by eliminating the pneumatic positioners and level indicating controllers, thereby reducing the number of components that can cause an SPV. In addition, the impact of the SPVs remaining are reduced by detection and annunciation of faults and failures in the DCS and DVCs.

The new system provides redundant inputs, redundant processors, networks, power supplies with backup power, etc. The new system is designated as "Quality Related" and is designed to meet Quality Related requirements; the reliability of new system is superior to the old analog system. The modification does not negatively impact any SSC that is important to safety nor does it impact the consequences or the frequency of their occurrence. The upgraded DCS does not cause a new type of malfunction or accident to be created. The upgraded DCS reduces the likelihood of failures and their consequences by providing more reliable and redundant control system. In addition, this modification provides the capability to reduce manual operator actions, thereby, allowing greater opportunity for assessment, monitoring and response.

The upgrade to DCS and replacement systems results in overall improvement in the plant and the ability to function with individual devices out of service as:

DCS provides for use of additional input signals for control. The DCS will continue to maintain function with the loss of a single input for controls with multiple inputs. In the case of a single input, the last good value prior to the failure will be used. The DCS will provide an alarm on the DCS VDU for loss of an input.

The DCS is powered from redundant power sources, thus in the loss of any single power supply or power source the DCS will continue to maintain control in normal operation.

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Enclosure The signal output to plant control devices, such as valves, use redundant FBMs such that should one FBM fail the other FBM maintains the control of the device.

Important functions are separated on a pair of DCS processors for segregation.

As the primary and secondary sources for the FWH equipment are both powered from U2 MOV boards (supplied by 6.9kV Unit Boards), there is maximum of a six-cycle period where the DCS equipment will not have power due to the buses transferring to a common board on a unit trip. It is likely that the voltage will not drop enough in this period for the equipment to lose power, however, in this event the power is lost, all FWH strings' level control valves will isolate and close. As feedwater heater flow is not required after a unit trip and as the buses are manually transferred (one board at a time) to normal power on system startup, no UFSAR function is impacted by this failure. The HDT equipment is powered from the TSC Board 2 and Vital Instrument Power Boards. Therefore, losing power to both DCS power sources would require dual failures of equipment, which are not considered in the failure analysis.

The proposed modification does not increase the frequency or likelihood of accidents or malfunctions or create a new type of accident. A hardware-related Common Cause Failure (CCF) conclusion of unlikely (Not credible) and a software-related CCF conclusion of not unlikely (Credible) were determined from the design change's Single Point Failure (SPF)

Analysis and other supporting documentation. The likelihood of these failures were qualitatively determined to result in no more than a minimal increase in the frequency of occurrence. In addition, the system was segmented so as to prevent software and hardware CCFs for critical functions. The segmentation strategy was implemented using the previous evaluation performed for the Unit 2 DCS calculation DCSSEGMENT. Therefore, the credible software CCFs do not introduce a new failure mode failure or cause more than a minimal increase in the failures already assumed in the UFSAR.

As a result of this evaluation, it is concluded that this activity does not meet any of the criteria of 10CFR50.59(c)(2), and therefore obtaining prior NRC approval is not required to implement this activity.

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Enclosure WBN-19-759 Rev. 3, Evaluation Rev. 1 Activity Description This activity, design change WBN-19-759, upgrades the existing Westinghouse ICCM86 Watts Bar Unit 1 Inadequate Core Cooling Monitoring (ICCM) Post Accident Monitoring System (PAMS). ICCM PAMS is a safety-related digital system that provides Post Accident Monitoring (PAM) parameters to the Main Control Room (MCR) operators. The existing Westinghouse ICCM86 ICCM PAMS system is obsolete.

This activity installs a Westinghouse Common Q ICCM PAMS system in Unit 1, similar to the Common Q ICCM PAMS system installed in Unit 2. This activity also replaces the Unit 2 subcooling margin monitor displays on MCR panel 2-M-4.

