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OF VALIDATION ACTIVITIES This Chapter provides, in punch-list format, a summary of the validation activities to be performed in order to enable confirmation of the results presented herein and submission of the final report for the WBN2 IPEEE implementation.
OF VALIDATION ACTIVITIES This Chapter provides, in punch-list format, a summary of the validation activities to be performed in order to enable confirmation of the results presented herein and submission of the final report for the WBN2 IPEEE implementation.
5.1 SSEL VALIDATION 5.1.1 Upon availability of the WBN2 final design and design drawings as well as Master Equipment List (MEL), verify the WBN2 SSEL equipment ID, description, flow diagram, category, seismic category, and elevation/room against the completed MEL.5.1.2 Obtain TVA Probability Risk Assessment (PRA) Section and Operations reviews of the WBN2 SSEL and resolve comments as required.
5.1 SSEL VALIDATION 5.1.1 Upon availability of the WBN2 final design and design drawings as well as Master Equipment List (MEL), verify the WBN2 SSEL equipment ID, description, flow diagram, category, seismic category, and elevation/room against the completed MEL.5.1.2 Obtain TVA Probability Risk Assessment (PRA) Section and Operations reviews of the WBN2 SSEL and resolve comments as required.
It will be confirmed that the SSEL and success paths are those normally preferred by operations for shutdown following a seismic event.5.1.3 An independent peer review of the WBN2 SSEL development process and results will be performed similar to the independent peer review that was conducted for the WBN1 SSEL.5.2 RELAY REVIEW VALIDATION 5.2.1 Evaluate any changes to any of the design documents utilized in the development of the WBN2 relay review design report.5.2.2 Confirm that the bad actor relay identification encompasses the scope of equipment in the final WBN2 SSEL.5.2.3 Prepare a Control and Power Interface to document any differences between WBN1 and WBN2, including but not limited to the SSEL control and/or power and fail safe circuitry.
It will be confirmed that the SSEL and success paths are those normally preferred by operations for shutdown following a seismic event.5.1.3 An independent peer review of the WBN2 SSEL development process and results will be performed similar to the independent peer review that was conducted for the WBN1 SSEL.5.2 RELAY REVIEW VALIDATION
This Interface will be the source document for further investigation of bad actor relays.5.2.4 An independent peer review of the WBN2 Relay Review will be performed similar to the independent peer review that was conducted for the WBN1 Relay Review.WBN2 IPEEE DESIGN REPORT Page 31 Revision 0 April 27, 2010 5.3 SEISMIC MARGIN ASSESSMENT 5.3.1 Confirm completion of corrective action programs of interest to the seismic margin assessment, in particular the HAAUP, lIP, and ESQ programs.5.3.2 Review the final SSEL and identify any new items or other significant deviations for seismic assessment.
 
====5.2.1 Evaluate====
any changes to any of the design documents utilized in the development of the WBN2 relay review design report.5.2.2 Confirm that the bad actor relay identification encompasses the scope of equipment in the final WBN2 SSEL.5.2.3 Prepare a Control and Power Interface to document any differences between WBN1 and WBN2, including but not limited to the SSEL control and/or power and fail safe circuitry.
This Interface will be the source document for further investigation of bad actor relays.5.2.4 An independent peer review of the WBN2 Relay Review will be performed similar to the independent peer review that was conducted for the WBN1 Relay Review.WBN2 IPEEE DESIGN REPORT Page 31 Revision 0 April 27, 2010 5.3 SEISMIC MARGIN ASSESSMENT
 
====5.3.1 Confirm====
completion of corrective action programs of interest to the seismic margin assessment, in particular the HAAUP, lIP, and ESQ programs.5.3.2 Review the final SSEL and identify any new items or other significant deviations for seismic assessment.
5.3,3 Once the bulk of WBN2 construction work is completed, perform the seismic margins walkdown for the items of equipment that are unique to the WBN2 SSEL.5.3,4 Perform a walk-by for common items that were previously addressed in the WBN1 program as deemed necessary due to possible significant changes from WBN2 construction completion modifications.
5.3,3 Once the bulk of WBN2 construction work is completed, perform the seismic margins walkdown for the items of equipment that are unique to the WBN2 SSEL.5.3,4 Perform a walk-by for common items that were previously addressed in the WBN1 program as deemed necessary due to possible significant changes from WBN2 construction completion modifications.
5.3,5 Perform specific verification walkdown screening evaluations for the Centrifugal Charging Pump (CCP 2A& 2B) & Residual Heat Removal (RHR)Pump Room Coolers and Lower Compartment Cooler Fans.5.3.6 Perform an independent walkdown of the WBN2 RWST to assess if there is any new seismic vulnerability in relation to the WBN1 RWST seismic margin evaluation.
5.3,5 Perform specific verification walkdown screening evaluations for the Centrifugal Charging Pump (CCP 2A& 2B) & Residual Heat Removal (RHR)Pump Room Coolers and Lower Compartment Cooler Fans.5.3.6 Perform an independent walkdown of the WBN2 RWST to assess if there is any new seismic vulnerability in relation to the WBN1 RWST seismic margin evaluation.
5.3.7 Confirm the seismic margin of WBN2 masonry walls by comparison to WBN1 masonry walls.5.3.8 Confirm that the same detail as used in WBN1 is used for the WBN2 ice basket lower seal.5.3.9 An independent peer review of the WBN2 SMA will be performed similar to the independent peer review that was conducted for the WBN1 SMA, in accordance with the peer review requirements as described in the EPRI NP-6041-SL (Ref. 3).5.4 OTHER EXTERNAL EVENTS ASSESSMENT 5.4.1 Perform a general site walkdown to identify potential objects that could be picked up by a tornado that may not be enveloped by the WBN1 evaluations.
 
====5.3.7 Confirm====
the seismic margin of WBN2 masonry walls by comparison to WBN1 masonry walls.5.3.8 Confirm that the same detail as used in WBN1 is used for the WBN2 ice basket lower seal.5.3.9 An independent peer review of the WBN2 SMA will be performed similar to the independent peer review that was conducted for the WBN1 SMA, in accordance with the peer review requirements as described in the EPRI NP-6041-SL (Ref. 3).5.4 OTHER EXTERNAL EVENTS ASSESSMENT
 
====5.4.1 Perform====
a general site walkdown to identify potential objects that could be picked up by a tornado that may not be enveloped by the WBN1 evaluations.
WBN2 IPEEE DESIGN REPORT Page 32 Revision 0 April 27, 2010 5.4.2 Confirm that the metal-wall panels for the Boric Acid Mixing Building (BAMB)are bounded by the WBN1 IPEEE tornado missile assumptions based on the Turbine Building wall.5.5 FIVE VALIDATION 5.5.1 The Unit 2 population of rooms with Appendix R Safe Shutdown (SSD)Equipment will be reviewed to ensure that no safe shutdown components or plant trip initiators have been added to the scope. If any of these are discovered, they will be evaluated via the FIVE process. A representative population of rooms will be reviewed to ensure that each room's configuration, barrier ratings, room use, etc. has not changed. Based on the results of this review, rooms will be reanalyzed as necessary and changes incorporated into the analysis (Phase I).5.5.2 A representative population of Unit 2 rooms will be reviewed to verify that there have been no significant changes in the room ignition frequencies which would result in a less conservative analysis result. New walkdowns will be performed and incorporated into the analysis as necessary (Phase II, Step 1).5.5.3 The "as built" equipment and location data for the Unit 2 Appendix R SSD equipment and safety injection/recirculation equipment will be reviewed and incorporated into the Plant Probabilistic Risk Assessment (PRA) as necessary to update the analysis.
WBN2 IPEEE DESIGN REPORT Page 32 Revision 0 April 27, 2010 5.4.2 Confirm that the metal-wall panels for the Boric Acid Mixing Building (BAMB)are bounded by the WBN1 IPEEE tornado missile assumptions based on the Turbine Building wall.5.5 FIVE VALIDATION 5.5.1 The Unit 2 population of rooms with Appendix R Safe Shutdown (SSD)Equipment will be reviewed to ensure that no safe shutdown components or plant trip initiators have been added to the scope. If any of these are discovered, they will be evaluated via the FIVE process. A representative population of rooms will be reviewed to ensure that each room's configuration, barrier ratings, room use, etc. has not changed. Based on the results of this review, rooms will be reanalyzed as necessary and changes incorporated into the analysis (Phase I).5.5.2 A representative population of Unit 2 rooms will be reviewed to verify that there have been no significant changes in the room ignition frequencies which would result in a less conservative analysis result. New walkdowns will be performed and incorporated into the analysis as necessary (Phase II, Step 1).5.5.3 The "as built" equipment and location data for the Unit 2 Appendix R SSD equipment and safety injection/recirculation equipment will be reviewed and incorporated into the Plant Probabilistic Risk Assessment (PRA) as necessary to update the analysis.
Manual actions credited in the analysis will be confirmed.
Manual actions credited in the analysis will be confirmed.
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===1.2 Methodology===
===1.2 Methodology===
1.3 Work Scope Statement 1.4 Safe Stable Shutdown States 2.0 SAFE SHUTDOWN PATHS 13 2.1 Assumptions 2.2 Resources Used 2.3 Safe Shutdown Path Descriptions 2.3.1 Intact RCS 2.3.2 Small LOCA 2.4 Dependency Matrices 3.0 SYSTEM DESCRIPTIONS 28 3.1 Main Steam 3.2 Main Feedwater 3.3 Auxiliary Feedwater 3.4 Chemical and Volume Control 3.5 Safety Injection 3.6 Essential Raw Cooling Water 3.7 Component Cooling System 3.8 Residual Heat Removal 3.9 Reactor Protection System 3.10 Electrical Power System 3.11 Miscellaneous Systems 4.0 SAFE SHUTDOWN EQUIPMENT LIST 64 4.1 Approach 4.2 Definition of Table Entries 4.3 Safe Shutdown Equipment List 5.0 PEER REVIEW 72 WBN2 IPEEE DESIGN REPORT Page 39 ATTACHMENT 1
1.3 Work Scope Statement 1.4 Safe Stable Shutdown States 2.0 SAFE SHUTDOWN PATHS 13 2.1 Assumptions
 
===2.2 Resources===
 
Used 2.3 Safe Shutdown Path Descriptions
 
====2.3.1 Intact====
RCS 2.3.2 Small LOCA 2.4 Dependency Matrices 3.0 SYSTEM DESCRIPTIONS 28 3.1 Main Steam 3.2 Main Feedwater 3.3 Auxiliary Feedwater 3.4 Chemical and Volume Control 3.5 Safety Injection 3.6 Essential Raw Cooling Water 3.7 Component Cooling System 3.8 Residual Heat Removal 3.9 Reactor Protection System 3.10 Electrical Power System 3.11 Miscellaneous Systems 4.0 SAFE SHUTDOWN EQUIPMENT LIST 64 4.1 Approach 4.2 Definition of Table Entries 4.3 Safe Shutdown Equipment List 5.0 PEER REVIEW 72 WBN2 IPEEE DESIGN REPORT Page 39 ATTACHMENT 1
Watts Bar Nuclear Plant Unit 2 IPEEE Seismic Report No: WBNIPEEE-003 Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List Revision 0 Page: 4 TABLE OF CONTENTS (cont.)Section Title Page 6.0 7.0  
Watts Bar Nuclear Plant Unit 2 IPEEE Seismic Report No: WBNIPEEE-003 Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List Revision 0 Page: 4 TABLE OF CONTENTS (cont.)Section Title Page 6.0 7.0  


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Potential for human error is minimized by selecting Safe Shutdown Paths that require either no operator actions (such as for a loss of offsite power with AFW available), or a minimum of operator actions. In all cases operator actions required to achieve the Safe Shutdown Path are those normally included in the operator training program and trained on by operators; for example, performing bleed and feed operations in the event of a loss of AFW. This approach allows human error failures to be screened out using the guidelines identified in Section 2.4.3 of Volume 2 of NUREG/CR-4826 (Reference 4).2.2 Resources Used in the Safe Shutdown Path Development.
Potential for human error is minimized by selecting Safe Shutdown Paths that require either no operator actions (such as for a loss of offsite power with AFW available), or a minimum of operator actions. In all cases operator actions required to achieve the Safe Shutdown Path are those normally included in the operator training program and trained on by operators; for example, performing bleed and feed operations in the event of a loss of AFW. This approach allows human error failures to be screened out using the guidelines identified in Section 2.4.3 of Volume 2 of NUREG/CR-4826 (Reference 4).2.2 Resources Used in the Safe Shutdown Path Development.
Resources listed in table 2.1 were reviewed and used, as determined applicable, in the development of the safe shutdown paths and critical equipment list developed for the WBN Seismic Margins assessment.
Resources listed in table 2.1 were reviewed and used, as determined applicable, in the development of the safe shutdown paths and critical equipment list developed for the WBN Seismic Margins assessment.
2.3 Shutdown Path Descriptions Consistent with the EPRI methodology, two scenarios were considered in developing the Safe Shutdown Path for the WBN seismic margins assessment; Scenario 1 A "No-LOCA" scenario, which assumes the RCS remains intact throughout the 72 hour time frame of interest, and, Scenario 2 A "Small LOCA" scenario, which assumes the breach in the primary system, occurs at the initiation of the SME.The reason for including small LOCA Safe Shutdown Path in the WBN SME is that it was judged that the effort necessary to demonstrate the availability of a source of makeup water to the RCS would be considerably less than the effort required to demonstrate that the integrity of the reactor coolant pump (RCP) seals and other small instrument and sensing lines would be maintained.
 
===2.3 Shutdown===
Path Descriptions Consistent with the EPRI methodology, two scenarios were considered in developing the Safe Shutdown Path for the WBN seismic margins assessment; Scenario 1 A "No-LOCA" scenario, which assumes the RCS remains intact throughout the 72 hour time frame of interest, and, Scenario 2 A "Small LOCA" scenario, which assumes the breach in the primary system, occurs at the initiation of the SME.The reason for including small LOCA Safe Shutdown Path in the WBN SME is that it was judged that the effort necessary to demonstrate the availability of a source of makeup water to the RCS would be considerably less than the effort required to demonstrate that the integrity of the reactor coolant pump (RCP) seals and other small instrument and sensing lines would be maintained.
Preferred and alternate Safe Shutdown Paths were developed for both scenarios.
Preferred and alternate Safe Shutdown Paths were developed for both scenarios.
They were developed to be consistent with plant operator training and plant procedures.
They were developed to be consistent with plant operator training and plant procedures.
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Watts Bar Nuclear Plant Unit 2 IPEEE Seismic Report No: WBNIPEEE-003 Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List Revision 0 Page: 15 condition be established and maintained for 72 hours, accomplishing the following basic long-term safety functions:
Watts Bar Nuclear Plant Unit 2 IPEEE Seismic Report No: WBNIPEEE-003 Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List Revision 0 Page: 15 condition be established and maintained for 72 hours, accomplishing the following basic long-term safety functions:
: 1) Reactivity Control 2) RCS pressure control 3) RCS inventory control 4) Decay heat removal 5) Containment performance All systems used to accomplish the long terms safety functions are multi-train systems. In addition, equipment components necessary to ensure containment integrity have been identified and included in the SSEL. Check valves which perform a containment isolation function are generically screened and are not included in the SSEL (Ref. EPRI NP 6041-SL, RI, page 3-29).Success Path Logic Diagrams (SPLDs) are used to show both the preferred and alternate Safe Shutdown Paths developed for the WBN SME. The SPLD for Scenario 1 (Intact RCS) is shown in Figure 2-1, and the SPLD for Scenario 2 (Break Size 1 Inch Equivalent Diameter) is shown in Figures 2-2. Both preferred and alternate paths are shown for both scenarios.
: 1) Reactivity Control 2) RCS pressure control 3) RCS inventory control 4) Decay heat removal 5) Containment performance All systems used to accomplish the long terms safety functions are multi-train systems. In addition, equipment components necessary to ensure containment integrity have been identified and included in the SSEL. Check valves which perform a containment isolation function are generically screened and are not included in the SSEL (Ref. EPRI NP 6041-SL, RI, page 3-29).Success Path Logic Diagrams (SPLDs) are used to show both the preferred and alternate Safe Shutdown Paths developed for the WBN SME. The SPLD for Scenario 1 (Intact RCS) is shown in Figure 2-1, and the SPLD for Scenario 2 (Break Size 1 Inch Equivalent Diameter) is shown in Figures 2-2. Both preferred and alternate paths are shown for both scenarios.
2.3.1 Intact RCS Considering Figure 2-1, the SPLD with an intact RCS, core sub criticality is required as is removal of decay heat, RCS pressure control, and RCS inventory control. Core sub criticality is attained by insertion of the control rods into the core. The preferred Safe Shutdown Path is for decay heat removal to be accomplished by supplying auxiliary feedwater to the steam generators and removing heat from the steam generators through the secondary side PORVs. Should decay heat removal via auxiliary feedwater and PORVs either fail or be unavailable, the alternate Safe Shutdown Path is for decay heat removal to be accomplished by means of a bleed and feed operation on the primary side. RCS pressure control is maintained by use of the pressurized PORVs and Safety Valves, and RCS injection via the CCPs, SIPs, and RHR Pumps. RCS inventory control is maintained by RCS injection using CCPs, SIPs, and RHR Pumps and maintaining R.C. Pump seal flow.WBN2 IPEEE DESIGN REPORT Page 51 ATTACHMENT 1
 