PAMS performs multiple functions, including:

Inadequate Core Cooling Monitoring (ICCM)

Reactor Vessel Water Level Indicating System (RVLIS)

Display of Core Exit Thermocouple (CET) values Subcooled Margin Monitoring (SMM) and display Assists in detecting the presence of a gas bubble or void in the reactor vessel, as well as approach to inadequate core cooling This activity re-uses the existing PAMS inputs and input cables, including the existing incore thermocouples, Reactor Coolant System (RCS) wide range pressure transmitters, and Eagle 21 inputs.

This activity was screened in for evaluation due to the use of common software in redundant trains of safety-related equipment.

Summary of Evaluation This activity replaces the PAMS input/output hardware, the processor, the operator interface, and the software with new and more reliable Common Q equipment. This activity does not affect or change the PAMS inputs. Both the new and existing ICCM PAMS systems are digital systems.

PAMS is a monitoring system; PAMS does not control any plant SSC, or initiate any plant action. PAMS provides Technical Specification 3.3.3 required PAM information used by operators to initiate manual actions. PAMS is described in UFSAR Section 7.5.1.8.

The new PAMS equipment is safety-related. The new PAMS consists of two independent and redundant trains of hardware and software.

Based on the results of this 10CFR50.59 Evaluation, installation of the new Common Q PAMS for Watts Bar Unit 1 does not require notification to or permission from the Nuclear Regulatory Commission prior to implementation.

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Enclosure WBN-19-760, Evaluation Rev. 0 Activity Description In order to maintain the required water level in each steam generator, the Auxiliary Feedwater System currently operates two parallel modulating level control valves (LCVs) per steam generator in both Units 1 and 2. Currently each steam generator has both a 4-inch and a parallel 2-inch LCV, (which are normally closed) between each Motor-Driven Auxiliary Feedwater (MDAFW) pump and the respective steam generator. The 2-inch LCV fails closed and is designed for extended operation at low flows and high pressure drops. The 4-inch LCV fails open and is utilized when flow demand increases beyond the capabilities of the 2-inch LCV.

Maintaining 2 separate valves to perform this function imposes a significant maintenance cost to WBN. This modification will replace the parallel 2 and 4-inch valves with a single 4-inch Control Components Incorporated (CCI) DRAG Valve which fails open. The parallel piping currently used for the 2 LCV will be capped. The DRAG valve is specifically designed to provide multi-path, multi-stage velocity control technology which prevents cavitation, erosion, noise, and vibration, thus removing the need for 2 separate LCVs.

Existing 4 inch LCVs utilize a spring to fail open. The replacement valves will fail open utilizing an air accumulator.

The control air to the new LCVs will now be vented to ensure that all valves fail open if required meet an Appendix R shutdown requirement. New 1/2 Inch vent valves and their associated FCVs are listed below.

New vent valves and associated LCVs:

WBN-1-VTV-032-3745A-B, Control Air Supply to 1-LCV-3-148 Vent Valve WBN-1-VTV-032-3747A-B, Control Air Supply to 1-LCV-3-171 Vent Valve WBN-1-VTV-032-3761A-A, Control Air Supply to 1-LCV-3-156 Vent Valve WBN-1-VTV-032-3765A-A, Control Air Supply to 1-LCV-3-164 Vent Valve WBN-2-VTV-032-3787A-B, Control Air Supply to 2-LCV-3-148 Vent Valve WBN-2-VTV-032-3789A-B, Control Air Supply to 2-LCV-3-171 Vent Valve WBN-2-VTV-032-3751A-A, Control Air Supply to 2-LCV-3-156 Vent Valve WBN-2-VTV-032-3753A-A, Control Air Supply to 2-LCV-3-164 Vent Valve Summary of Evaluation Replacement of the 2 inch and 4 LCVs with a 4 inch DRAG valve has been evaluated due to deletion of the 2 inch bypass LCV which was a deletion of an automatic design function and the effects of loss of flow through the now deleted and capped 2 bypass piping. The 2 inch LCV supported operation of the AFW system by operating at low flow and high-pressure ranges which would cause cavitation in the 4 inch valves. The replacement DRAG valves are designed to operate at reduced flow and high pressures without cavitating. The 50.59 Evaluation has demonstrated that the proposed activity may be implemented without prior NRC approval.

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Enclosure WBN-21-022 Revision 4, Evaluation Rev. 1 Activity Description The proposed activity implements a reduction in the reactor vessel average temperature

("Tavg") for WBN Unit 2 with the Original Steam Generators (OSGs). WBN Unit 2 is currently operating at reduced power and reduced reactor coolant system (RCS) temperatures in operating Cycle 4 due to concerns with indications in the OSGs. TVA plans to return to full power following a mid-cycle outage in the Fall of 2021, and to ensure the OSGs remain viable through the end of Cycle 4, TVA may elect to operate at a reduced vessel Tavg.

TVA plans to implement a Tavg operating window of 581.2 degrees F to 588.2 degrees F (i.e., 7 degrees F) at Unit 2 starting in in the mid-Cycle 4 outage. This will allow Unit 2 to operate at 100 percent power following the mid-cycle outage with a reduced Tavg of up to 7 degrees F, and also provides more operating flexibility while satisfying the current Cycle 4 operational requirements. At the end of Cycle 4, the Unit 2 OSGs will be replaced.

The proposed activity also evaluates both WBN Units 1 & 2 with Replacement Steam Generators (RSGs) to operate at a reduced Tavg of up to 7 degrees F, creating a Tavg window of operation between 581.2 degrees F and 588.2 degrees F. A Tavg operating window will allow WBN to select a full power Tavg within the 7 degrees F operating band at the start of a fuel cycle. Operating both units at a reduced Tavg will extend the qualified life of the RSGs. In addition, for operation with RSGs, an evaluation is performed to support a reactor coolant system (RCS) Tavg and power coastdown operating strategy at end-of-cycle (EOC) for both units.

Summary of Evaluation Westinghouse evaluations performed in support of the Tavg Reduction Program with Unit 2 Original Steam Generators (OSGs) and Units 1 & 2 Replacement Steam Generators (RSGs) identified the following for review by performing a 10 CFR 50.59 Evaluation: 1) Changes to the NSSS control system settings were established for the Tavg reduction. For the 50 percent load rejection, over 5 percent margin is gained to the Over Temperature Delta Temperature (OTT) reactor trip system (RTS) function by implementing the revised loss of load controller settings.

The margin to trip analyses were rerun to address the changes to the control systems settings;

2) To address the Tavg window, the following non-LOCA events were reanalyzed: Loss of Normal Feedwater, Loss of Offsite Power to the Station Auxiliaries, Inadvertent Operation of the Emergency Core Cooling System, and Chemical and Volume Control System Malfunction; and,

3) The computer program ANSYS was used for analysis of the Reactor Equipment System Model (RESM), which is an alternate method of evaluation that replaces the method of evaluation described in the UFSAR (i.e., WECAN).

In addition, for Units 1 & 2 with RSGs, an evaluation was performed to support an end-of-cycle (EOC) reactor coolant system (RCS) Tavg and power coastdown operating strategy for both units.

Also, for Unit 2 with RSGs, the steam generator tube rupture (SGTR) margin to overfill (MTO) analysis was performed at a high Tavg of 588.2 degrees F and a low Tavg of 581.2 degrees F, and the need to credit a reduced time critical operator action time for isolating auxiliary feedwater flow to obtain acceptable results in the margin to overfill the steam generators was identified. The EOC coastdown evaluation and the need to credit a reduced time critical E36 OF E45 WBL-23-045

Enclosure operator action time following a Unit 2 SGTR event also required review by performing a 10 CFR 50.59 Evaluation.

These activities were evaluated and it was concluded that the operability and margin to trip analyses show that, with the necessary NSSS control system changes, the margins for the impacted transients are acceptable for Unit 2 with OSGs and Units 1 & 2 with RSGs. The reanalysis of the four non-LOCA transients for the Tavg window for Unit 2 with OSGs and Units 1 & 2 with RSGs, and for the EOC coastdown conditions for Units 1 & 2 with RSGs, has demonstrated that the appropriate criteria continue to be met for these analyses. Also, the implementation of the revised operator action time for the Unit 2 SGTR event with RSGs is demonstrated to be acceptable. These activities do not increase the frequency of accidents or likelihood of malfunctions, does not increase the consequences of an accident or malfunction, or create a new type of accident or malfunction. Also, these activities will not result in a design basis limit for a fission product barrier as described in the UFSAR being exceeded or altered. In addition, the use of ANSYS for analysis of the RESM does not involve a departure from a method of evaluation described in the UFSAR used in establishing the design bases or in the safety analyses.