====2.3.1 Intact====
RCS Considering Figure 2-1, the SPLD with an intact RCS, core sub criticality is required as is removal of decay heat, RCS pressure control, and RCS inventory control. Core sub criticality is attained by insertion of the control rods into the core. The preferred Safe Shutdown Path is for decay heat removal to be accomplished by supplying auxiliary feedwater to the steam generators and removing heat from the steam generators through the secondary side PORVs. Should decay heat removal via auxiliary feedwater and PORVs either fail or be unavailable, the alternate Safe Shutdown Path is for decay heat removal to be accomplished by means of a bleed and feed operation on the primary side. RCS pressure control is maintained by use of the pressurized PORVs and Safety Valves, and RCS injection via the CCPs, SIPs, and RHR Pumps. RCS inventory control is maintained by RCS injection using CCPs, SIPs, and RHR Pumps and maintaining R.C. Pump seal flow.WBN2 IPEEE DESIGN REPORT Page 51 ATTACHMENT 1
Watts Bar Nuclear Plant Unit 2 IPEEE Seismic Report No: WBNIPEEE-003 Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List Revision 0 Page: 16 Note that the preferred Safe Shutdown Path results in the plant being maintained at Hot Standby (average coolant temperature 3500 F), and that the alternate Safe Shutdown Path results in the plant going to Cold Shutdown.
Watts Bar Nuclear Plant Unit 2 IPEEE Seismic Report No: WBNIPEEE-003 Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List Revision 0 Page: 16 Note that the preferred Safe Shutdown Path results in the plant being maintained at Hot Standby (average coolant temperature 3500 F), and that the alternate Safe Shutdown Path results in the plant going to Cold Shutdown.
This is discussed in some detail in Section 1.4.Insertion of the control rods into the core will occur when the electro-magnetic coils that hold the control rods lose power. This can occur: 1) Automatically upon a signal from the reactor protection system (RPS), 2) Manually by an operator action to trip the plant, or, 3) As a consequence of the loss of offsite power.Loss of offsite power is assumed to occur in this assessment.
This is discussed in some detail in Section 1.4.Insertion of the control rods into the core will occur when the electro-magnetic coils that hold the control rods lose power. This can occur: 1) Automatically upon a signal from the reactor protection system (RPS), 2) Manually by an operator action to trip the plant, or, 3) As a consequence of the loss of offsite power.Loss of offsite power is assumed to occur in this assessment.
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Cold leg recirculation requires drawing water from the sump via the residual heat removal pumps and supplying it to either the charging or safety injection pumps per procedure ES-1.3. The charging and/or safety injection pumps then inject the water back into the RCS cold legs. Cold leg recirculation requires the following actions to be accomplished:
Cold leg recirculation requires drawing water from the sump via the residual heat removal pumps and supplying it to either the charging or safety injection pumps per procedure ES-1.3. The charging and/or safety injection pumps then inject the water back into the RCS cold legs. Cold leg recirculation requires the following actions to be accomplished:
: 1) Establishing the flow path from the sump to the RHR pumps, 2) Starting the RHR pumps if they are not yet running, 3) Establishing the flow path from the RHR pumps to the charging and/or safety injection pumps, and, 4) Isolating the RWST.Procedures direct the operators to switch to hot leg recirculation as early as three hours after the beginning of a LOCA to preclude precipitation of Boron out of solution per procedure ES-1.4. Hot leg recirculation requires switching the RCS injection point from the cold legs to the hot legs. This alternative path is adequate to keep the plant in cold shutdown for 72 hours.For path 1, RCS pressure control is maintained by RCS letdown and / or RCS injection via the CCPs. As a precaution, the Pressurizer PORVs and Safety Valves are listed to insure RCS pressure relief is available for this path. RCS inventory control for path 1 is maintained by RCS injection by the CCPs and by maintaining RCP Seal flow. For path 2, RCS pressure control is maintained by the Pressurizer PORVs and Safety Valves and by RCS injection via the CCPs, SIPs and RHR Pumps. RCS inventory control for path 2 is maintained by the CCPs, SIPs and RHR Pumps and by maintaining RCP Seal flow.See discussion in Section 2.3.2 about containment performance.
: 1) Establishing the flow path from the sump to the RHR pumps, 2) Starting the RHR pumps if they are not yet running, 3) Establishing the flow path from the RHR pumps to the charging and/or safety injection pumps, and, 4) Isolating the RWST.Procedures direct the operators to switch to hot leg recirculation as early as three hours after the beginning of a LOCA to preclude precipitation of Boron out of solution per procedure ES-1.4. Hot leg recirculation requires switching the RCS injection point from the cold legs to the hot legs. This alternative path is adequate to keep the plant in cold shutdown for 72 hours.For path 1, RCS pressure control is maintained by RCS letdown and / or RCS injection via the CCPs. As a precaution, the Pressurizer PORVs and Safety Valves are listed to insure RCS pressure relief is available for this path. RCS inventory control for path 1 is maintained by RCS injection by the CCPs and by maintaining RCP Seal flow. For path 2, RCS pressure control is maintained by the Pressurizer PORVs and Safety Valves and by RCS injection via the CCPs, SIPs and RHR Pumps. RCS inventory control for path 2 is maintained by the CCPs, SIPs and RHR Pumps and by maintaining RCP Seal flow.See discussion in Section 2.3.2 about containment performance.
2.3.2 Small Break (1-Inch Equivalent Break) LOCA Figure 2-2 shows the SPLD for the small LOCA scenario.
 
====2.3.2 Small====
Break (1-Inch Equivalent Break) LOCA Figure 2-2 shows the SPLD for the small LOCA scenario.
Core sub criticality, removal of decay heat, and RCS inventory control are required for the small LOCA. Sub criticality is achieved by insertion of the control rods into the core. Decay heat removal is accomplished by supplying auxiliary feedwater to the steam generators and removing heat from the steam generators through WBN2 IPEEE DESIGN REPORT Page 53 ATTACHMENT 1
Core sub criticality, removal of decay heat, and RCS inventory control are required for the small LOCA. Sub criticality is achieved by insertion of the control rods into the core. Decay heat removal is accomplished by supplying auxiliary feedwater to the steam generators and removing heat from the steam generators through WBN2 IPEEE DESIGN REPORT Page 53 ATTACHMENT 1
Watts Bar Nuclear Plant Unit 2 IPEEE Seismic Report No: WBNIPEEE-003 Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List Revision 0 Page: 18 the secondary side PORVs. Should decay heat removal via auxiliary feedwater and PORVs fail, then heat can be removed via a bleed and feed operation on the primary side as previously described.
Watts Bar Nuclear Plant Unit 2 IPEEE Seismic Report No: WBNIPEEE-003 Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List Revision 0 Page: 18 the secondary side PORVs. Should decay heat removal via auxiliary feedwater and PORVs fail, then heat can be removed via a bleed and feed operation on the primary side as previously described.
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The function of the CCS is to serve as an intermediate heat conductor for the removal of heat from potentially radioactive heat loads during normal and accident conditions.
The function of the CCS is to serve as an intermediate heat conductor for the removal of heat from potentially radioactive heat loads during normal and accident conditions.
This function is accomplished through the use of a closed-loop system in which the CCS removes heat from the various component coolers (CCS loads) and transfers it to the CCS heat exchangers where the heat is transferred to the ERCW system. Those portions of the CCS required following a SME are included in the SSEL. Portions of this system are isolated following a LOCA; the isolation valves are included in the SSEL. Check valves 2-CKV-070-0679,-0687,-0698, & 0790 also function as containment isolation valves, but are generically screened and are not listed on the SSEL. Other components serviced by the CCS such as the Waste Gas Compressor and Spent Fuel Cooling Heat Exchangers are not specifically required for the SMA, but are listed on the SSEL for CCS pressure boundary integrity.
This function is accomplished through the use of a closed-loop system in which the CCS removes heat from the various component coolers (CCS loads) and transfers it to the CCS heat exchangers where the heat is transferred to the ERCW system. Those portions of the CCS required following a SME are included in the SSEL. Portions of this system are isolated following a LOCA; the isolation valves are included in the SSEL. Check valves 2-CKV-070-0679,-0687,-0698, & 0790 also function as containment isolation valves, but are generically screened and are not listed on the SSEL. Other components serviced by the CCS such as the Waste Gas Compressor and Spent Fuel Cooling Heat Exchangers are not specifically required for the SMA, but are listed on the SSEL for CCS pressure boundary integrity.
3.8 Residual Heat Removal The RHR system is a safety-related system designed to perform functions during startup and cool down operations, shutdown operations, and during accident conditions.
 
===3.8 Residual===
Heat Removal The RHR system is a safety-related system designed to perform functions during startup and cool down operations, shutdown operations, and during accident conditions.
The RHR consists of two independent pump trains in each unit. With the exception of the common piping described below, each loop is capable of performing the safety-related and normal operating functions of the system.Each loop consists of a pump, pump miniflow loop, a heat exchanger, and flow control and isolation valves. Both loops share a common heat exchanger bypass line, suction piping from the RCS, suction and discharge to the RWST.The normal functions of the RHR system are used during reactor startup, cool down, shutdown, and refueling.
The RHR consists of two independent pump trains in each unit. With the exception of the common piping described below, each loop is capable of performing the safety-related and normal operating functions of the system.Each loop consists of a pump, pump miniflow loop, a heat exchanger, and flow control and isolation valves. Both loops share a common heat exchanger bypass line, suction piping from the RCS, suction and discharge to the RWST.The normal functions of the RHR system are used during reactor startup, cool down, shutdown, and refueling.
These normal functions of the RHR are: 1) To transfer decay heat from the RCS to the component cooling system when the RCS pressure and temperature are below RHR system design conditions.
These normal functions of the RHR are: 1) To transfer decay heat from the RCS to the component cooling system when the RCS pressure and temperature are below RHR system design conditions.
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TENNESSEE VALLEY AUTHORITY Title: Watts Bar Nuclear Plant, Report No.IPEEE Seismic Margins Evaluation, Unit 2 Relay Design Report WBNIPEEE-004 Prepared By: Michael G. Eason Date: 04/26/2010 Page 13 of 25 provides a normally closed contact to the "Safety Shutdown Reset" circuit and therefore qualifies as a Bad Actor Relay. The other relays are screened based on thleir operating mode.The function of relay RRXlA is to provide remote reset from main control room panel 0-M-26 (see Figure 5.3). Since there would be no reason for the circuit to be set, contact chatter would have no adverse effect and therefore the relay is screened.5.8 LOCA Containment Hydrogen Monitor Sampling Isolation Valves There is one GE Model 12HGAI 1J52 relay in the circuit of the four isolation valve pairs. The relays are designated as 42X1 & 42X2. During normal operation each relay is de-energized and two normally open contacts are used in the Auto-Open circuit of each valve pair (see Figure 5.4). GE type HGA relays used in this operating configuration do not qualify as Bad Actor Relays.5.9 WBNU1 "and WBNU2 System Differences In order to accept like for like equipment, it is also necessary to accept like for unlike equipment.
TENNESSEE VALLEY AUTHORITY Title: Watts Bar Nuclear Plant, Report No.IPEEE Seismic Margins Evaluation, Unit 2 Relay Design Report WBNIPEEE-004 Prepared By: Michael G. Eason Date: 04/26/2010 Page 13 of 25 provides a normally closed contact to the "Safety Shutdown Reset" circuit and therefore qualifies as a Bad Actor Relay. The other relays are screened based on thleir operating mode.The function of relay RRXlA is to provide remote reset from main control room panel 0-M-26 (see Figure 5.3). Since there would be no reason for the circuit to be set, contact chatter would have no adverse effect and therefore the relay is screened.5.8 LOCA Containment Hydrogen Monitor Sampling Isolation Valves There is one GE Model 12HGAI 1J52 relay in the circuit of the four isolation valve pairs. The relays are designated as 42X1 & 42X2. During normal operation each relay is de-energized and two normally open contacts are used in the Auto-Open circuit of each valve pair (see Figure 5.4). GE type HGA relays used in this operating configuration do not qualify as Bad Actor Relays.5.9 WBNU1 "and WBNU2 System Differences In order to accept like for like equipment, it is also necessary to accept like for unlike equipment.
This section identifies these differences at a design level and will be an integral portion of the verification activities at WBN which will be included in the final IPEEE report.The below systems have been .identified by the design Relay Evaluation team to have dissimilar or new qualities than that of the WBNU1 relay evaluation.
This section identifies these differences at a design level and will be an integral portion of the verification activities at WBN which will be included in the final IPEEE report.The below systems have been .identified by the design Relay Evaluation team to have dissimilar or new qualities than that of the WBNU1 relay evaluation.
5.9.1 System 043 -Sampling and Water Quality The only safety related function that System 43 provides is containment isolation of the containment penetrations to which the sample lines are attached.
 
====5.9.1 System====
043 -Sampling and Water Quality The only safety related function that System 43 provides is containment isolation of the containment penetrations to which the sample lines are attached.
Sampling System has fail-closed inboard and outboard containment isolation valves for each penetration to which a sample line is attached.
Sampling System has fail-closed inboard and outboard containment isolation valves for each penetration to which a sample line is attached.
Sampling and Water Quality System has no outgoing signals other than valve position indication and illumination.
Sampling and Water Quality System has no outgoing signals other than valve position indication and illumination.
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Shutdown and control rods are raised or lowered by a prescribed set of electromechanical actions performed by the control rod drive (CRD) mechanisms.
Shutdown and control rods are raised or lowered by a prescribed set of electromechanical actions performed by the control rod drive (CRD) mechanisms.
A review of the Vendor Technical Manual (WBN-VTD-W120-2568), Reference 2.13, for the Full Length Rod Control System was completed in order to verify whether any installed relays could negatively affect components identified by the SSEL. The review determined that Bad Actor Relays, as identified in Appendix E of Reference 2.7 "Low Ruggedness Relays", are not used in the Unit 1 Control Rod Drive System and will not be used in the Unit 2 system.After this design review, it was determined that the shutdown rods identified in the SSEL supply enough negative reactivity to maintain the reactor in a safe shutdown condition.
A review of the Vendor Technical Manual (WBN-VTD-W120-2568), Reference 2.13, for the Full Length Rod Control System was completed in order to verify whether any installed relays could negatively affect components identified by the SSEL. The review determined that Bad Actor Relays, as identified in Appendix E of Reference 2.7 "Low Ruggedness Relays", are not used in the Unit 1 Control Rod Drive System and will not be used in the Unit 2 system.After this design review, it was determined that the shutdown rods identified in the SSEL supply enough negative reactivity to maintain the reactor in a safe shutdown condition.
5.9.3 System 090 -Radiation Monitoring System There are ten flow control valves and four radiation monitors on the SSEL for this system. The four radiation monitors are for indication only and have no control function, therefore having no impact to the safe shutdown of WBNU2.The Unit 2 controls for these valves will be like the Unit I valve controls, therefore no bad actor relays will be used.5.9.4 System 092 -Neutron Monitoring System The WBNU2 Power Range and Auxiliary Equipment drawers are refurbished by Westinghouse and the relays used will be similar to Unit 1 but improved due to certain upgrades.
 
====5.9.3 System====
090 -Radiation Monitoring System There are ten flow control valves and four radiation monitors on the SSEL for this system. The four radiation monitors are for indication only and have no control function, therefore having no impact to the safe shutdown of WBNU2.The Unit 2 controls for these valves will be like the Unit I valve controls, therefore no bad actor relays will be used.5.9.4 System 092 -Neutron Monitoring System The WBNU2 Power Range and Auxiliary Equipment drawers are refurbished by Westinghouse and the relays used will be similar to Unit 1 but improved due to certain upgrades.
The function of this part of System 92 is identical to Unit 1 except a time delay circuit is added to Unit 2 Flux Deviation drawer, whereas Unit 1 has a time delay in the annunciator system for the QPTR alarms. Like the Unit 1 system, no bad actor relays will be used in the Unit 2 Neutron Monitoring System.WBN2 IPEEE DESIGN REPORT Page 193 ATTACHMENT 2
The function of this part of System 92 is identical to Unit 1 except a time delay circuit is added to Unit 2 Flux Deviation drawer, whereas Unit 1 has a time delay in the annunciator system for the QPTR alarms. Like the Unit 1 system, no bad actor relays will be used in the Unit 2 Neutron Monitoring System.WBN2 IPEEE DESIGN REPORT Page 193 ATTACHMENT 2
TENNESSEE VALLEY AUTHORITY Title: Watts Bar Nuclear Plant, Report No.IPEEE Seismic Margins Evaluation, Unit 2 Relay Design Report WBNIPEEE-004 Prepared By: Michael G. Eason Date: 04/26/2010 Page 15 of 25 5.9.5 System 099 -Reactor Protection System The purpose of the Reactor Protection System (RPS) is to provide automatic protection against unsafe and improper reactor operation during steady-state and transient power operations and to provide initiating signals to mitigate the consequences of faulted conditions.
TENNESSEE VALLEY AUTHORITY Title: Watts Bar Nuclear Plant, Report No.IPEEE Seismic Margins Evaluation, Unit 2 Relay Design Report WBNIPEEE-004 Prepared By: Michael G. Eason Date: 04/26/2010 Page 15 of 25 5.9.5 System 099 -Reactor Protection System The purpose of the Reactor Protection System (RPS) is to provide automatic protection against unsafe and improper reactor operation during steady-state and transient power operations and to provide initiating signals to mitigate the consequences of faulted conditions.
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TENNESSEE VALLEY AUTHORITY Title: Watts Bar Nuclear Plant, Report No.IPEEE Seismic Margins Evaluation, Unit 2 Relay Design Report WBNIPEEE-004 Prepared By: Michael G. Eason Date: 04/26/2010 Page 16 of 25 Currently, designed system, components do not house or rely on Bad Actor Relays.Verification activities will include confirming the SSEL components of this system are not subject to unacceptable relay chatter.5.9.6 System 236 -125 VDC Vital Power An addition to the WBNU2 SSEL included the Fifth Vital Battery and the components to distribute and transfer power from them to either Train A or Train B via a transfer switch located on the panel; making it completely independent and able to serve either Train as a backup for maintenance or accident purposes.
TENNESSEE VALLEY AUTHORITY Title: Watts Bar Nuclear Plant, Report No.IPEEE Seismic Margins Evaluation, Unit 2 Relay Design Report WBNIPEEE-004 Prepared By: Michael G. Eason Date: 04/26/2010 Page 16 of 25 Currently, designed system, components do not house or rely on Bad Actor Relays.Verification activities will include confirming the SSEL components of this system are not subject to unacceptable relay chatter.5.9.6 System 236 -125 VDC Vital Power An addition to the WBNU2 SSEL included the Fifth Vital Battery and the components to distribute and transfer power from them to either Train A or Train B via a transfer switch located on the panel; making it completely independent and able to serve either Train as a backup for maintenance or accident purposes.
The batteries, distribution panefs A&B and the 125 V Vital Battery Board V were confirmed to have no low seismic ruggedness relays and fail safe circuitry is adequate for the system. The charger for Battery V is not included in the SSEL since during the use of the batteries, charging them is performed through the use of the SPARE charger for the system it is serving. Both SPARE chargers were evaluated and qualified by the WBNU1 IPEEE.There are no Inverters specifically for the Vital Battery V loop. Existing inverters, also evaluated and qualified by the WBNUI IPEEE, are utilized from the battery that is unavailable.
The batteries, distribution panefs A&B and the 125 V Vital Battery Board V were confirmed to have no low seismic ruggedness relays and fail safe circuitry is adequate for the system. The charger for Battery V is not included in the SSEL since during the use of the batteries, charging them is performed through the use of the SPARE charger for the system it is serving. Both SPARE chargers were evaluated and qualified by the WBNU1 IPEEE.There are no Inverters specifically for the Vital Battery V loop. Existing inverters, also evaluated and qualified by the WBNUI IPEEE, are utilized from the battery that is unavailable.
6.0 RESULTS Through initial design review, there are no low-seismic-ruggedness relays used in applications which would qualify them as an essential relay; therefore no corrective action is anticipated.
 