Results Implement the activity per plant procedures without obtaining a License Amendment.

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Enclosure WBN-21-047, Evaluation Rev. 0 Activity Description NRC Information Notice (IN) 2011-21 identified an issue with the modeling of thermal conductivity in the Emergency Core Cooling System (ECCS) evaluation models due to degradation of fuel conductivity from irradiation effects. Design Change Package (DCP) WBN-21-047 is a document only change package for implementation of NRC-approved WCAP-17642-P-A, Revision 1, "Westinghouse Performance Analysis and Design Model (PAD5),"

which properly accounts for the effects of fuel conductivity degradation with burnup.

As part of the PAD5 fuel performance code adoption, Westinghouse re-performed the non-LOCA safety analysis described in Chapter 15 of the Watts Bar Units 1 and 2 Updated Final Safety Analysis Report (UFSAR) for when both Watts Bar units are operating with Westinghouse Model 68AXP replacement steam generators (RSGs). Eight UFSAR events were reanalyzed for PAD5 implementation. This activity involves replacing an UFSAR described evaluation methodology that is used in the safety analyses and therefore requires a 50.59 Evaluation.

Additionally, as a result of PAD5 implementation, the reanalysis of the Steam Generator Tube Rupture (SGTR) event resulted in new mass releases. The new mass releases resulted in a revision to the dose analysis for control room operator and offsite doses due to a SGTR. The revision to the analysis and resulting increase in dose consequences therefore require a 50.59 Evaluation.

Summary of Evaluation The adoption of PAD5 does not constitute a departure from a method of evaluation described in the UFSAR used in establishing the design bases or in the safety analyses because it replaces a method of evaluation with a method of evaluation that is approved by the NRC for the intended application. The associated NRC Safety Evaluation is included as part of WCAP-17642-P-A, Revision 1, "Westinghouse Performance Analysis and Design Model (PAD5)."

Section 4.0 of the Safety Evaluation outlines acceptable cladding and fuel materials, fuel fabrication specifications, burnable absorbers, and reactor characteristics that must be satisfied in order to use the PAD5 models and methodology. It was confirmed that the limitations and conditions outlined in Section 4.0 of the Safety Evaluation are met for the PAD5 implementation at Watts Bar Units 1 and 2. Therefore, the PAD5 computer code is considered to be approved for use in the Watts Bar Unit 2 licensing basis and prior NRC approval is not required for this change.

Implementation of PAD5 does not affect the frequency of occurrence of an accident previously evaluated in the UFSAR, the likelihood of occurrence of a malfunction of an SSC important to safety previously evaluated in the UFSAR, or the consequences of a malfunction of an SSC important to safety previously evaluated in the UFSAR. The change does not create a possibility for an accident of a different type than any previously evaluated in the UFSAR or create a possibility for a malfunction of an SSC important to safety with a different result than any previously evaluated in UFSAR. Design basis limits for a fission product barrier, as described in the UFSAR, are not exceeded or altered. The resulting increases in operator and offsite doses due to the increased mass releases for a Steam Generator Tube Rupture event meet the criteria in NEI 96-07, Revision 1, for a no more than minimal increase, and the associated regulatory limits are not exceeded. Therefore, the proposed activity does not result in more than E38 OF E45 WBL-23-045

Enclosure a minimal increase in the consequences of an accident previously evaluated in the UFSAR and prior NRC approval is not required for this change.

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Enclosure WBN-22-015, Evaluation Rev. 0 Activity Description WBN-22-015, Compensatory measures for Loss of Shutdown Board Room (SDBR) HVAC cooling, is a document only change to enable the plant to implement compensatory measures in the event of a total loss of SDBR HVAC cooling. The area served by the SDBR HVAC system, located on elevation 757.0' of the Auxiliary Building is classified as Mild (Ref. 1). A Mild Zone is defined as a building zone where (1) the temperature, pressure, and relative humidity resulting from the direct effects of a DBA (such as temperature rise due to steam release) are no more severe than those which would occur during an abnormal plant operational condition; (2) the temperature will not exceed 130°F due to the indirect effects of OBA (such as increased heat loads from electrical equipment).