===6.0 RESULTS===
Through initial design review, there are no low-seismic-ruggedness relays used in applications which would qualify them as an essential relay; therefore no corrective action is anticipated.


==7.0 VERIFICATION==
==7.0 VERIFICATION==
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This report describes the process used to evaluate the fire hazards for Unit 2 at Watts Bar Nuclear Plant and the results of that process. The evaluation was performed in response to the Individual Plant Examination for External Events (IPEEE) requested by Supplement 4 of Generic Letter 88-20 (NUREG-1407, Reference 19). The methodology used to perform this examination is based on the Fire Induced Vulnerability Evaluation (FIVE) methodology that was developed by the Electric Power Research Institute (EPRI), as described in Reference 2.The scope of the evaluation included the rooms that contain systems and equipment required to safely shutdown the unit in the event of a fire as identified in the Unit 2 Appendix R analysis.
This report describes the process used to evaluate the fire hazards for Unit 2 at Watts Bar Nuclear Plant and the results of that process. The evaluation was performed in response to the Individual Plant Examination for External Events (IPEEE) requested by Supplement 4 of Generic Letter 88-20 (NUREG-1407, Reference 19). The methodology used to perform this examination is based on the Fire Induced Vulnerability Evaluation (FIVE) methodology that was developed by the Electric Power Research Institute (EPRI), as described in Reference 2.The scope of the evaluation included the rooms that contain systems and equipment required to safely shutdown the unit in the event of a fire as identified in the Unit 2 Appendix R analysis.
The evaluation was based on the "as designed" configuration of Unit 2. A validation of this analysis will be conducted when plant construction is complete to confirm this evaluation.
The evaluation was based on the "as designed" configuration of Unit 2. A validation of this analysis will be conducted when plant construction is complete to confirm this evaluation.
1.1 Overview of the FIVE Methodology The EPRI FIVE methodology was used as a basis for evaluation of fire hazards and for screening fires from further consideration, based on screening criteria of less than 1 E-06 core damage frequency due to fire related initiating events.The FIVE documentation describes the fire evaluation process in three phases. The steps involved in each of these phases are described below.Phase I Qualitative screening and fire compartment interaction analysis During this phase, plant areas can be removed from further consideration based on the absence of safe shutdown equipment and no identified need for plant trip. In addition, fire boundaries are reviewed to ensure that a fire could not develop and then spread to other areas that may contain safe shutdown equipment or components.
 
===1.1 Overview===
of the FIVE Methodology The EPRI FIVE methodology was used as a basis for evaluation of fire hazards and for screening fires from further consideration, based on screening criteria of less than 1 E-06 core damage frequency due to fire related initiating events.The FIVE documentation describes the fire evaluation process in three phases. The steps involved in each of these phases are described below.Phase I Qualitative screening and fire compartment interaction analysis During this phase, plant areas can be removed from further consideration based on the absence of safe shutdown equipment and no identified need for plant trip. In addition, fire boundaries are reviewed to ensure that a fire could not develop and then spread to other areas that may contain safe shutdown equipment or components.
Phase II Quantitative evaluation of plant areas This phase accounts for the largest portion of effort for the fire hazard evaluation process and consists of the following steps.Phase II (Step 1) identified room specific and generic plant fire hazards and their associated fire ignition frequencies for those rooms that were not screened out during the Phase I analysis.
Phase II Quantitative evaluation of plant areas This phase accounts for the largest portion of effort for the fire hazard evaluation process and consists of the following steps.Phase II (Step 1) identified room specific and generic plant fire hazards and their associated fire ignition frequencies for those rooms that were not screened out during the Phase I analysis.
A total fire ignition frequency for each plant area is calculated as the sum of the individual ignition source frequencies in that area. EPRI identifies this frequency as F 1 andý if it has a value of less than 1E-06, then the room can be screened from further evaluation.
A total fire ignition frequency for each plant area is calculated as the sum of the individual ignition source frequencies in that area. EPRI identifies this frequency as F 1 andý if it has a value of less than 1E-06, then the room can be screened from further evaluation.
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Phase II (Step 3) If F 2 for a room is greater than 1 E-06, then additional evaluations were conducted.
Phase II (Step 3) If F 2 for a room is greater than 1 E-06, then additional evaluations were conducted.
These additional evaluations included determining the zone of influence from fire ignition sources and dividing the fires into severity probabilities using an "event tree" approach.PHASE III Results and Issues The final phase of the fire evaluation process consists of the documentation of results and identification of any new or remaining issues, including those addressed by the Sandia Fire Risk Scoping Study (NUREG/CR-5088) and the evaluation of containment performance.
These additional evaluations included determining the zone of influence from fire ignition sources and dividing the fires into severity probabilities using an "event tree" approach.PHASE III Results and Issues The final phase of the fire evaluation process consists of the documentation of results and identification of any new or remaining issues, including those addressed by the Sandia Fire Risk Scoping Study (NUREG/CR-5088) and the evaluation of containment performance.
1.2 Implementation of the EPRI FIVE Methodology The implementation of the EPRI FIVE methodology and the organization of this report is shown graphically in Figure 1-1. This implementation can be described as follows.Phase I The qualitative screening process is described in Section 2 of this report. A Fire Compartment Interaction Analysis (FCIA) was performed to determine if a given room contained any components required for fire safe shutdown (FSSD) or could require a plant trip. In addition, the room was evaluated for the potential for fire spread from an exposed room to an adjacent room.There were 17 rooms out of 140 rooms, listed in Table 6.1, screened from further evaluation through this process. This evaluation is documented in the Fire Compartment Interaction Analysis Report (Reference 8).Phase II The quantitative evaluation of the fire hazard frequency for each of the remaining rooms was then performed.
 
===1.2 Implementation===
 
of the EPRI FIVE Methodology The implementation of the EPRI FIVE methodology and the organization of this report is shown graphically in Figure 1-1. This implementation can be described as follows.Phase I The qualitative screening process is described in Section 2 of this report. A Fire Compartment Interaction Analysis (FCIA) was performed to determine if a given room contained any components required for fire safe shutdown (FSSD) or could require a plant trip. In addition, the room was evaluated for the potential for fire spread from an exposed room to an adjacent room.There were 17 rooms out of 140 rooms, listed in Table 6.1, screened from further evaluation through this process. This evaluation is documented in the Fire Compartment Interaction Analysis Report (Reference 8).Phase II The quantitative evaluation of the fire hazard frequency for each of the remaining rooms was then performed.
This was based on the guidance given in the EPRI FIVE documentation, which was implemented in a three step process.WBN2 IPEEE DESIGN REPORT.Page 211 ATTACHMENT 3
This was based on the guidance given in the EPRI FIVE documentation, which was implemented in a three step process.WBN2 IPEEE DESIGN REPORT.Page 211 ATTACHMENT 3
7 Phase II (Step 1) used the guidance in the EPRI FIVE documentation to generate fire ignition frequencies (F 1 values) for each room. These calculations are based on the plant-specific data listed in Sections 2 and 3.This process consisted of first allocating a plant area fire ignition frequency based on the assignment of each plant location to a generic type of area such as "Auxiliary Building, Switchgear Room", etc. Then the ignition sources were identified in each room and assigned a fire ignition frequency using the information provided by the FIVE documentation.
7 Phase II (Step 1) used the guidance in the EPRI FIVE documentation to generate fire ignition frequencies (F 1 values) for each room. These calculations are based on the plant-specific data listed in Sections 2 and 3.This process consisted of first allocating a plant area fire ignition frequency based on the assignment of each plant location to a generic type of area such as "Auxiliary Building, Switchgear Room", etc. Then the ignition sources were identified in each room and assigned a fire ignition frequency using the information provided by the FIVE documentation.
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Response Discussions with plant operators have confirmed that they regularly receive training in the use of the fire safe shutdown instructions.
Response Discussions with plant operators have confirmed that they regularly receive training in the use of the fire safe shutdown instructions.
: c. If, in performance of these procedures, operators are expected to pass through or perform manual actions in areas that may contain fire or smoke suitable SCBA equipment and other protective equipment are available for operators to perform their function.Response SCBA equipment is located in key locations throughout the plant, in addition to the equipment that is located in the fire brigade lockers.Plant operators receive regularly scheduled training in the effective use of this equipment.
: c. If, in performance of these procedures, operators are expected to pass through or perform manual actions in areas that may contain fire or smoke suitable SCBA equipment and other protective equipment are available for operators to perform their function.Response SCBA equipment is located in key locations throughout the plant, in addition to the equipment that is located in the fire brigade lockers.Plant operators receive regularly scheduled training in the effective use of this equipment.
7.2.5 Control Systems Interactions This issue centers on the concern that safe shutdown circuits are physically independent of, or can be isolated from, the control room for a fire in the control room fire area.Response The remote shutdown system at Watts Bar consists of the Auxiliary Control Room and shutdown boards that are located in the Auxiliary Building.
 
====7.2.5 Control====
Systems Interactions This issue centers on the concern that safe shutdown circuits are physically independent of, or can be isolated from, the control room for a fire in the control room fire area.Response The remote shutdown system at Watts Bar consists of the Auxiliary Control Room and shutdown boards that are located in the Auxiliary Building.
The remote shutdown system circuits are physically independent of, or can be electrically isolated from, the Main Control Room. Therefore, safe shutdown can be accomplished from outside the Control Building in the event of a severe fire in the Control Building that would cause Main Control Room abandonment.
The remote shutdown system circuits are physically independent of, or can be electrically isolated from, the Main Control Room. Therefore, safe shutdown can be accomplished from outside the Control Building in the event of a severe fire in the Control Building that would cause Main Control Room abandonment.
This capability is described in Part IV of the Fire Protection Report. The implementation of this capability is directed by Appendix C.69 of Abnormal Operating Instruction 30.2, Fire Safe Shutdown.7.2.6 Improved Analytical Codes WBN2 IPEEE DESIGN REPORT Page 253 ATTACHMENT 3
This capability is described in Part IV of the Fire Protection Report. The implementation of this capability is directed by Appendix C.69 of Abnormal Operating Instruction 30.2, Fire Safe Shutdown.7.2.6 Improved Analytical Codes WBN2 IPEEE DESIGN REPORT Page 253 ATTACHMENT 3

Revision as of 05:11, 14 October 2018

Watts Bar, Unit 2 - Individual Plant Examination of External Events Design Report
ML101240992
Person / Time
Site: Watts Bar Tennessee Valley Authority icon.png
Issue date: 04/30/2010
From: Bajestani M
Tennessee Valley Authority
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML101240992 (266)
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Text

Tennessee Valley Authority, Post Office Box 2000, Spring City, TN 37381-2000 April 30, 2010 10 CFR 50.54(f)U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Watts Bar Nuclear Plant, Unit 2 NRC Docket No. 50-391

Subject:

WATTS BAR NUCLEAR PLANT (WBN) UNIT 2 -INDIVIDUAL PLANT EXAMINATION OF EXTERNAL EVENTS DESIGN REPORT

References:

1. TVA letter dated January 29, 2008, 'Watts Bar Nuclear Plant (WBN) -Unit 2 -Regulatory Framework for the Completion of Construction and Licensing Activities for Unit 2" 2. NRC letter dated October 23, 2007, "Watts Bar Nuclear Plant, Unit 2 -Information Needed for Licensing Review Reconstitution" The purpose of this letter is to provide a Design Report describing the Individual Plant Examination for External Events (IPEEE) for WBN Unit 2. Enclosure 1 provides the IPEEE Design Report. In Reference 1, TVA committed to complete the evaluation for WBN Unit 2.The IPEEE for WBN Unit 2 is being completed in accordance with the U.S. Nuclear Regulatory Commission (NRC) Generic Letter (GL) 88-20, Supplements 4 and 5. In Reference 1, Enclosure 2, Item 128 provided a response to NRC's letter (Reference
2) of October 23, 2007, and NRC GL 88-20.The WBN Unit 2 IPEEE program is being conducted with an approach consistent with that used for the WBN Unit 1 program with the exception of the Low-Seismic-Ruggedness Relays that are described in Attachment 2 of the IPEEE Design Report. The outcome of the IPEEE Design Report is acceptable and is similar to that resulting from the WBN Unit 1 IPEEE. The corrective action programs from WBN Unit 1 are being implemented for WBN Unit 2, and thorough follow-up confirmatory actions are included in the WBN Unit 2 IPEEE program.Printed on recycled paper U.S. Nuclear Regulatory Commission Page 2 April 30, 2010 Enclosure 2 provides a list of commitments made in this submittal.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 30th day of April, 2010.If you have any questions, please contact me at (423) 365-2351.Sincerely, Masoud festani Watts nit 2 Vice President

Enclosures:

1. IPEEE Design Report 2. List of Commitments cc (Enclosures):

U. S. Nuclear Regulatory Commission Region II Marquis One Tower 245 Peachtree Center Ave., NE Suite 1200 Atlanta, Georgia 30303-1257 NRC Resident Inspector Unit 2 Watts Bar Nuclear Plant 1260 Nuclear Plant Road Spring City, Tennessee 37381 Enclosure 1 IPEEE Design Report Watts Bar Nuclear Plant Unit 2 (WBN2)Individual Plant Examination of External Events (IPEEE)Design Report Prepared by: Reviewed by: henJEder, P.E.v John 0. Dizon, P.E.Revision 0 April 27, 2010 WBN2 IPEEE DESIGN REPORT Page I Revision 0 April 27, 2010 TABLE OF CONTENTS Page 1. INTRODUCTION 3 1.1 Plant Familiarization 3 1.2 Seismic IPEEE 4 1.3 Other External Events 4 1.4 Fire-Induced Vulnerability Evaluation 5 1.5 Corrective Action Programs 5 1.6 List of Acronyms 8 2. SEISMIC IPEEE 10 2.1 Safe Shutdown Equipment List 10 2.2 Seismic Demand 10 2.3 Seismic Capabilities of Structures and Components 11 2.4 WBN2 Equipment Anchorage 15 2.5 WBN2 Seismic IPEEE Approach 16 3. Other External Events 22 3.1 High Winds and Tornados 23 3.2 Approach for WBN2 Other External Events 24 4. Fire-Induced Vulnerability Evaluation 28 4.1 Summary of the WBN1 FIVE Approach and Results 28 4.2 WBN2 FIVE Approach 29 5. Summary of Validation Activities 31 5.1 SSEL Validation 31 5.2 Relay Review Validation 31 5.3 Seismic Margin Assessment 32 5.4 Other External Events Assessment 32 5.5 FIVE Validation 33 6. References 34 Attachment 1: Report No. WBNIPEEE-003, "Watts Bar Nuclear Plant Unit 2, 36 Safe Shutdown Paths and Safe Shutdown Equipment List" Attachment 2: Report No. WBNIPEEE-004 "Watts Bar Nuclear Plant, IPEEE 179 Seismic Margins Evaluation, Unit 2 Relay Design Report" Attachment 3: Report No. WBN-IPEEE-005, "Watts Bar Nuclear Plant Unit 2 205 IPEEE Summary of Fire Evaluation Process" WBN2 IPEEE DESIGN REPORT Page 2 Revision 0 April 27, 2010

1.0 INTRODUCTION

This report describes the Individual Plant Examination for External Events (IPEEE) for Tennessee Valley Authority (TVA) Watts Bar Nuclear Plant Unit 2 (WBN2), including seismic, other external events, and the fire-induced vulnerability evaluations.

The results summarized herein provide a "design report," based largely on the ongoing and planned WBN2 final design and construction completion activities.

A final report will be submitted following certain validation activities, described herein, as WBN2 draws closer to operation.

As documented in TVA RIMS No. T04-980217-539 (Ref. 1), the IPEEE for WBN1 was completed in accordance with the U.S. Nuclear Regulatory Commission (NRC) Generic Letter (GL) No. 88-20, Supplement 4 (Ref. 2). In the WBN1 IPEEE investigation, no issues were found that would require mitigation through severe accident guidelines.

The WBN2 IPEEE program is being performed with an approach consistent with that used for the WBN1 program, and in general WBN2 is just like WBNI. In particular, the same corrective action programs from WBN1 are being implemented for WBN2. It is thus anticipated that the WBN2 IPEEE will have an acceptable outcome similar to that resulting from the WBN1 IPEEE. Thorough follow-up confirmatory actions are included in the WBN2 IPEEE program for validation of this anticipated outcome.The WBN1 seismic study and its applicability to WBN2 are summarized in Section 1.2.The IPEEE for other events and its applicability to WBN2 are summarized in Section 1.3, and the Fire-Induced Vulnerability Evaluation (FIVE) is described in Section 1.4.Background on the corrective action programs is provided in Section 1.5.1.1 PLANT FAMILIARIZATION Watts Bar Nuclear Plant is located in southeastern Tennessee on the west shore of Chickamauga Lake, approximately 50 miles northeast of Chattanooga and 31 miles northeast of the Sequoyah Nuclear Plant site. WBN is a two unit plant. Each unit employs a pressurized water reactor nuclear steam supply system with four coolant loops furnished by Westinghouse Electric Corporation.