The scope of the ECP is to create compensatory measures to maintain temperatures within the limits for the area served by the SDBR HVAC system upon loss of cooling. The compensatory measures required for the SDBR area will prevent the temperature from exceeding 130°F during accident conditions. The site would also utilize the compensatory measures during normal operation upon a total loss of SDBR HVAC cooling. During normal operation of the plant, temperatures within the SDBR area are maintained by the SDBR HVAC system within the range specified (normal limit no greater than 85°F) in the Technical Requirements Manual, Section 3.7.5 (Ref. 3). The abnormal limit temperature specified within the TRM for the SDBR area is no greater than 104°F. Therefore, the site's goal is to maintain temperature below 104°F upon a total loss of SDBR HVAC cooling for as long as possible by utilizing the compensatory measures added by WBN 015.

Prior to the development of this ECP, Train B SDBR chiller (0-CHR-031-49/2) was permanently removed from service due to a malfunctioning compressor. Therefore, only Train A chiller (0-CHR-031-36/2) remains available for service. No credit can be taken for the temporary chiller located out in the yard during accident conditions, which supplies chilled water to the SDBR air handling units D-B and C-B. Therefore, compensatory measures are required for the plant to utilize in the event of a total loss of SDBR HVAC cooling.

Due to total loss of SDBR HVAC Cooling system, the temperature of the areas served by the HVAC system elevates following a concurrent LOCA event (Ref. 4 & 5). The compensatory measures required to prevent the area temperature from exceeding 130°F include deployment and operation of 10 portable fans (6,000 CFM each) to circulate air on elevations 757.0' and 772.0'. In addition, 19 doors located throughout both elevations should be opened to allow air to circulate freely between the 480V Boardrooms (El. 772.0') and Shutdown Boardrooms (El.

757.0'). Cooling is still maintained on the 772.0' elevation. Cooling on the 772.0' elevation is performed by the 480V Board Room HVAC units. Therefore, this change utilizes the 480V Board Room cooling system as part of the compensatory measures along with the fans to circulate air from the 772.0' elevation to the 757.0' elevation. Class 1E 480V power sources (designated compartments within Units 1 and 2 C&A Vent Boards, El. 757.0') will be used to power the fans to circulate air on both elevations. The compartments within the designated C&A vent boards are not used for any other purpose than for powering the fans as part of the compensatory measures.

As part of this ECP, TVA procedure 0-SOl-30.07 (Operating Instruction for Shutdown Board Rooms HVAC El. 757 and 772) requires an update to address actions required for deployment and operation of the equipment associated with these compensatory measures. Due to the E40 OF E45 WBL-23-045

Enclosure changes to the procedure to add manual actions in place of automatic actions, this 10 CFR 50.59 has screened in.

Summary of Evaluation Accidents evaluated in Chapter 15 of the UFSAR for Watts Bar will not increase more than minimally as a result of these compensatory measures, due to considerations taken in the added actions to 0-SOI-30.07 for deployment and operation of the compensatory measures.

Malfunctions to SSCs important to safety will not see more than a minimal increase due to the added actions. Radiological consequences from an accident or from malfunctions of SSCs important to safety will not increase more than minimally as a result of added actions. Barriers protecting against radiological doses will not be impacted by the added actions. The design considerations included as part of these compensatory measures, in conjunction with the plant's existing design considerations, preclude any new accident types or malfunctions to SSCs important to safety from being introduced by the added actions. No fission product barriers are impacted by added actions and no methods of evaluation are impacted by these added actions.

Results Implement the activity per plant procedures without obtaining a License Amendment.

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Enclosure WBN-22-060 Rev. 0, Evaluation Rev. 0 Activity Description The proposed change would increase the testing and maintenance intervals associated with portions of the Watts Bar Nuclear Station main turbine overspeed protection system (OPS).