WBN1 and WBN2 are essentially identical.

WBN1 and WBN2 shared structures include the Auxiliary and Control Building, Turbine and Service Buildings, Diesel Generator WBN2 IPEEE DESIGN REPORT Page 3 Revision 0 April 27, 2010 Buildings, and the Intake Pumping Station. Major independent structures for each WBN unit consist of an ice-condenser containment with free standing steel vessel, a reinforced concrete Shield Building, a main cooling tower, and a refueling water storage tank.1.2 SEISMIC IPEEE A Seismic Margin Assessment (SMA) was performed for WBN1 in accordance with the Electric Power Research Institute (EPRI) NP-6041-SL (Ref. 3) seismic margins methodology.

No design change recommendations resulted from the WBN1 IPEEE seismic evaluation.

A number of minor anomalies were noted during the WBN1 walkdowns, mostly requiring housekeeping or maintenance actions. The IPEEE program did not identify any adverse spatial interactions or any equipment or components with seismic capacity below the reference level of the Review Level Earthquake (RLE), i.e., 0.3g. In fact, the governing high confidence low probability of failure (HCLPF) seismic capacity of WBN1 was found to be 0.36g.In general, WBN2 is just like Watts Bar Nuclear Plant Unit 1 (WBN1). The ongoing WBN2 construction completion work is implementing the same design criteria and similar implementation procedures as WBN1. All of the same corrective action programs from WBN1 are being implemented for WBN2 (see Section 1.5). Therefore, the WBN2 HCLPF seismic capacity is expected to be at least as high as that determined for WBN1, and the results of the WBN1 seismic IPEEE should be fully applicable to WBN2. That is, the WBN2 HCLPF seismic capacity is expected to be at least 0.36g.Confirmatory seismic margins capacity walkdowns and evaluations will be performed after the bulk of the construction completion activities is completed.

The methodology that will be used for these confirmatory evaluations is described in Chapter 2. Validation activities are summarized in Chapter 5.1.3 OTHER EXTERNAL EVENTS The other external events for IPEEE include high winds, floods, transportation, and nearby facility accidents.

WBN1 performed the screening described in Supplement 4 to Generic Letter 88-20 (Ref. 2) and NUREG-1407 (Ref. 4). Because WBN1 was designed prior to the 1975 Standard Review Plan (SRP, Ref. 5), the general approach taken was WBN2 IPEEE DESIGN REPORT Page 4 Revision 0 April 27, 2010 to review the design bases and compare them to the SRP requirements.

A review was performed to determine if any changes around and at WBN had taken place since issuance of the operating license in November 1995. The WBN1 IPEEE evaluation revealed that the plant meets the 1975 SRP criteria for these external events, and only one recommendation for plant improvement resulted; namely, a recommendation to modify an Auxiliary Building concrete canopy to provide additional protection against tornado missiles.

This modification is common for WBN2.A goal of the WBN2 construction completion project is to make WBN2 as much as possible like WBN1. Thus, a similar effort as performed for WBN1 is used for the WBN2 other external events portion of the IPEEE. Additional discussion is provided in Chapter 3.1.4 FIRE-INDUCED VULNERABILITY EVALUATION The WBN1 plant vulnerability to internal fire events evaluation was assessed based on the Fire Induced Vulnerability Evaluation (FIVE) methodology developed by EPRI (Ref. 6). For those areas that did not pass the screen at the initial levels of evaluation, more detailed review techniques were utilized (e.g., zone of influence reviews for potential fires, segmentation of fire scenarios utilizing event trees). The results of the WBN1 detailed evaluation process were that all remaining plant areas were screened from further consideration while maintaining a conservative level of assumed system failures within the analysis.

This WBN1 evaluation confirmed that there are no fire-induced vulnerabilities associated with continued operation.

For WBN2, the same FIVE process as used for WBN1 was repeated for plant areas supporting WBN2. Additional discussion is provided in Chapter 4.1.5 CORRECTIVE ACTION PROGRAMS WBN2 is implementing the same CAPs as WBN1. A brief summary of the WBN1 Civil/Seismic CAPs follows: Seismic reanalysis of nine Category I structures was performed for WBN1 restart using the methodology described in the Seismic Analysis CAP (Ref. 7) and Design Criteria WB-DC-20-24 (Ref. 8), and approved by NRC Inspection Reports 50-390-89/21 and 50-391-89/21.

The nine structures were reanalyzed using current analysis WBN2 IPEEE DESIGN REPORT Page 5 Revision 0 April 27, 2010 methods, upgraded structural models, and site specific response spectra. The seismic models of the Interior Concrete Structures and Auxiliary Control Building were revised to include actual location of shear centers. Also, the torsional constants in the seismic models of the Interior Concrete Structure (ICS) and North Steam Valve Room (NSVR) were revised to consider effects of warping and new analyses were performed.

These upgraded building models were used in the development of the evaluation basis (Set B) and the new design/modification (Set B+C) amplified response spectra. The same seismic response spectra are being used for WBN2 evaluations (Set B) and new designs (Set B+C). Block walls were evaluated and accepted as part of the TVA response to the NRC Bulletin 80-11 (Ref. 9) for WBNI. Any WBN2 block walls that may not have been addressed in the WBN1 80-11 program are included in the WBN2 construction completion project.The Hanger and Analysis Update Program (HAAUP) addressed seismic as well as other issues relating to piping and pipe supports which were previously identified to the NRC. The scope of the HAAUP included TVA responses to NRC Bulletin 79-02 (Ref. 10) for verification of pipe support base plate flexibility and expansion anchorage factor of safety, and also Bulletin 79-14 (Ref. 11) for verification of input data used in the seismic qualification of piping systems. TVA completed action plans for four distinct categories of piping under the WBN1 HAAUP: 1) ASME Large Bore Piping and Supports;

2) ASME Small Bore Piping and Supports;
3) Category I(L)Piping and Supports; and 4) Instrument Lines and Supports.

Additional program -elements that were covered under the HAAUP include 1) Pipe Support Component Substitution;

2) Pipe Rupture; 3) Buried Piping; and 4) Equipment Interfaces.

A similar rigorous HAAUP is being implemented for WBN2 piping and pipe supports.The Integrated Interaction Program (liP) identified and evaluated potential seismic interaction hazards as well as interactions due to piping thermal expansion.

The different types of seismic interactions that were evaluated include the following: " Interaction due to structural failure and falling;" Spray interaction;" Impact due to flexure/displacement;" Commodity/component deformation due to building interface differential displacement, termed "shakespace" interaction.

WBN2 IPEFE DESIGN REPORT Page 6 Revision 0 April 27, 2010 The objective of the lIP walkdown was two-fold:

1) to perform "screening evaluation" using conservative screening criteria; and 2) to identify outliers, i.e., items that did not meet the screening criteria requirements; as well as bounding cases.Subsequent acceptance criteria evaluations of bounding cases and outliers were performed to ensure that the safety-related functional capability of commodities will not be compromised by seismic systems interaction.

The WBN1 program included comprehensive plant walkdowns of all WBN1 and common systems and areas. The WBN2 liP covers all of the applicable plant systems and areas not previously addressed in the WBN1 lIP.The Equipment Seismic Qualification (ESQ) Program covered all Category I equipment, including walkdowns of all items and 100-percent inspection of the anchorage.

Attributes of the ESQ Program include (1) completeness and retrievabililty of qualification documentation; (2) evaluation of equipment mounting conditions; and (3) resolution of all discrepancies between design documents and as installed conditions.

A similar rigorous ESQ program has been implemented for WBN2 and appropriate modifications are in progress.Category I(L) equipment was addressed in the lIP as described above (structural failure and falling interactions) on an area-by-area basis. WBN2 is using the same approach as WBN1 for seismic verification of Category I(L) equipment." The WBN1 and Common major commodities (cable tray, conduit, and HVAC systems) and their supports were re-evaluated and screened and/or walked down for location, structural adequacy, and anchorage issues. This was accomplished in accordance with their respective CAPs. The comparable WBN2 programs are covering all of the applicable plant systems and areas not previously addressed in the WBN1 programs.The above programs provided documentation of the WBN1 design basis seismic qualification.

This documentation was used throughout the WBN1 IPEEE to screen, assess, and verify seismic margin beyond design basis. Results of the WBN2 CAP will similarly be used in the WBN2 seismic IPEEE program. See Chapter 2 for additional discussion.

In addition to the above-described civil/seismic CAPs, another CAP of interest to IPEEE is for fire protection.

WBN2 is implementing the same Fire Protection CAP approach as WBN2 IPEEE DESIGN REPORT.Page 7 Revision 0 April 27, 2010 WBN1. The program includes documentation of the measures taken to evaluate violation of the Appendix.R requirements and issuance of design change notices (DCNs)to correct the deficiencies; review of SQN Appendix R allegations, as well as issues raised by the NRC during SQN inspections, for applicability to WBN and issuance of DCNs to correct the deficiencies; and Fire Protection Compliance Review to ensure WBN conformance with NRC requirements and applicable guidelines.

As with WBN1, the results of the Compliance Review will be used as the basis for developing the remaining scope of work (calculations, analyses, DCNs, and document updates) and the consolidation of fire protection documentation into an organized package to support and substantiate the Compliance Review.1.6 LIST OF ACRONYMS ACB Auxiliary Control Building ACI American Concrete Institute, AHU Air Handling Unit AOI Abnormal Operating Instruction ARS Amplified Response Spectrum ASME American Society of Mechanical Engineers BAMB Boric Acid Mixing Building CAP Corrective Action Program CCS Component Cooling System CDFM Conservative Deterministic Failure Mechanism DBT Design Basis Tornado DCN Design Change Notice DG Diesel Generator DGB Diesel Generator Building EMPAC Enterprise Maintenance Planning and Control EPRI Electric Power Research Institute ERCW Essential Raw Cooling Water ESQ Equipment Seismic Qualification FIVE Fire-Induced Vulnerability Evaluation FRC Facility Risk Consultants, Inc.FSAR Final Safety Analysis Report HAAUP Hangar and Analysis Update Program WBN2 IPEEE DESIGN REPORT Page 8 Revision 0 April 27, 2010 HCLPF High Confidence Low Probability of Failure ICS Interior Concrete Structure lip Integrated Interaction Program IPE Individual Plant Examination IPEEE Individual Plant Examination for External Events LOCA Loss of Coolant Accident LOOP Loss of Offsite Power MEL Master Equipment List MOV Motor Operated Valve NRC Nuclear Regulatory Commission NSVR North Steam Valve Room PGA Peak Ground Acceleration PMF Probable Maximum Flood PMP Probable Maximum Precipitation PRA Probabilistic Risk Assessment PSA Plant Safety Assessment RHR Residual Heat Removal RLE Review Level Earthquake RWST Refueling Water Storage Tank SB Shield Building SC Seismic Capacity Ratio SCV Steel Containment Vessel SD Seismic Demand Ratio SMA Seismic Margin Assessment SQN Sequoyah Nuclear Plant SRP Standard Review Plan SRT Seismic Review Team SSD Safe Shutdown SSE Safe Shutdown Earthquake SSEL Safe Shutdown Equipment List TVA Tennessee Valley Authority UNID Unique Identification WBN Watts Bar Nuclear Plant WBN2 IPEEE DESIGN REPORT Page 9 Revision 0 April 27, 2010 2.0 SEISMIC IPEEE WBN1 performed the seismic IPEEE using the EPRI Seismic Margin Assessment (SMA)approach (Ref. 3). This is a deterministic method which involves determination of the high-confidence-low-probability-of-failure (HCLPF) capacity for a subset of essential components and subsystems necessary to safely shutdown the reactor in the event of a specified earthquake greater than the design basis. The objective of the SMA is to define the margin above the design basis Safe Shutdown Earthquake (SSE) for the plant. HCLPF capacity is expressed in terms of peak ground acceleration (PGA).The walkdown effort for the WBN1 seismic IPEEE took credit for the extensive walkdowns associated with the numerous corrective action programs performed prior to issuance of the low power operating license. Salient features are described in Sections 2.1 through 2.4. In general, the same approach is used for WBN2, as described in Section 2.5.2.1 SAFE SHUTDOWN EQUIPMENT LIST The WBN1 Safe Shutdown Equipment List (SSEL) for the seismic IPEEE was developed using the EPRI SMA methodology (Ref. 3). The WBN1 SSEL identified Unit 1, common, and Unit 2 equipment required to achieve and maintain a Unit 1 safe shutdown condition for at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> following a seismic event and a seismic event concurrent with a small break Loss of Coolant Accident (LOCA).WBN2 systems and components function in the same manner as the Unit 1 systems and components.

Development of the WBN2 SSEL is summarized in Section 2.5.1.2.2 SEISMIC DEMAND The Set B+C new design/modification spectra was used in the WBN1 IPEEE program as the basis for defining seismic demand for equipment on the SSEL. Appropriate scale factors were used to boost those amplified response spectra (ARS) to the demand level of the RLE. The primary scale factors for the amplified portion of the ARS for WBN1 and Common Category I structures are shown in Table 2-1. A 1.25 scale factor for set B+C was applied to both the horizontal and vertical ARS in non-amplified portions of the ARS for all buildings.

In addition, scale factors for items supported on base slabs were WBN2 IPEEE DESIGN REPORT Page 10 Revision 0 April 27, 2010 determined on a case-by-case basis using the natural frequency of the item and the difference between the RLE spectrum and the Set B+C composite spectrum.These same scale factors and general approach are used for WBN2.2.3 SEISMIC CAPABILITIES OF STRUCTURES AND COMPONENTS In the WBN1 seismic IPEEE, HCLPF calculations were prepared for items which could not be screened by other means. In all cases, the calculated HCLPF values exceeded the RLE. HCLPF seismic capacities for WBN2 structures, systems, and components will also be confirmed to exceed the RLE. Table 2-2 summarizes the WBN1 HCLPF calculations and results. The evaluations for other key items in the WBN1 seismic IPEEE were as follows.Refueling Water Storage Tank (RWST)The WBN1 RWST was reevaluated during the Seismic CAP for Set B seismic responses and found to be adequate.

The design features of the tank were also extensively investigated by TVA and the NRC and any seismic vulnerabilities were addressed.

The WBN1 RWST was screened from further review based on (1) the conservative HCLPF of 0.48g calculated for the Sequoyah RWST, (2) the more rugged design of the WBN RWST, and (3) the well-designed foundation and anchorage of the WBN RWST.The seismic review team will perform an independent walkdown of the WBN2 RWST to assess if there are any new seismic vulnerabilities that could lead to structural failure during the RLE.Category I Block Walls WBN1 contains approximately 85 masonry block walls located in, or in association with, Category I Buildings (Control Building, Diesel Generator Building, Auxiliary Building, and Reactor Building).

They consist of three distinct types or may be divided into the following three categories:

1. Reinforced

-Hollow Core Mortared Masonry Walls *:* 25 -8" thick block walls in the Control Building* 13 -8" thick block walls in the Diesel Generator Building WBN2 IPEEE DESIGN REPORT Page 11 Revision 0 April 27, 2010 0 1 -8" thick block wall in the Reactor Building 0 19 -8" thick block walls in the Auxiliary Building 2. Unreinforced

-Mortared Solid Block Masonry Walls **-* 15 -6"x 8"x 12" block walls in the Auxiliary Building 3. Unreinforced

-Unmortared Solid Block Masonry Shielding Walls ***2 walls in the Reactor Building 10 walls in the Auxiliary Building* All the walls have some degree of fixity at top, bottom, and at the sides.** Unreinforced mortared solid block walls either have physical restraints or designated safe fall zones.Unreinforced unmortared solid stackable shield walls either have physical restraints or designated safe fall zones.All of these walls were reviewed as part of the NRC Bulletin 80-11 (Ref. 9) program prior to WBN1 start-up, by assessing a sample of WBN1 and common worst case masonry walls. The SMA of the WBN1 block walls utilized information from this design basis calculation to the extent possible including choice of most highly loaded walls, wall properties, and material properties.

Walls in the Control Building and Diesel Generator Building were selected as representative of the worst cases (greatest height, applied attachment loads, type of fixity, and design basis evaluation interaction ratios). All of these walls are 8" thickness, mortared and reinforced.

The WBN1 walls were evaluated using the procedures outlined in Appendix R of EPRI NP-6041 -SL (Ref. 3). A minimum conservative deterministic failure method (CDFM)capacity of 0.53g was found for the walls evaluated.

Therefore, in the WBN1 seismic IPEEE, failure of the masonry walls was not considered credible for the RLE.The seismic margin of WBN2 masonry walls will be confirmed for the WBN2 IPEEE program prior to start-up.Diesel Generator Building (DGB)The DGB could not be automatically screened out based on the criteria in Table 2-3 of EPRI NP-6041-SL (Ref. 3). The WBN1 SRT performed a walkdown of the DGB to WBN2 IPEEE DESIGN REPORT Page 12 Revision 0 April 27, 2010 identify any structural vulnerabilities.

No significant structural vulnerabilities that could lead to structural damage due to the RLE were identified.

During the WBN1 Seismic CAP, a detailed evaluation of worst-case concrete walls and slabs was performed to confirm the original design. None of the DGB walls and slabs were identified as worst-case features because of their robust design and moderate loads. Therefore, the DGB was screened from further evaluation based on the SRT walkdown coupled with the evaluation performed under the WBN1 seismic CAP.The DGB is common for WBN1 and WBN2. No further evaluation is required.Impact between Adjacent Structures The gap between the WBN1 Shield Building and Auxiliary Control Building (ACB) is 1" wide, and the gap between the ACB and the Waste Packaging Building is 2" wide. The SQN IPEEE demonstrated a large margin of approximately 5:1 against closure of the gap during out-of-phase seismic motion between the ACB and the Shield Building.

Although the scaled Set B+C responses are more conservative than those used for the SQN IPEEE evaluation, the gaps between WBN structures is more than adequate to prevent closure during an RLE.The gap between the WBN2 Shield Building and the ACB is also 1" wide, therefore no further evaluation is required for the WBN2 IPEEE program.Ice Basket Lower Seal (Divider Barrier Seal)The WBN1 SRT reviewed the determination of seismic margin for the seal during the SQN IPEEE. The SQN review determined that a margin of approximately 16:1 existed against exceedence of the tangential allowable movement of 1-in. The SRT confirmed that the WBN1 seal design and gap is similar to that of Sequoyah and that allowable displacement limits are similar.It will be confirmed that the same detail is used for the WBN2 ice basket lower seal, and this will establish the basis that this seal is screened from further review.Buckling of Containment The WBN1 steel containment was screened from evaluation in accordance with the criteria in EPRI NP-6041-SL (Ref. 3). Further, the containment structure was reviewed for the SQN IPEEE and determined to be acceptable with a margin of failure against WBN2 IPEEE DESIGN REPORT Page 13 Revision 0 April 27, 2010 buckling of 27:1. Based on the EPRI NP-6041-SL (Ref. 3) screening criteria and the large margin established during the SQN IPEEE evaluation, no further review was determined to be necessary for WBN1.The WBN2 containment shell is the same design as WBN1 and is, likewise, considered acceptable.

Auxiliary/Control Building The SQN IPEEE seismic evaluation determined that for all site building features the controlling HCLPF value was for the Aux. Bldg. High Bay Roof Diaphragm slab located between the two reactor buildings at elevation 791.75'. Since this feature was the controlling building item it was evaluated also for the WBN1 IPEEE. The SQN and WBN Auxiliary Buildings are based on the same design and are essentially the same with the exception of building elevations, seismic input levels and rotation of building axes by 90 degrees (i.e., SQN Aux. Bldg. N-S direction corresponds to WBN Aux. Bldg. E-W direction, and SQN elevation 791.75 corresponds to WBN elevation 814.75). The wall and slab thicknesses and reinforcing are equivalent.

The WBN1 seismic IPEEE approach included scaling SQN IPEEE seismic loads to WBN levels (this was done to develop both translational and torsional components of in-plane loads for the diaphragm) and evaluating the diaphragm in detail. The resulting HCLPF for the roof diaphragm was computed to be 0.75g for the RLE.The Auxiliary/Control Building is common to WBN1 and WBN2. No further evaluation is required for WBN2.Relay Evaluation Watts Bar Nuclear Plant is listed in NUREG-1407 Table 3.1 as a 0.3g focused scope plant. This designation requires a relay chatter evaluation as part of the IPEEE.Subsequent to NUREG-1407, Generic Letter GL 88-20 Supplement 5 (Ref. 13) was issued which modified the relay evaluation requirements for focused scope plants to: "drastically reduce the scope of relay chatter evaluation, retaining only the identification of bad actor relays." To comply with this requirement, WBN1 concluded the relay evaluation as specified, and no low seismic ruggedness relays were found to be installed in unacceptable configurations at WBN1. This verified design basis expectations.

See Section 2.5.2 for a summary of the WBN2 seismic IPEEE relay review.WBN2 IPEEE DESIGN REPORT Page 14 Revision 0 April 27, 2010 2.4 WBN2 EQUIPMENT ANCHORAGE For newly-installed base-anchored electrical equipment, the WBN2 ESQ program anchorage design basis envelopes the requirements for a 0.30g RLE HCLPF capacity as follows: Seismic Demand. IPEEE seismic demand is obtained from set B+C floor response spectra scaled up by a maximum factor of 1.25. The appropriate damping value for HCLPF capacity determination is 5% per EPRI NP-6041-SL (Ref. 3). The TVA WBN design criteria (WB-DC-40-31.2, Ref. 12) specifies 3%damping. Thus the HCLPF / design seismic demand ratio is as follows: SD = 1.25 x (3 / 5)1/2 = 0.97 < 1.0 This means that the 0.30g RLE HCLPF seismic demand level is less than the design criteria seismic demand level.Seismic Capacity.

In general, for expansion anchors, the EPRI margins HCLPF capacity of expansion anchors is based on a factor of safety ranging from 2.4 up to 3.6 for multiple bolts depending on the presence of cracks. In general, the upper bound factor of safety for WBN electrical panels is 3.2. The TVA anchor design specification DS-C-1.7.1 (Ref. 14) lower bound factor of safety, for wedge bolts, is 4.0. Thus the HCLPF / design seismic capacity ratio is: SC = 4 / 3.2 = 1.25 This means that the HCLPF seismic capacity level is higher than the design criteria seismic capacity level.Using the above 2 ratios, the minimum equivalent RLE for HCLPF capacity of WBN2 new electrical equipment panel anchorage is: HCLPF = 0.30g x ( 1.25 / 0.97 ) = 0.39g This is higher than the lower bound WBN1 HCLPF capacity as reported in Table 2-2.Thus no additional work is required for new electrical panel anchorage for WBN2 seismic IPEEE.WBN2 IPEEE DESIGN REPORT Page 15 Revision 0 April 27, 2010 2.5 WBN2 SEISMIC IPEEE APPROACH The approach being implemented for the WBN2 seismic IPEEE is summarized in the following Sections which address the following:

  • Safe Shutdown Equipment List* Relay Review* Walkdown and HCLPF Evaluations" Seismic Review Team and Peer Reviewers 2.5.1 Safe Shutdown Equipment List The WBN2 SSEL design report is included as Attachment 1 to this report. The WBN2 SSEL was developed primarily based on the WBN1 SSEL. That is, the original WBN1 SSEL was used as a starting point, formatting the equipment IDs to TVA's unique identification (UNID) format. DCNs for each SSEL system issued between 12/15/1997 and 12/31/1998 were reviewed to identify any equipment additions and deletions to the systems (note that the WBN1 SSEL was issued on 12/15/1997).

Any equipment additions and deletions were incorporated in the WBN1 SSEL as required.Changes to the Master Equipment List (MEL) after 12/31/1998 were reviewed to identify equipment additions and deletions (note that the TVA EMPAC MEL was issued on or about 12/31/1998).

The WBN1 SSEL was then updated and converted into the design report WBN2 SSEL.As described in Attachment 1, the final WBN2 SSEL will be issued following later validation and confirmation to include the following: " Upon availability of the WBN2 final design and design drawings as well as Master Equipment List (MEL), verify the WBN2 SSEL equipment ID, description, flow diagram, category, seismic category, and elevation/room against the completed MEL." Obtain TVA Probability Risk Assessment (PRA) Section and Operations reviews of the WBN2 SSEL and resolve comments as required.

It will be confirmed that the SSEL and success paths are those normally preferred by operations for shutdown following a seismic event.WBN2 IPEEE DESIGN REPORT Page 16 Revision 0 April 27, 2010 2.5.2 WBN2 Relay Review The WBN1 seismic IPEEE relay evaluation concluded that no low-seismic-ruggedness relays were used in applications which would qualify them as essential relays. No corrective actions were required.A similar evaluation will be performed for WBN2, using the 0.3g Focused Scope plant relay evaluation process in accordance with Supplement 5 of GL 88-20 (Ref. 13). That is, the evaluation is limited to only a review for low seismic ruggedness relays (bad actor relays). Bad actor relays are those subject to "chatter" during a review level earthquake."Chatter" is considered to be the inadvertent opening or closing of a relay contact with a sustained output of two milliseconds or more.The WBN2 relay review design report is included in Attachment

2. The design report concludes that WBN2, like WBN1, requires no corrective actions for the issue of Relay Chatter. As described in Attachment 2, the final WBN2 relay review will be issued following later verification including the following:
  • Evaluation of any changes to design documents utilized in the development of the WBN2 relay review design report;* Confirmation of the final WBN2 SSEL as described in Section 2.5.1;* A Control and Power Interface will be created to document any differences between Unit 1 and Unit 2, including but not limited to SSEL, control and/or power, and fail safe circuitry.

This Interface will be the source document for further investigation of Bad Actor Relays.If no differences in Unit 1 and Unit 2 are identified, this verification report will conclude that WBN requires no corrective actions for the issue of Relay Chatter.2.5.3 WBN2 Walkdown and HCLPF Evaluations Once the bulk of construction work on WBN2 is completed, a seismic margins walkdown will be performed for the items of equipment that are unique to the WBN2 SSEL in accordance with the requirements of the EPRI SMA methodology.

In addition, a walk-by will be performed for common items that were previously addressed in the WBN1 program as deemed necessary due to possible significant changes from WBN2 construction completion modifications.

WBN2 IPEEE DESIGN REPORT Page 17 Revision 0 April 27, 2010 Consistent with the WBN1 program, the WBN2 seismic margins review will factor in the results of other ongoing construction completion efforts, including corrective action programs (CAPs), the equipment seismic qualification program (ESQ), and the integrated interaction program (liP). WBN2 is implementing the same CAPs, ESQ, and lIP as WBN1 (see Section 1.5), using a similar approach and the same design criteria.The WBN2 seismic capacity calculations will be performed following the walkdowns, and will be based largely on those performed for WBN1 (see Section 2.3).One of the controlling features for the WBN1 0.36g HCLPF is the control room ceiling.After the WBN1 IPEEE program was completed, as described in Section 3.7.3.18 of the UFSAR (Ref. 15), a detailed evaluation including non-linear finite element analysis of the controlling features of the ceiling was performed.

This evaluation provides considerable insight to the seismic performance of the control room ceiling. Utilizing information from the detailed ceiling evaluation, an updated HCLPF capacity of the WBN main control room ceiling structure is found to be 0.52g (Ref. 16).As described in the Attachment 1 WBN2 SSEL design report, equipment items which were directly or indirectly referred to in the WBN1 SSEL report but not specifically listed in the SSEL were added to the WBN2 SSEL. The seismic margin assessment for these new items is as follows:* The Diesel Generator Seven Day Fuel Oil Tanks are screened out from further evaluation based on verification of engineered compacted foundation eliminating likelihood of differential settlement." Sump Strainers are screened out as inherently rugged." Walkdown evaluation of Fifth Vital Battery verified consistency with other WBN1 SSEL batteries as well as design drawings and seismic qualification basis." Walkdown evaluation of Control Rod Drive Mechanisms verified seismic bracing at top of frame enabling these to be screened out.* ERCW flow control valves, Safety Injection System (SIS) pump discharge isolation valves, and ERCW to Auxiliary Feedwater System isolation valves are screened out, noting that the WBN2 liP will perform the seismic proximity interaction reviews for these items.WBN2 IPEEE DESIGN REPORT Page 18 Revision 0 April 27, 2010 Later verification walkdown screening evaluations will be performed for the Centrifugal Charging Pump (CCP 2A& 2B) & Residual Heat Removal (RHR)Pump Room Coolers and Lower Compartment Cooler Fans.2.5.4 Seismic Review Team and Peer Reviewers The WBN2 Seismic Review Team (SRT) consists of Richard D. Cutsinger, John 0.Dizon, and Stephen J. Eder. These individuals are fully trained and meet all SRT requirements for the SMA. These individuals have a working knowledge of the WBN1 and WBN2 seismic programs, and recently were responsible for the successful completion of the SMA for TVA Browns Ferry Unit 1. In particular, the WBN2 SRT is currently responsible for the WBN2 lIP. Outlier resolution and bounding evaluation efforts from the lIP will directly consider seismic IPEEE implications and will be factored into the margins review as applicable.

The peer review team for the WBN2 seismic IPEEE will consist of former TVA individuals who served as WBN1 seismic IPEEE SRT members, namely Robert 0. Enis and Bill C. Perkins.WBN2 IPEEE DESIGN REPORT Page 19 Revision 0 April 27, 2010 Table 2-1: RLE Scale Factors for Category I Structures Scale Factors*Structure N-S Direction E-W Direction Vertical Auxiliary Control Building 1.0 1.16 1.25 Interior Concrete Structure 1.0 1.0 1.0 Shield Building 1.0 1.0 1.33 Steel Containment Vessel 1.0 1.0 1.25 Notes:* Scale Factor is for ARS in amplified portion of buildings WBN2 IPEEE DESIGN REPORT Page 20 Revision 0 April 27, 2010 Table 2-2: Resolution of WBN1 Seismic IPEEE Outliers.Item Building Resolution HCLPF Issue Spatial Interactions Various Interactions Seismic Spatial determined to be Interaction acceptable Masonry Walls Various 0.53g Stability 6900V Shutdown AUXILIARY 0.45g Structural Integrity Boards DG Air Intake Filters D.G. 1.78g Anchorage Main Control Rm CONTROL 0.56g Anchorage AHU Auxiliary Bldg. Roof AUXILIARY 0.75g Structural Integrity Diaphragm RHR Pumps AUXILIARY 0.50g Anchorage 480V Shutdown AUXILIARY 0.38g Structural Integrity Board Transformers Control Air Pre & Aft. AUXILIARY 1.08g Anchorage Filters CCS Heat AUXILIARY 0.38g Anchorage Exchangers ERCW Pumps IPS > 0.40g Anchorage IPS Screen Wash IPS 0.36g Anchorage Pumps 480V -Reactor AUXILIARY 0.40g Anchorage MOV, Reactor Vent, Control & Aux.Boards Main Cont. Rm. CONTROL 0.36g Anchorage Ceiling Structure Main Cont. Rm. CONTROL 0.70g Structural Integrity Electrical Panels WBN2 IPEEE DESIGN REPORT Page 21 Revision 0 April 27, 2010 3.0 OTHER EXTERNAL EVENTS To address the potential vulnerabilities at nuclear power plants from the effects of high winds, floods, and transportation and nearby facility accidents, NUREG-1407 (Ref. 4)recommends a progressive screening approach.

The first step in the process is to review the plant-specific design hazard information and licensing bases, including the resolution given for each event noted. The next step is to review the site for any significant changes since the operational license was issued with respect to the following:

1. Military and industrial facilities within 5 miles of the site;2. Onsite storage or other activities involving hazardous materials;
3. Transportation; or 4. Development that could affect the original design conditions.

Then the analyst should compare the information from the reviews conducted in the previous steps for conformance to the 1975 Standard Review Plan (SRP, Ref. 5) and perform a confirmatory walkdown.

If the comparison indicates that the plant conforms to the 1975 SRP and the walkdown reveals no potential vulnerabilities which were not included in the original design basis analysis, the IPEEE screening criteria are met.WBN is not a 1975 SRP plant, hence reviews of WBN1 were required of the design basis and any changes since the design for comparison to the SRP criteria.

The reviews and subsequent walkdowns confirmed that the plant does meet the 1975 SRP criteria and that these events may be screened out.Selection of external events for the IPEEE and the technical approach recommended for evaluation of such external events are discussed in NUREG-1407.

In the WBN1 IPEEE, the information and criteria used in the selection of external events included in NUREG-1407 was reviewed and the applicability of these to WBN examined, in order to confirm that no unique external events are excluded in the IPEEE. Table 3-1 documents the results of the WBN1 evaluations for High Winds, Floods, and Other External Events for TVA Watts Bar Nuclear Plant (WBN).As described in Table 3-1, the one exception noted and resolved during the WBN1 was for high winds and tornadoes.

This is further described in Section 3.1 below.WBN2 IPEEE DESIGN REPORT Page 22 Revision 0 April 27, 2010 3.1 HIGH WINDS AND TORNADOS A walkdown was performed during the WBN1 IPEEE to identify potential high wind and tornado concerns.

This evaluation concentrated'on outdoor tanks and equipment, entrances to concrete buildings, openings in buildings such as air intakes, diesel exhaust stacks, and louvers, block walls in structures with openings, structures which could collapse and impact buildings containing safety-related equipment, and availability of objects which could become missiles in a tornado.The following is a summary of the WBN1 walkdown observations with respect to high winds and tornadoes: " Metal-sided structures on site were verified to not contain Category I equipment.

The most significant metal-sided building, closest to a Category I structure, is the Turbine Building.

The Turbine Building is a metal-sided building whose panels are assumed to fail at loads less than Design Basis Tornado (DBT). The impact of the resulting missiles on other Category I structures has been evaluated in the design calculations and found to be acceptable.

No other metal-sided structures on site were found to be of greater significance than the Turbine Building as a source of tornado missiles." During the walkdown, it was confirmed that Category I building entrances and exterior openings in walls and slabs, which were determined to require protection as part of the design basis, are protected against tornado generated missiles which could penetrate and hit safety related equipment.

The only exception found was an opening in the concrete canopy on the unit 2 side of the Auxiliary Building, which had the potential to allow tornado missiles to penetrate the Auxiliary Building from the Unit 2 area. Plant modifications were implemented to correct this situation.

  • Block walls were qualified for tornado depressurization during the design basis evaluation.

During the confirmatory walkdown, no modifications to block wall were observed that would compromise the design basis evaluation." The only outdoor safety related tank is the RWST. Although it is not designed to withstand a DBT event, a storage basin is located around the tank to retain WBN2 IPEEE DESIGN REPORT Page 23 Revision 0 April 27, 2010 sufficient borated water in the event of a rupture of the tank due to tornado missiles.The number of potential objects available to be picked up by a tornado and become missiles was found to be not unusually large, in particular because there was no ongoing major construction activity at the plant site at the time of the WBN1 IPEEE program.3.2 APPROACH FOR WBN2 OTHER EXTERNAL EVENTS The findings and results from the WBN1 other external events IPEEE are fully applicable to WBN2. The following confirmatory reviews have already been performed for WBN2: Severe weather, lightning, and external fires were addressed in the WBN1 IPEEE program by reference to Loss of Offsite Power (LOOP) evaluations performed under the PSA for the internal events IPE. Similarly, detailed LOOP evaluations are included in the state-of-the-art PSA for the internal events IPE performed for WBN2 (Ref. 20) in accordance with RG 1.200 (Ref. 21)., It has been re-confirmed that WBN1 has no additional operating experience indicating that anything other than loss of offsite power would result from lightning strikes, In addition, WBN2 meets the requirements of NFPA Code No. 78-1975 (Ref. 18). Lightning protection was evaluated in detail (Ref. 19) and corrective actions were implemented including addition of air terminals on the WBN2 Reactor Building parapet wall (verified by walkdown 04/07/2010).