Specifically, the frequencies associated with testing the Trip Block solenoid valves (TBSVs) and stroke testing and maintenance overhauls of the main turbine steam admission valves, i.e.,

governor valves (GVs), throttle valves (TVs), reheat stop valves (RSVs) and intercept valves (IVs) will be extended as part of the proposed change. These valves are key elements of the turbine OPS due to their function and use in terminating steam flow to the high pressure (HP) turbine and low pressure (LP) turbines. Prompt and effective termination of steam flow to the main turbine during a load separation event leading to a turbine overspeed condition limits the potential that turbine missiles can be generated as a result of turbine rotor failure.

The above mentioned TBSV and turbine valve tests are currently performed at Watts Bar Units 1 and 2 every six months. The maintenance overhauls of the HP valves (GVs and TVs) and LP valves (RSVs and IVs) are currently conducted every three years and six years, respectively.

These frequencies are an input to the current turbine missile probability analysis as they affect the component reliabilities and the overall system failure probability that correlates to the missile generation probability. The method used to calculate turbine missile probability is described in UFSAR Section 3.5.1.3.5. The change to an input parameter (e.g., valve reliability as a function of test and maintenance intervals) for a method described in the UFSAR is evaluated herein, since extending the interval between valve stroke test or maintenance overhaul increases the overall turbine missile generation probability.

A review of turbine valve component failure modes and effects, a review of turbine Operations Operating Experience (OPS OE), as well as a quantitative analysis of the impact from various test and maintenance interval extension cases (eleven scenarios) on turbine missile probability P1 was documented in Reference [1] *. These cases are shown to meet the NRC turbine missile generation acceptance criteria; Additionally, Reference [1] provides other provisions and recommendations on how Watts Bar should exercise increasing such test intervals. The proposed change in this 50.59 Evaluation is inclusive of any of those eleven test and maintenance interval extension cases (see Table 5-1 in Reference [1]) after enacting the respective recommendations. As shown in Table 5-1 of Reference [1], the turbine missile probability P1 would change from the current baseline intervals (2.22 E-6 and 3.27 E-7 per year for Unit 1 and Unit 2, respectively) to a maximum of 5.70 E-6 and 3.80 E-6 per year for Unit 1 and Unit 2, respectively, for the most extreme test and maintenance interval extension case (Case 11). These maximum values are still less than the NRC acceptance limit of 1 E-4 per year per the US NRC Regulatory Guide 1.115 for plants with favorably oriented turbine.

• Reference [1]: MPR Report 2048-0056-RPT-001, "Watts Bar Turbine Valve Maintenance and Test Interval Extension Assessment", Rev. 0.

Summary of Evaluation The proposed change includes decreasing the frequency of stroke tests of the main turbine valves GVs, TVs, RSVs, and IVs, extending test intervals of TBSVs, and decreasing maintenance overhaul frequencies of the steam admission valves. Based on statistical data analysis, reducing the frequency of periodic valve tests does not more than minimally increase E42 OF E45 WBL-23-045

Enclosure the likelihood of valve failure during operation; additionally, the reliability of the turbine overspeed protection system minimally decreases due to reduced testing frequencies. With the implementation of the proposed activity, the evaluation of the change in the reliability of the overspeed protection system (Reference [1]) demonstrated that the turbine missile generation probability will continue to meet the acceptance criterion described in the UFSAR Sections 3.5.1.3.3 and 3.5.1.3.5. The 50.59 evaluation responses to each of the eight criteria are "No".

Therefore, prior NRC approval is not required for implementation of the proposed change.

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Enclosure WBN-TS-22-09, Evaluation Rev. 0 Activity

Description:

This activity is a change to the UFSAR and to the Unit 1 and Unit 2 Tech Spec Bases to document that the Tech Spec Surveillance Requirements (SRs) 3.1.4.1, 3.1.4.2, and 3.1.4.3 may be performed via verification from cycle design values in lieu of a plant measurement provided sufficient margin to the limits exists with a 95 percent confidence level that 95 percent of the measurements would be within the limit. Should sufficient margin not be available for the applicable surveillance, then a measurement shall be obtained. These surveillances pertain to the Beginning of Cycle (BOC) and End of Life (EOL) Moderator Temperature Coefficient (MTC) values. These changes are being made in accordance with PWROG-19014-P, 'Verification Versus Measurement of the Beginning of Cycle Life and End of Cycle Life Moderator Temperature Coefficient." WBN was a participant in this report; as such, the conclusions based upon the statistical data used in the report contain WBN data.