Therefore, as in WBN1, the generic data for screening out lightning for IPEEE is applicable to WBN2.* The transportation accident statistics and industrial and military facilities within 5 miles of the site were reviewed for any significant changes that may have.occurred since the WBN1 FSAR. This study was conducted as part of the preparation of the WBN2 FSAR (Ref. 22). It was concluded that none of the activities being performed in the vicinity of the site are a potential hazard to the plant.* Probable maximum flood (PMF) levels are currently being re-established by a TVA hydrology for all sites (Ref. 23). The results of this ongoing hydrology study WBN2 IPEEE DESIGN REPORT Page 24 Revision 0 April 27, 2010 indicate that flood levels at WBN may increase.

Any changes in PMF elevation are going to be handled by TVA as a WBN2 design issue; no further discussion is provided in this report." WBN2 is being outfitted with a new Siemens turbine generator system (Ref. 24)and this included a new study on turbine missiles (Ref. 25). The report concludes that the turbine missile probabilities remain below NRC limits and thus no further consideration is required for the WBN2 IPEEE." WBN Abnormal Operating Instruction AOI-8 (Ref. 26) requires that in the event of a tornado watch or warning that loose equipment be secured or removed in the General Yard, Switchyard, Transformer yard, Intake Pumping Station, Power Stores dock, and Cooling Tower areas. Nevertheless, additional walkdowns will be performed as described below for tornado missile effects review.Additional confirmatory review will be performed for WBN2 after the bulk of the construction completion activities is completed.

Items that will be specifically addressed include the following: " A site walkdown will be performed to identify potential objects that could be picked up by a tornado that may not be enveloped by the WBN1 evaluations.

This WBN2 walkdown will not be performed until after the WBN2 construction work is complete and associated loose materials and equipment have been taken away from the site. Any significant discoveries will be removed from the site or evaluated in detail for acceptability.

  • A new metal-sided Boric Acid Mixing Building (BAMB) is being added on the east side adjacent to the Turbine Building.

It will be confirmed that the tornado missile assumptions based on the Turbine Building wall panels are bounding relative to possible missiles originating from the BAMB.WBN2 IPEEE DESIGN REPORT Page 25 Revision 0 April 27, 2010 Table 3-1: Screening of External Events for the WBNI IPEEE Program Event Generic Basis Applicability to WBN1 High Winds NUREG- 1407 (Ref. 4) requests that this event A review of the UFSAR (Ref. 15) indicates that and Tornadoes be examined in the IPEEE. A progressive tornado wind design does not strictly meet the screening approach is recommended.

If the Standard Review Plan. Additional evaluations plant does not meet the NRC criteria (1975 were conducted as described in Section 4.1.version of the Standard Review Plan), more detailed examination is required.External NUREG-1407 requests that flooding be A review of the FSAR indicated that the design Floods evaluated if the plant design basis does not meets the NRC regulatory position 2 of the meet the criteria (Regulatory Guide 1.59, Regulatory Guide 1.59. The new PMP criteria Ref. 17). It also requires the use of the latest were reviewed and it was determined that WBN probable maximum precipitation (PMP) criteria is designed to withstand this flood and prevents which may result in higher site flooding levels water from entering safety related structures.

and greater roof ponding loads than have been used in the plant design basis.Transportation NUREG-1407 requests that older plants need The FSAR previously examined the impact of and Nearby systematic examination for plant specific potential transportation and nearby facility Facility vulnerabilities from these events, accidents and concluded that their contribution Accidents to plant risk is negligible.

The transportation accident statistics and nearby facilities were reviewed for any changes to this conclusion as part of the WBN1 IPEEE and none were found.Lightning In accordance with NUREG-1407, the primary Walls Bar meets the requirements of NFPA impact of lightning on nuclear power plants is Code No. 78-1975 (Ref. 18) and has no loss of offsite power which is included as part of additional operating experience indicating that the internal events IPE. The NRC staff has anything other than loss of offsite power would judged that the probability of a severe accident result from lightning strikes. Lightning protection caused by lightning (other that one due to loss of was evaluated in calculation WBN-EEB-MSTI-offsite power) is relatively low and further 190025 (Ref. 19). Therefore, the generic data consideration of lightning effects should be used in screening lightning is applicable to Watts performed only for plant sites where lightning Bar.strikes are likely to cause more than just loss of offsite power or a scram.Severe In accordance with NUREG-1407, the effects of Watts Bar site is not exposed to temperature Temperature these events are usually limited to reducing the transients more severe than other nuclear power Transients capacity of the ultimate heat sink and loss of plants in the U.S. Therefore, the generic data offsite power. The capacity reduction of the used in screening this event is applicable to ultimate heat sink would be a slow process that Watts Bar.allows plant operators sufficient time to take proper actions such as reducing power output level or achieving safe shutdown.

The other potential impact on the plant, loss of offsite power, will be considered within the realm of the station blackout rule. Therefore, the temperature transients need not be addressed in the IPEEE.WBN2 IPEEE DESIGN REPORT Page 26 Revision 0 April 27, 2010 Table 3-1: Screening of External Events for the WBN1 IPEEE Program, Continued Event Generic Basis Applicability to WBN1 Severe In accordance with NUREG-1407, the potential Watts Bar has no additional information to Weather effects of severe weather storms are loss of supplement NUREG-1407.

Therefore, the Storms offsite power and station blackout; these will be generic data used in screening of this event is addressed in the internal events IPE. Thus, applicable to Watts Bar.severe weather storms need not be examined further in the IPEEE.External Fires In accordance with NUREG-1407, the potential Watts Bar agrees with the generic basis and effects on the plant could be loss of offsite confirms that the plant site is generally cleared power, forced isolation of the plant ventilation, which would preclude the possibility of an and possible control room evacuation.

Usually, external fire spreading onsite. Therefore, external fires are unable to spread onsite external fires will not be considered further in the because of site clearing during construction IPEEE.stage. The effect of loss of offsite power will be addressed in the internal events IPE. The other effects have been evaluated during operating license review against sufficiently conservative criteria; thus they do not need to be reassessed in the IPEEE.Extraterrestrial In accordance with NUREG-1407, the probability Watts Bar agrees with the generic basis;Activity of a meteorite or satellite strike is estimated to therefore, this event will not be considered be negligibly small (less than 10-9) and the event further in the IPEEE.is dismissed on the basis of low event frequency.

Volcanic In accordance with NUREG- 1407, plant sites Watts Bar is far removed from an active volcano;Activity too far away from active volcanoes to expect therefore, this event will not be considered any effect need not be considered in the IPEEE. further in the IPEEE.Turbine Missile Based on the regular inspection of low pressure The plant arrangement for WBN is such that turbine discs and overspeed protection system safety related structures, systems and followed by the utilities, the probability of turbine components are essentially protected from low failure leading to missiles is considered trajectory turbine missiles.

FSAR Section 10.2.3 acceptably small. describes the analysis performed to estimate the probability of damage to WBN from turbine missiles.

The probability was determined to be less than 1 X 10-7 per year. Also, WBN is committed to an inspection program of the turbine discs on a regular basis. This provides the basis for not considering the turbine missiles further in the IPEEE.WBN2 IPEEE DESIGN REPORT Page 27 Revision 0 April 27, 2010 4.0 FIRE-INDUCED VULNERABILITY EVALUATION Determination of the WBN1 vulnerability to internal fire events, in response to Supplement 4 to Generic Letter 88-20 (Ref. 2), was accomplished by use of the EPRI FIVE methodology (Ref. 6). Section 4.1 provides an overview of the WBN1 FIVE program. The WBN2 fire vulnerability evaluation is an extension of and uses a methodology consistent with WBN1 as described in Section 4.2. The WBN2 FIVE evaluation is included as Attachment 3 to this report.4.1

SUMMARY

OF THE WBN1 FIVE APPROACH AND RESULTS The FIVE methodology consists of a progressive screening evaluation in which plant fire areas are screened from consideration based on qualitative information (Phase I) or by quantitative analysis (Phase II). Half (135 of 269) of the plant areas included within the scope of the WBN1 FIVE program were screened in Phase I. This phase consisted of screening fire areas based on area fire boundary integrity, the absence of safe shutdown components, and the lack of plant trip initiators.

Containment fires were also screened based on qualitative factors. The Phase 11 quantitative analysis then consisted of an initial quantitative evaluation, followed by a more detailed quantitative evaluation for areas that were not screened, based on a fire-induced core damage frequency of less than 1 E-06.The WBN1 initial quantitative evaluation consisted of generating an area-specific fire ignition frequency, then assuming that all fires totally engulf the affected area. Plant components that could be damaged by these fires were evaluated in the Appendix R analysis and documented in the Fire Protection Report. A "conditional core damage frequency" was then generated for each area by incorporating the failed components into the Plant Safety Assessment (PSA) model. Nearly half (60 of 132) of the remaining areas were screened during this phase.The detailed quantitative evaluation was then performed for the remaining areas (i.e.those that were not screened from consideration in the initial quantitative evaluation).

For those Control Building and Auxiliary Building areas that did not screen at an initial level of evaluation, more detailed review techniques were utilized (e.g., zone of influence reviews for potential fires, segmentation of fire scenarios utilizing event trees).WBN2 IPEEE DESIGN REPORT Page 28 Revision 0 April 27, 2010 The results of the detailed evaluation processes were that all remaining plant areas were screened from further consideration while maintaining a conservative level of assumed system failures within the analysis.

This evaluation has confirmed that there are no fire-induced vulnerabilities associated with the continued operation of the Watts Bar Nuclear Plant.The last part of the fire evaluation process addressed the response to and resolution of the Sandia Fire Risk Scoping Study (NUREG/CR-5088, Ref. 27) issues and the evaluation of containment isolation and heat removal. These issues were resolved by a combination of referral to the WBN1 Fire Protection Report (Ref. 28) or by discussion in Attachment 4 of the WBN1 IPEEE report (Ref. 1).4.2 WBN2 FIVE APPROACH The WBN2 FIVE will repeat the above-described process that was used for the WBN1 FIVE for fire areas that support WBN2 operation.

This will also include consideration of the following potential program additions:

  • Areas that support WBN2 operation will be examined from a dual unit perspective;
  • Identification of combustible load additions/deletions and ignition source additions/deletions that occurred after the WBN1 IPEEE report was published for common unit areas that support both WBN1 and WBN2.The WBN2 Fire Induced Vulnerability Evaluation (FIVE) is documented in Attachment 3.The FIVE represents an analysis of the "as designed" condition of WBN2. Since Generic Letter (GL) 88-20 specifies that the IPEEE evaluation should be based upon the"as built" configuration of the plant, the FIVE analyses of the "as designed" configuration will be validated when construction is complete to meet the "as built" GL 88-20 criterion.

The FIVE validation effort will be comprised of the following activities:

Validation Activities for Phase I The Unit 2 population of rooms with Appendix R Safe Shutdown (SSD)Equipment will be reviewed to ensure that no safe shutdown components or plant trip initiators have been added to the scope. If any of these are discovered, they will be evaluated via the FIVE process. A representative population of rooms will WBN2 IPEEE DESIGN REPORT Page 29 Revision 0 April 27, 2010 be reviewed to ensure that each room's configuration, barrier ratings, room use, etc. has not changed. Based on the results of this review, rooms will be reanalyzed as necessary and changes incorporated into the analysis.Validation Activities for Phase II, Step 1* A representative population of Unit 2 rooms will be reviewed to verify that there have been no significant changes in the room ignition frequencies which would result in a less conservative analysis result. New walkdowns will be performed and incorporated into the analysis as necessary.

Validation Activities for Phase II, Step 2 The "as built" equipment and location data for the Unit 2 Appendix R SSD equipment and safety injection/recirculation equipment will be reviewed and incorporated into the Plant Probabilistic Risk Assessment (PRA) as necessary to update the analysis.

Manual actions credited in the analysis will be confirmed.

Also, the latest Plant PRA will be compared to the "as designed" version of the model and updated if needed.Validation Activities for Phase II, Step 3* Report WBN-IPE-005 U2 will be updated as necessary.

This includes reviewing and updating both the assumptions and event trees as required.Validation Activities for Phase III* All applicable reports, including the summary, associated with the Unit 2 FIVE evaluation will be updated as necessary.

Other Validation Activities A peer review of the Unit 2 analysis will be performed prior to submittal of the Validated "As Built" Analysis Report. This review will be similar to the review performed for the Unit I evaluation.

WBN2 IPEEE DESIGN REPORT Page 30 Revision 0 April 27, 2010 5.0

SUMMARY

OF VALIDATION ACTIVITIES This Chapter provides, in punch-list format, a summary of the validation activities to be performed in order to enable confirmation of the results presented herein and submission of the final report for the WBN2 IPEEE implementation.

5.1 SSEL VALIDATION 5.1.1 Upon availability of the WBN2 final design and design drawings as well as Master Equipment List (MEL), verify the WBN2 SSEL equipment ID, description, flow diagram, category, seismic category, and elevation/room against the completed MEL.5.1.2 Obtain TVA Probability Risk Assessment (PRA) Section and Operations reviews of the WBN2 SSEL and resolve comments as required.

It will be confirmed that the SSEL and success paths are those normally preferred by operations for shutdown following a seismic event.5.1.3 An independent peer review of the WBN2 SSEL development process and results will be performed similar to the independent peer review that was conducted for the WBN1 SSEL.5.2 RELAY REVIEW VALIDATION

5.2.1 Evaluate

any changes to any of the design documents utilized in the development of the WBN2 relay review design report.5.2.2 Confirm that the bad actor relay identification encompasses the scope of equipment in the final WBN2 SSEL.5.2.3 Prepare a Control and Power Interface to document any differences between WBN1 and WBN2, including but not limited to the SSEL control and/or power and fail safe circuitry.

This Interface will be the source document for further investigation of bad actor relays.5.2.4 An independent peer review of the WBN2 Relay Review will be performed similar to the independent peer review that was conducted for the WBN1 Relay Review.WBN2 IPEEE DESIGN REPORT Page 31 Revision 0 April 27, 2010 5.3 SEISMIC MARGIN ASSESSMENT

5.3.1 Confirm

completion of corrective action programs of interest to the seismic margin assessment, in particular the HAAUP, lIP, and ESQ programs.5.3.2 Review the final SSEL and identify any new items or other significant deviations for seismic assessment.

5.3,3 Once the bulk of WBN2 construction work is completed, perform the seismic margins walkdown for the items of equipment that are unique to the WBN2 SSEL.5.3,4 Perform a walk-by for common items that were previously addressed in the WBN1 program as deemed necessary due to possible significant changes from WBN2 construction completion modifications.

5.3,5 Perform specific verification walkdown screening evaluations for the Centrifugal Charging Pump (CCP 2A& 2B) & Residual Heat Removal (RHR)Pump Room Coolers and Lower Compartment Cooler Fans.5.3.6 Perform an independent walkdown of the WBN2 RWST to assess if there is any new seismic vulnerability in relation to the WBN1 RWST seismic margin evaluation.

5.3.7 Confirm

the seismic margin of WBN2 masonry walls by comparison to WBN1 masonry walls.5.3.8 Confirm that the same detail as used in WBN1 is used for the WBN2 ice basket lower seal.5.3.9 An independent peer review of the WBN2 SMA will be performed similar to the independent peer review that was conducted for the WBN1 SMA, in accordance with the peer review requirements as described in the EPRI NP-6041-SL (Ref. 3).5.4 OTHER EXTERNAL EVENTS ASSESSMENT

5.4.1 Perform

a general site walkdown to identify potential objects that could be picked up by a tornado that may not be enveloped by the WBN1 evaluations.

WBN2 IPEEE DESIGN REPORT Page 32 Revision 0 April 27, 2010 5.4.2 Confirm that the metal-wall panels for the Boric Acid Mixing Building (BAMB)are bounded by the WBN1 IPEEE tornado missile assumptions based on the Turbine Building wall.5.5 FIVE VALIDATION 5.5.1 The Unit 2 population of rooms with Appendix R Safe Shutdown (SSD)Equipment will be reviewed to ensure that no safe shutdown components or plant trip initiators have been added to the scope. If any of these are discovered, they will be evaluated via the FIVE process. A representative population of rooms will be reviewed to ensure that each room's configuration, barrier ratings, room use, etc. has not changed. Based on the results of this review, rooms will be reanalyzed as necessary and changes incorporated into the analysis (Phase I).5.5.2 A representative population of Unit 2 rooms will be reviewed to verify that there have been no significant changes in the room ignition frequencies which would result in a less conservative analysis result. New walkdowns will be performed and incorporated into the analysis as necessary (Phase II, Step 1).5.5.3 The "as built" equipment and location data for the Unit 2 Appendix R SSD equipment and safety injection/recirculation equipment will be reviewed and incorporated into the Plant Probabilistic Risk Assessment (PRA) as necessary to update the analysis.

Manual actions credited in the analysis will be confirmed.

Also, the latest Plant PRA will be compared to the "as designed" version of the model and updated if needed (Phase II, Step 2).5.5.4 Report WBN-IPE-005 U2 will be updated as necessary.

This includes reviewing and updating both the assumptions and event trees as required.(Phase II, Step 3).5.5.5 All applicable reports, including the summary, associated with the Unit 2 FIVE evaluation will be updated as necessary (Phase III).5.5.6 A peer review of the Unit 2 analysis will be performed prior to submittal of the Validated "As Built" Analysis Report. This review will be similar to the review performed for the Unit 1 evaluation.

WBN2 IPEEE DESIGN REPORT Page 33 Revision 0 April 27, 2010

6.0 REFERENCES

1. Tennessee Valley Authority letter to United States Nuclear Regulatory Commission dated February 17, 1998, transmitting the Watts Bar Unit 1 Final Report, "Watts Bar Nuclear Plant (WBNP) Individual Plant Examination of External Events (IPEEE) Final Report," RIMS No. T04 980217 539.2. United States Nuclear Regulatory Commission, Generic Letter No. 88-20, Supplement 4 dated June 28, 1991, "Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities

-1OCFR 50.54(f)." 3. Electric Power Research Institute, "A Methodology for Assessment of Nuclear Power Plant Seismic Margin," Report No. EPRI NP-6041-SL, Revision 1, August 1991.4. United States Nuclear Regulatory Commission, Procedural and Submittal Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities," NUREG-1407, June 1991.5. United States Nuclear Regulatory Commission, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants," NUREG-0800, Revision 0, November 1975.*6. Electric Power Research Institute, "Fire-Induced Vulnerability Evaluation (FIVE)," Report No. EPRI TR-100370, Final Report, April 1992.7. Tennessee Valley Authority, "Watts Bar Nuclear Plant Seismic Analysis Corrective Action Closure Report," RIMS No. B26 910814 103, August 1991.8. Tennessee Valley Authority Design Criteria No. WB-DC-20-24, "Dynamic Earthquake Analysis Of Category I Structures And Earth Embankments," Revision 8, March 1991.9. United States Nuclear Regulatory Commission, IE Bulletin No.. 80-11, "Masonry Wall Design," May 1980.10. United States Nuclear Regulatory Commission, IE Bulletin No. 79-02, "Pipe Support Base Plate Design Using Concrete Expansion Anchor Bolts," March 1979.11. United States Nuclear Regulatory Commission, IE Bulletin No. 79-14, "Seismic Analyses for As-Built Safety-Related Piping Systems," July 1979.12. Tennessee Valley Authority Design Criteria No. WB-DC-40-31.2, "Seismic/Structural Qualification of Seismic Category I Electrical and Mechanical Equipment," Revision 11, April 2009.13. United States Nuclear Regulatory Commission, Generic Letter No. 88-20, Supplement 5 dated September 8, 1995, "Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities." WBN2 IPEEE DESIGN REPORT Page 34 Revision 0 April 27, 2010 14. Tennessee Valley Authority Nuclear Engineering Design Standard No. DS-C-1.7.1,"General Anchorage to Concrete," Revision 11, January 2006.15. Tennessee Valley Authority, "Updated Final Safety Analysis Report," Amendment 7, Submitted to the NRC September 2008.16. Tennessee Valley Authority Calculation No. WCG-2-617, "Seismic Margin Assessment of Main Control Room Ceiling," Revision 0, March 2010.17. United States Nuclear Regulatory Commission, Regulatory Guide 1.59, "Design Basis Floods for Nuclear Power Plants," Revision-2, August 1977.18. National Fire Protection Association, "Lightning Protection Code," Standard No.NFPA-78, 1975 Edition.19. Tennessee Valley Authority Calculation No. WBN EEB MSTI 190025, "Lightning Protection," Revision 2, RIMS No. T71080930800, September 2008.20. Tennessee Valley Authority Report, "Watts Bar Nuclear Plant Unit 2 Probabilistic Risk Assessment, WBN Unit 2 IPE Summary Report," February 2010, RIMS No. T02 100209 001.21. United States Nuclear Regulatory Commission, Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 2, March 2009.22. Tennessee Valley Authority, "Wafts Bar Nuclear Plant Unit 2 Living Final Safety Analysis Report A97." 23. Tennessee Valley Authority, Problem Evaluation Report (PER) No. 177501,"Hydrology

-Fort Loudon Dam Spillway Discharge Coefficient Inconsistencies," July 2009.24. Siemens Power Generation, Inc., "Tennessee Valley Authority Wafts Bar Nuclear Unit 2 Turbine Generator Island Project," TVA Contract No. 65008.25. Siemens Energy Inc., "Missile Report, TVA, Watts Bar 2," Report No. CT-27467, October 6, 2009, Rev. 0.26. TVA Watts Bar Nuclear Plant Abnormal Operating Instruction, "Tornado Watch or Warning," AOI-8, Rev. 47.27. Sandia National Laboratories, "Fire Risk Scoping Study: Investigation of Nuclear Power Plant Fire Risk, Including Previously Unaddressed Issues," prepared for the U.S. NRC, NUREG/CR-5088, January, 1989.28. Tennessee Valley Authority, "Wafts Bar Nuclear Plant Fire Protection Report," Revision 8, March 15, 1997.WBN2 IPEEE DESIGN REPORT Page 35 Revision 0 April 27, 2010 Attachment 1: Watts Bar Nuclear Plant Unit 2 Preliminary IPEEE Seismic Margins Evaluation Design Report, Safe Shutdown Paths and Safe Shutdown Equipment List Report No: WBNIPEEE-003 WBN2 IPEEE DESIGN REPORT Page 36 Watts Bar Nuclear Plant Unit 2 Preliminary IPEEE Seismic Margins Evaluation Design Report Report No: WBNIPEEE-003 SAFE SHUTDOWN PATHS And SAFE SHUTDOWN EQUIPMENT LIST April 26, 2010 WBN2 IPEEE DESIGN REPORT Page 37 ATTACHMENT 1

Prepared : Michael Snider Donald G. Fickey? " Reviewed by: by: eneh1 Awo Approved Approve d 4/04Io/t April 26, 2010 Page 2 WBN2 IPEEE DESIGN REPORT Page 38 ATTACHMENT 1

Watts Bar Nuclear Plant Unit 2 IPEEE Seismic Report No: WBNIPEEE-003 Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List Revision 0 Page: 3 TABLE OF CONTENTS Section Title Page

1.0 INTRODUCTION

8 1.1 Background

1.2 Methodology

1.3 Work Scope Statement 1.4 Safe Stable Shutdown States 2.0 SAFE SHUTDOWN PATHS 13 2.1 Assumptions

2.2 Resources

Used 2.3 Safe Shutdown Path Descriptions

2.3.1 Intact

RCS 2.3.2 Small LOCA 2.4 Dependency Matrices 3.0 SYSTEM DESCRIPTIONS 28 3.1 Main Steam 3.2 Main Feedwater 3.3 Auxiliary Feedwater 3.4 Chemical and Volume Control 3.5 Safety Injection 3.6 Essential Raw Cooling Water 3.7 Component Cooling System 3.8 Residual Heat Removal 3.9 Reactor Protection System 3.10 Electrical Power System 3.11 Miscellaneous Systems 4.0 SAFE SHUTDOWN EQUIPMENT LIST 64 4.1 Approach 4.2 Definition of Table Entries 4.3 Safe Shutdown Equipment List 5.0 PEER REVIEW 72 WBN2 IPEEE DESIGN REPORT Page 39 ATTACHMENT 1

Watts Bar Nuclear Plant Unit 2 IPEEE Seismic Report No: WBNIPEEE-003 Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List Revision 0 Page: 4 TABLE OF CONTENTS (cont.)Section Title Page 6.0 7.0

SUMMARY

REFERENCES 73 74 75 Appendix A SSEL of Unit 1 Required Equipment WBN2 IPEEE DESIGN REPORT Page 40 ATTACHMENT I

Watts Bar Nuclear Plant Unit 2 IPEEE Seismic Report No: WBNIPEEE-003 Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List Revision 0 Page: 5 THIS PAGE INTENTIONALLY LEFT BLANK WBN2 IPEEE DESIGN REPORT Page 41 ATTACHMENT 1

Watts Bar Nuclear Plant Unit 2 IPEEE Seismic Report No: WBNIPEEE-003 Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List Revision 0 Page: 6 Table Number IA 2.1 2.2 2.3 3.1 4.1 LIST OF TABLES Title SSEL Verification Activities Watts Bar IPEEE Safe Shutdown Paths, Detailed List of References Used Watts Bar Support to Front Line Dependency Matrix Watts Bar Support to Support Dependency Matrix Watts Bar Seismic Study, Systems Summary List Key for Reading Safe Shutdown Equipment List 9 24 26 27 29 66 Page WBN2 IPEEE DESIGN REPORT Page 42 ATTACHMENT I

Watts Bar Nuclear Plant Unit 2 IPEEE Seismic Report No: WBNIPEEE-003 Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List Revision 0 Page: 7 LIST OF FIGURES Figure Number 2-1 2-2 Title Success Path Logic Diagram, Intact RCS Scenario Success Path Logic Diagram, Small Break (1-Inch Equivalent Pipe Diameter)

LOCA Page 22 23 WBN2 IPEEE DESIGN REPORT Page 43 ATTACHMENT 1

Watts Bar Nuclear Plant Unit 2 IPEEE Seismic Report No: WBNIPEEE-003 Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List Revision 0 Page: 8

1.0 INTRODUCTION

The Unit 1 Safe Shutdown Equipment List (SSEL) was developed as part of the Unit 1 Individual Plant Examination of External Events (IPEEE) Seismic Margins Evaluation.

Using methodology described in the IPEEE Seismic Margins Evaluation Report, the SSEL identified Unit 1, common and Unit 2 equipment required to achieve and maintain a Unit 1 safe shutdown condition for at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> following two different seismic event scenarios.

One scenario is a seismic event beyond the design basis earthquake.

The second is a beyond design basis earthquake concurrent with a small break Loss of Coolant Accident (LOCA). The preliminary Unit 2 SSEL Design Report was developed utilizing the Unit 1 SSEL since the Unit 2 systems and components are designed and function the same as the Unit 1 systems and components.

Thus, the methodology used to develop the Unit 1 SSEL applies to Unit 2. The Unit 2 SSEL Design Report is preliminary since much of the Unit 2 design and documentation work remains to be done.The Unit 1 SSEL was first updated as, applicable, to account for changes to equipment and systems descriptions since the Unit 1 SSEL report was issued. Generally, Unit 2 equipment Unique Identification (UNID) numbers are the same as Unit 1 UNIDs except that the UNID number is preceded by the unit designation; i.e., 2- instead of 1-. Thus, the preliminary Unit 2 SSEL was established by converting the Unit 1 UNIDS and associated information to Unit 2 UNIDS. In similar manner, Unit 2 equipment UNIDs listed in the Unit 1 SSEL was converted to Unit 1 UNIDs. Equipment UNIDs common to both units were retained in the Unit 2 SSEL without change. The preliminary Unit 2 SSEL will be verified and validated (see following Table IA) when the Unit 2 design, drawings, and Unit 2 Master Equipment List (MEL) are completed and verified.

Upon completion of verification the preliminary Unit 2 SSEL Design Report will be issued as the final Unit 2 IPEEE Seismic Margins Evaluation Report.In addition to the above, the following equipment, which was directly or indirectly referred to in the Unit 1 report but not listed in the Unit 1 SSEL, was added to the preliminary Unit 2 SSEL Design Report: Fifth Vital Battery, Room Coolers for the following four pumps: Centrifugal Charging Pump (CCP 2A& 2B) and Residual Heat Removal pump (RHR 2A-A and 2B-B), ERCW flow control valves, Safety Injection System (SIS) pump discharge isolation valves, Control Rod Drive Mechanisms, ERCW to Auxiliary feed water system isolations valves, Lower Compartment Cooler Fans, Sump Strainers, and Diesel Generator Seven Day Fuel Oil Tanks.WBN2 IPEEE DESIGN REPORT Page 44 ATTACHMENT 1

Watts Bar Nuclear Plant Unit 2 IPEEE Seismic Report No: WBNIPEEE-003 Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List Revision 0 Page: 9 1.2.3.4.5.TABLE 1A SSEL VERIFICATION

/ VALIDATION ACTIVITIES Complete Unit 2 design and design drawings Complete Unit 2 Master Equipment List (MEL)Verify SSEL equipment ID, description, flow diagram, category, seismic category, and Elevation

/ room against the competed Unit 2 MEL.Verify SSEL against operations procedures Perform peer review.WBN2 IPEEE DESIGN REPORT Page 45 ATTACHMENT 1

Watts Bar Nuclear Plant Unit 2 IPEEE Seismic Report No: WBNIPEEE-003 Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List Revision 0 Page: 10 1.1 Background The general objectives of the Individual Plant Examination for External Events (IPEEE) are similar as stated in reference 3 are: 1) To develop an appreciation of severe accident behavior, 2) To understand the most likely severe accident sequences that could occur at the plant under full power operating conditions, 3) To gain a qualitative understanding of the overall likelihood of core damage and fission product releases, and, 4) If necessary, to reduce the overall likelihood of core damage and fission product releases by modifying, where appropriate, hardware and procedures that would help prevent severe accidents.

This report documents the definition of safe shutdown paths and the identification of associated critical equipment that would be used to accomplish the necessary safe shutdown and containment performance functions.

1.2 Methodology

References 2 and 3 document the request by the US Nuclear Regulatory Commission (NRC) that each active commercial nuclear power plant in the United States perform an IPEEE for specified beyond design basis events, one of which is a seismic event. The NRC has also identified three (3)acceptable approaches for the evaluation of the plant specific beyond design basis seismic event.The Tennessee Valley Authority (TVA) has chosen to employ the Electric Power Research Institute (EPRI) Seismic Margins Assessment method, (Ref. 3), at its Watts Bar Nuclear Plant (WBN).1.3 Work Scope Statement One element of EPRI's Seismic Margins Assessment methodology is a systems evaluation activity.The principle objective of the systems evaluation is to identify those components required for an operational sequence of plant systems that will bring the plant to a stable condition and maintain WBN2 IPEEE DESIGN REPORT Page 46 ATTACHMENT 1

that condition for at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. To accomplish this objective, it is necessary to develop an understanding of how the plant, as well as the plant operators, will perform following a seismic margin earthquake (SME).The specific steps to be followed in the systems evaluation are: Step 1 Determine the plant specific front line systems included in a preferred and alternate shutdown path that perform the following long-term safety functions:

0 Reactivity control El RCS pressure control 0 RCS inventory control 11 Decay heat removal El Containment performance Step 2 Determine the support systems required for the front line systems to operate using the documents listed in Table 2.1.Step 3 Determine the components that make up the front line and support systems again using the resources noted in Table 2.1.Step 4 Once the systems and system components are identified, define the active components of those systems that are required to accomplish the long term safety functions identified in Step 1.Step 5 Perform a walk down of the required components to gather detailed information that may not be available from drawings.

This includes anchoring of components and pipe runs, support of heavy loads above components, stiffening of adjacent electrical cabinets, and verification of component model and types.The execution of the above five steps provides for both the development of an understanding of the plant and plant operators response to an SME, and the identification of the minimum set of components required for an operational sequence of plant systems to bring the plant to a stable condition and maintain that condition for at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.WBN2 IPEEE DESIGN REPORT Page 47 ATTIACHMENT 1

Watts Bar Nuclear Plant Unit 2 IPEEE Seismic Report No: WBNIPEEE-003 Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List Revision 0 Page: 12 This report documents the results of activities performed to accomplish Steps 1 through 4. The walk downs will be discussed in greater detail in the seismic evaluation of the components comprising the Safe Shutdown Equipment List.1.4 Safe Stable Shutdown States For the purposes of this study, two safe, stable shutdown conditions are identified; one with the reactor coolant system (RCS) intact, and one with a loss of primary coolant from the RCS to the containment.

Intact RCS: Breech in RCS: This case assumes there is no uncontrolled release of primary coolant from the RCS to the containment.

The licensed long term safe, stable plant condition for WBN is Hot Standby, with the average primary coolant temperature greater than 3500 F and decay heat removal accomplished with auxiliary feedwater and atmospheric dumping of secondary side steam through the mainsteam PORVs.This case assumes a loss of primary coolant from the RCS to the containment, due either to a small leak which cannot be isolated, or due to the use of the pressurizer power operated relief valves (PORVs). The long term safe, stable plant condition for WBN is Cold Shutdown with the Residual Heat Removal (RHR) system recirculating coolant to the core from the containment sump.The reason for this second safe, stable shutdown condition is based on depletion of the refueling water storage tank (RWST). A small leak in the RCS which cannot be isolated will release high energy fluid to the containment, increasing containment pressure.

Similarly, if auxiliary feedwater is unavailable and bleed and feed is used to remove decay heat from the reactor core, the release of high energy primary coolant through the pressurizer PORVs to the containment will also act to increase containment pressure.

For WBN, virtually any release of high energy primary coolant to containment is predicted to actuate containment sprays. Operation of containment sprays will quickly deplete the refueling water storage tank (RWST) inventory, requiring that decay heat removal from the RCS be accomplished by recirculating water collected in the containment sump.Hence, if the RCS remains intact over the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> period of interest, the safe, stable long term shutdown condition for the SME is defined to be Hot Standby. If there is any loss of high energy coolant from the RCS to the containment, the safe, stable long term shutdown condition is defined to be Cold Shutdown with the RHR aligned for recirculation of fluid from the containment sump.WBN2 IPEEE DESIGN REPORT Page 48 ATTACHMENT 1

Watts Bar Nuclear Plant Unit 2 IPEEE Seismic Report No: WBNIPEEE-003 Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List Revision 0 Page: 13 2.0 SAFE SHUTDOWN PATHS This section describes the assumptions and process used to select the preferred and alternate Safe Shutdown Paths, and to define the paths.2.1 Assumptions The major systems-related assumptions and ground rules used in the systems evaluation portion of the WBN seismic margin evaluation are: 1) Path success is defined as the ability to achieve and maintain a safe, stable shutdown condition for at least a 72-hour period following the seismic 'event. For the case with the RCS intact and auxiliary feedwater available, this is Hot Standby.For the case with the RCS intact but auxiliary feedwater unavailable, or with a small LOCA having an equivalent break size of no more than a 1 -inch pipe, this is Cold Shutdown with recirculation from the containment sump.2) Offsite power is assumed to be failed due to the SME and unrecoverable during the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> time period of interest.

Credit for equipment operation (e.g., fail closed valve on L.O.P.) is not assumed.3) Only seismically induced transient events and small seismically induced primary coolant leakage events (referred to as "small LOCAs") are addressed.

The small LOCA is taken to be the combined leakage equivalent to a 1" diameter pipe break.4) Non-seismically caused component or system unavailability is not explicitly addressed.

However, both trains of safety related, equipment are listed on the SSEL.5) The potential effects of seismically induced relay chatter will be evaluated under a separate element of the WBN IPEEE study. This evaluation will be limited to low seismic ruggedness relays as defined in Appendix E of EPRI NP-7148-SL,"Procedure for Evaluating Nuclear Power Plant Relay Seismic Functionality".