A corresponding wording change to the applicable Core Operating Limits Reports will also be made.

It must be noted that any plant measurement for the MTC is not based upon pure measurements, but involves a mixture of measurements and cycle design predictions.

Examples of necessary predicted design values include the Doppler Temperature Coefficient (DTC), Inverse Boron Worth, Hot Full Power (HFP) Boron Letdown Curve, time dependent Xenon values, Boron Correction Factor, and, in the case of using a reactivity computer, the delayed neutron data for the 6 decay groups.

While both the UFSAR and Technical Specification Bases imply that these surveillance requirements are satisfied by performing measurements, each Tech Spec SR simply states

'Verify MTC is within" the applicable limit. It must be noted that the word "measured" appears once in the second Note for SR 3.1.4.3, but the context for that word usage is that the MTC value at the 300 ppm surveillance window was not within limit. As stated above, the revised Bases require MTC measurement if the MTC value cannot be verified within limit with a 95/95 confidence. In that case, a measurement will be performed at the 300 ppm window, and, if the MTC is not within limit, the measurement shall be repeated every 14 Effective Full Power Days (EFPD) in accordance with SR 3.1.4.3 Note 1. Note 2 explains the conditions that must exist for the repeated measurements at a 14 EFPD frequency to be discontinued.

Additional clarifications are made in the UFSAR. For instance, the UFSAR states that the estimated accuracy of the current analytical methods for the MTC is +/-2 pcm/F. While +/-2 pcm/F will continue to be used within the safety analysis for MTC and will remain as the review criteria for MTC measurements, if required to be performed, the UFSAR clarifies that PWROG-19014-P concludes that an uncertainty of +/-1.5 pcm/F bounds all MTC measurements in the industry with a 95/95 confidence interval.

The verification methodology described in PWROG-19014-P is to apply the bounding MTC measurement uncertainty of 1.5 pcm/F to the predicted MTC value from the cycle design and compare the result to the limit. If the verification is satisfied, the measurement is moot. In the case of the End of Life MTC verification, additional uncertainties are conservatively applied based upon comparisons of measured to predicted values for other reactivity parameters for the applicable fuel cycle. Mis-predictions in Axial Offset, core reactivity, and core burnup can require additional uncertainties to be applied.

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Enclosure Many design basis accidents are affected by the assumed MTC because it affects the core reactivity with a change of the moderator temperature. In order for the safety analysis to conservatively bound all conditions, some accidents assume the MTC is at upper limit (with uncertainty), some accidents assume the MTC is at the lower limit (with uncertainty), and a few accidents assume both, depending upon the parameters being modelled.Section V of this Screening Review Form lists the applicable design basis accidents.

Summary of Evaluation:

This 50.59 Evaluation concludes that this activity does not impact the frequency or likelihood of any assumed accident or malfunction of an SSC because this activity does not affect the initiating mechanisms by which an accident is initiated or any SSC malfunctions. Similarly this activity does not create the possibility of an accident of a different type or a malfunction of an SSC important to safety with a different result than previously evaluated in the UFSAR. This activity does not contain any mechanism by which to negatively impact the operation of any SSC. This 50.59 Evaluation concludes that this activity does not increase the consequences of any assumed accident or malfunction of an SSC because the MTC shall remain within the upper and lower limits assumed in the safety analyses. This 50.59 Evaluation concludes that this activity does not result in a design basis limit for a fission product barrier as described in the UFSAR being exceeded or altered and does not result in a departure from a method of evaluation described in the UFSAR used in establishing the design bases or in the safety analyses.

Based upon the results of this evaluation:

Implement the activity per plant procedures without obtaining a License Amendment.

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