Although the essential relays are not identified in this element of the study, the electrical panels and cabinets which house the electrical relays have been included in the Safe Shutdown Equipment List (SSEL).WBN2 IPEEE DESIGN REPORT Page 49 ATTACHMENT 1

Watts Bar Nuclear Plant Unit 2 IPEEE Seismic Report No: WBNIPEEE-003 Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List Revision 0 Page: 14 Assumption 4 states that non-seismic failures are not explicitly addressed.

Potential for human error is minimized by selecting Safe Shutdown Paths that require either no operator actions (such as for a loss of offsite power with AFW available), or a minimum of operator actions. In all cases operator actions required to achieve the Safe Shutdown Path are those normally included in the operator training program and trained on by operators; for example, performing bleed and feed operations in the event of a loss of AFW. This approach allows human error failures to be screened out using the guidelines identified in Section 2.4.3 of Volume 2 of NUREG/CR-4826 (Reference 4).2.2 Resources Used in the Safe Shutdown Path Development.

Resources listed in table 2.1 were reviewed and used, as determined applicable, in the development of the safe shutdown paths and critical equipment list developed for the WBN Seismic Margins assessment.

2.3 Shutdown

Path Descriptions Consistent with the EPRI methodology, two scenarios were considered in developing the Safe Shutdown Path for the WBN seismic margins assessment; Scenario 1 A "No-LOCA" scenario, which assumes the RCS remains intact throughout the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> time frame of interest, and, Scenario 2 A "Small LOCA" scenario, which assumes the breach in the primary system, occurs at the initiation of the SME.The reason for including small LOCA Safe Shutdown Path in the WBN SME is that it was judged that the effort necessary to demonstrate the availability of a source of makeup water to the RCS would be considerably less than the effort required to demonstrate that the integrity of the reactor coolant pump (RCP) seals and other small instrument and sensing lines would be maintained.

Preferred and alternate Safe Shutdown Paths were developed for both scenarios.

They were developed to be consistent with plant operator training and plant procedures.

Both Safe Shutdown Paths satisfy the plant safe shutdown criteria that the reactor is subcritical and that a stable cooling WBN2 IPEEE DESIGN REPORT Page 50 ATTACHMENT 1

Watts Bar Nuclear Plant Unit 2 IPEEE Seismic Report No: WBNIPEEE-003 Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List Revision 0 Page: 15 condition be established and maintained for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, accomplishing the following basic long-term safety functions:

1) Reactivity Control 2) RCS pressure control 3) RCS inventory control 4) Decay heat removal 5) Containment performance All systems used to accomplish the long terms safety functions are multi-train systems. In addition, equipment components necessary to ensure containment integrity have been identified and included in the SSEL. Check valves which perform a containment isolation function are generically screened and are not included in the SSEL (Ref. EPRI NP 6041-SL, RI, page 3-29).Success Path Logic Diagrams (SPLDs) are used to show both the preferred and alternate Safe Shutdown Paths developed for the WBN SME. The SPLD for Scenario 1 (Intact RCS) is shown in Figure 2-1, and the SPLD for Scenario 2 (Break Size 1 Inch Equivalent Diameter) is shown in Figures 2-2. Both preferred and alternate paths are shown for both scenarios.

2.3.1 Intact

RCS Considering Figure 2-1, the SPLD with an intact RCS, core sub criticality is required as is removal of decay heat, RCS pressure control, and RCS inventory control. Core sub criticality is attained by insertion of the control rods into the core. The preferred Safe Shutdown Path is for decay heat removal to be accomplished by supplying auxiliary feedwater to the steam generators and removing heat from the steam generators through the secondary side PORVs. Should decay heat removal via auxiliary feedwater and PORVs either fail or be unavailable, the alternate Safe Shutdown Path is for decay heat removal to be accomplished by means of a bleed and feed operation on the primary side. RCS pressure control is maintained by use of the pressurized PORVs and Safety Valves, and RCS injection via the CCPs, SIPs, and RHR Pumps. RCS inventory control is maintained by RCS injection using CCPs, SIPs, and RHR Pumps and maintaining R.C. Pump seal flow.WBN2 IPEEE DESIGN REPORT Page 51 ATTACHMENT 1

Watts Bar Nuclear Plant Unit 2 IPEEE Seismic Report No: WBNIPEEE-003 Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List Revision 0 Page: 16 Note that the preferred Safe Shutdown Path results in the plant being maintained at Hot Standby (average coolant temperature 3500 F), and that the alternate Safe Shutdown Path results in the plant going to Cold Shutdown.

This is discussed in some detail in Section 1.4.Insertion of the control rods into the core will occur when the electro-magnetic coils that hold the control rods lose power. This can occur: 1) Automatically upon a signal from the reactor protection system (RPS), 2) Manually by an operator action to trip the plant, or, 3) As a consequence of the loss of offsite power.Loss of offsite power is assumed to occur in this assessment.

Credit for loss of offsite power;however, is not assumed for this evaluation.

Consequently, the control rods will fall into the core upon receipt of a signal from the RPS. Emergency boration provides an alternate method of assuring sub criticality.

However, since sub criticality is achieved via control rod insertion by a signal from the RPS, only this mode for achieving sub criticality is-considered for the WBN SME.On loss of offsite power, the reactor coolant pumps will coast down to a stop and both the normal feedwater and the condenser system will be lost; procedure E-0 will be entered. Decay heat removal is accomplished via the auxiliary feedwater system (AFW) supplying feedwater to the steam generators and using the steam generator PORVs to remove heat per procedure ES-0. 1.Since the condensate storage tank, which is the normal supply of water to the auxiliary feedwater system, is not seismically qualified, the AFW system will be automatically aligned to the essential raw cooling water system (ERCW), its backup supply. This arrangement, considered the preferred path, is sufficient to keep the plant in Hot Standby conditions for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.Note that primary coolant temperature at Hot Standby, the licensed long term safe shutdown condition for Watts Bar when the RCS is intact, is above that for RHR cut in. Thus, the RHR system is not utilized for the preferred safe shutdown path with the RCS intact and AFW available.

An alternative success path for decay heat removal is the use of a bleed and feed operation on the primary side. This approach is independent of the preferred AFW and PORV path. With this approach, heat is removed from the core through the pressurizer PORVs (bleed) and made up by injecting water from the RWST via the centrifugal charging pumps or the safety injection pumps (feed) as directed by procedure FR-H. 1. With the bleed operation, the containment pressure will WBN2 IPEEE DESIGN REPORT Page 52 ATTACHMENT 1

Watts Bar Nuclear Plant Unit 2 IPEEE Seismic Report No: WBNIPEEE-003 Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List Revision 0 Page: 17 rise and actuate the containment spray system which also draws from the RWST. The water inventory in the RWST is limited and once it is depleted, cold leg recirculation from the containment sump is necessary.

Cold leg recirculation requires drawing water from the sump via the residual heat removal pumps and supplying it to either the charging or safety injection pumps per procedure ES-1.3. The charging and/or safety injection pumps then inject the water back into the RCS cold legs. Cold leg recirculation requires the following actions to be accomplished:

1) Establishing the flow path from the sump to the RHR pumps, 2) Starting the RHR pumps if they are not yet running, 3) Establishing the flow path from the RHR pumps to the charging and/or safety injection pumps, and, 4) Isolating the RWST.Procedures direct the operators to switch to hot leg recirculation as early as three hours after the beginning of a LOCA to preclude precipitation of Boron out of solution per procedure ES-1.4. Hot leg recirculation requires switching the RCS injection point from the cold legs to the hot legs. This alternative path is adequate to keep the plant in cold shutdown for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.For path 1, RCS pressure control is maintained by RCS letdown and / or RCS injection via the CCPs. As a precaution, the Pressurizer PORVs and Safety Valves are listed to insure RCS pressure relief is available for this path. RCS inventory control for path 1 is maintained by RCS injection by the CCPs and by maintaining RCP Seal flow. For path 2, RCS pressure control is maintained by the Pressurizer PORVs and Safety Valves and by RCS injection via the CCPs, SIPs and RHR Pumps. RCS inventory control for path 2 is maintained by the CCPs, SIPs and RHR Pumps and by maintaining RCP Seal flow.See discussion in Section 2.3.2 about containment performance.

2.3.2 Small

Break (1-Inch Equivalent Break) LOCA Figure 2-2 shows the SPLD for the small LOCA scenario.

Core sub criticality, removal of decay heat, and RCS inventory control are required for the small LOCA. Sub criticality is achieved by insertion of the control rods into the core. Decay heat removal is accomplished by supplying auxiliary feedwater to the steam generators and removing heat from the steam generators through WBN2 IPEEE DESIGN REPORT Page 53 ATTACHMENT 1

Watts Bar Nuclear Plant Unit 2 IPEEE Seismic Report No: WBNIPEEE-003 Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List Revision 0 Page: 18 the secondary side PORVs. Should decay heat removal via auxiliary feedwater and PORVs fail, then heat can be removed via a bleed and feed operation on the primary side as previously described.

RCS pressure control is maintained by use of the pressurized PORVs and Safety Valves, and RCS injection via the CCPs, SIPs, and RHR Pumps. RCS inventory control is maintained by RCS injection using CCPs, SIPs, and RHR Pumps and maintaining R.C. Pump seal flow.As discussed in Section 2.3.1 for the no-LOCA case, sub criticality for the small LOCA case is accomplished by insertion of the control rods into the core occurring when the electro-magnetic coils that hold the control rods lose power as a result of any one of the three conditions previously identified and procedure E-0 is entered. Decay heat removal is again accomplished by supplying feedwater to the steam generators via the AFW system, which has been aligned to draw suction from the ERCW system, and using the steam generator PORVs to remove heat per procedure E- 1.Similar to the removal of decay heat in the no-LOCA case, RCS inventory control is maintained by injecting water from the RWST into the RCS via either the charging pumps or safety injection pumps. With a small LOCA, the containment pressure will rise and actuate the containment spray system which also draws from the RWST. Following depletion of the RWST inventory, cold leg recirculation from the containment sump is necessary per procedure ES-1.3. This requires drawing water from the sump via the residual heat removal pumps and supplying it to either the charging or safety injection pumps. The charging and/or safety injection pumps then inject the water back into the RCS cold legs. The actions required to accomplish cold leg recirculation are identical to those described above.Again, an alternative success path for decay heat removal is using a bleed and feed operation on the primary side. This approach is independent of the preferred AFW and PORV path for decay heat removal. As discussed in Section 2.3.1, with this approach decay heat is removed from the core through the pressurizer PORVs (bleed) and made up by injecting water from the RWST via the centrifugal charging pumps or the safety injection pumps (feed) per procedure FR-H. 1. With the bleed operation, the containment pressure will rise and actuate the containment spray system which also draws from the RWST. Once the limited water inventory in the RWST is depleted, cold leg recirculation from the containment sump is necessary per procedure ES-1.3.For either the primary or. the alternate success paths, as early as three hours following the onset of a LOCA, procedures direct the operators to switch to hot leg recirculation so as to preclude Boron precipitation per procedure ES-1.4. Hot leg recirculation requires switching the RCS injection WBN2 IPEEE DESIGN REPORT Page 54 ATTACHMENT 1

point from the cold legs to the hot legs. This flow path is adequate to keep the plant in cold shutdown for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Thus, either the preferred or alternate Safe Shutdown Paths are adequate to keep the plant in a safe stable state for 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> time period of interest to the SME.Thus, the SME preferred paths for safe shutdown for WBN include control rod insertion, the AFW system, the steam generator PORVs, the RHR system in both the open and closed loop recirculation alignments, the high pressure safety injection and charging pumps, and the RWST.Also, all support systems that provide cooling and electrical power to equipment used to achieve safe shutdown, as well as the minimum instrumentation and safeguards actuation and control systems are included in the SME success paths.For path 1 (Fig. 2.2), RCS pressure control is maintained by RCS injection via the CCPs, SIPs and RHR Pumps. As a precaution, the Pressurizer PORVs and Safety Valves are listed to insure RCS pressure relief is available for this path. RCS inventory control for path 1 is maintained by RCS injection by the CCPs, SIPs and RHR Pumps, and by maintaining RCP Seal flow. For path 2, RCS pressure control is maintained by the Pressurizer PORVs and Safety Valves and by RCS injection via the CCPs, SIPS and RHR pumps. RCS inventory control for path 2 is maintained by RCS injection by the CCPs, SIPs and RHR pumps, and by maintaining RCP Seal flow.Although not specifically shown on Figures 2.1 and 2.2, equipment necessary to ensure containment performance subsequent to a seismic event has been included on the SSEL and will be evaluated in the SME evaluation.

This equipment includes containment isolation equipment, Ice Condenser, containment spray equipment, Lower Compartment Cooler Fans, and Containment Air Return Fans.The containment isolation system provides the means of isolating fluid systems that pass through containment penetrations so as to confine to the containment any radioactivity that may be released in the containment following a design basis event. The containment isolation systems are required to function following any design basis event that initiate a Phase A or Phase B containment isolation signal or releases radioactive materials into containment to isolate non-safety related fluid systems penetrating the containment.

Isolation design is achieved by applying common criteria to penetrations in many different fluid systems and by using Engineered Safety Features Actuation signals to actuate the appropriate equipment.

Phase A containment isolation, whose function is to prevent fission product release by isolating all lines not essential to reactor protection, can be generated manually or by a safety injection signal (SIS). A SIS is generated by one or more of the following:

WBN2 IPEEE DESIGN REPORT Page 55 ATTACHMENT 1

Watts Bar Nuclear Plant Unit 2 IPEEE Seismic Report No: WBNIPEEE-003 Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List Revision 0 Page: 20 1. Low steamline pressure in any steamline 2. Low pressurizer pressure 3. High containment pressure 4. Manually Phase B containment isolation is generated by the following:

1. High-high containment pressure 2. Manually Also, containment vent isolation signal is generated by the following:
1. A manual Phase A or B isolation actuation 2. A safety injection signal (SIS)3. High radiation signal from the containment purge air exhaust monitors The containment isolation signal is "fail safe" to the following extent: 1. Most of the containment isolation valves fail closed on loss of power or air and many are powered and controlled by battery backed power supplies.2. A number of isolation valves are normally closed air-operated valves or solenoid valves with de-energized control solenoid coils.3. A modified control circuit design is provided for specific motor operated valves whose possible inadvertent misalignment due to spurious operation could result in the loss of a system safety function.The glycol lines to the chillers inside the Ice Condenser are isolated following a containment isolation signal; these isolation valves are listed on the SSEL. However, the Ice Condenser and Ice Condenser doors, which are listed on the SSEL, are required to function for cases when primary coolant is released to the containment to reduce containment pressure and temperature.

The containment Air Return Fans are also on the SSEL since they are required to circulate from upper containment to lower containment, forcing any steam through the Ice Condenser.

The Containment Spray System is initiated upon receipt of Phase B containment isolation signal.This system consists of Containment Spray Pumps, Heat Exchangers and associated valves. The system serves to reduce containment pressure and temperature after a release of high energy fluids to the containment; this equipment is listed on the SSEL WBN2 IPEEE DESIGN REPORT Page 56 ATTACHMENT 1

Watts Bar Nuclear Plant Unit 2 IPEEE Seismic Report No: WBNIPEEE-003 Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List Revision 0 Page: 21 The Combustible Gas Control System (CGCS) is designed to detect and control the concentration of combustible gas mixtures inside primary containment following a large bore LOCA successfully mitigated by the Emergency Core Cooling System. The CGCS includes the Hydrogen Recombiners, the Hydrogen Analyzer, Containment Air Return Fans, which are redundant, safety grade subsystems and the non-safety grade Hydrogen Mitigation System (HMS). The HMS, which consists of two groups of 34 igniters distributed throughout the containment, was originally designed, procured and installed as completely safety-grade system that was redundant, Seismic Category I and Class IE (except for the igniter coil). However, a licensing commitment to install a complete safety-grade system was never made by TVA. As indicated above, the CGCS is required following a large break LOCA. Since the SME assessment involves (1) an intact RCS and (2) a small break (bore) LOCA, the CGCS is not included on the SSEL and is not evaluated as a part of the SME evaluation.

The WBN design does not include Penetration cooling as part of containment performance, and does not include inflatable seals around personnel hatches.2.4 Dependency Matrices Dependency matrices for the SME safe shutdown paths were developed to assure that all systems required to perform or support performance of the basic long-term safety functions identified in Section 2.3 would be considered in the development of the Safe Shutdown Equipment List. The WBN SME support system-to-front line system dependency matrix is given in Table 2.2. The.WBN SME support system-to-support system dependency matrix is given in Table 2.3. The support systems providing support are listed in the left hand column. The required support system(s) is identified by an "X" in the column associated with the system or function in question.The steam generator cooling mode and bleed and feed cooling mode which are considered for WBN require Essential Raw Cooling Water (ERCW), Component Cooling System (CCS), essential AC power (6.9-kV and 480V), vital instrument and control power (120V AC and 125V DC), the refueling water storage tank (RWST), the containment sump, Auxiliary Control (essential) air, heating, ventilating, and air conditioning (HVAC), the reactor protection system (Eagle 21), and the engineered safety features actuation system (ESFAS). These systems are included in the dependency matrices.WBN2 IPEEE DESIGN REPORT Page 57 ATTACHMENT 1

~3cs FIGURE 2-1 WITH SECONDARY SIDE HEAT SINK Lii 0 2: I-0: w z 0: C cj~w z 0 w I w 0 w I-WITHOUT SECONDARY SIDE HEAT SINK WBN2 IPEEE DESIGN REPORT Page 58 ATTACHMENT 1

Watts Bar Nuclear Plant Unit 2 IPEEE Seismic Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List Revision 0 Page: 23 Report No: WBNIPEEE-003 SMALL LOCA (_S I-INCH BREAK)FIGURE 2-2 WITH SECONDARY SIDE HEAT SINK Lu It Z (D U, z 0 i-I-o)z;SEE TABLES 2-2 AND 2-3 FOR A COMPLETE LIST OF SUPPORT SYSTEMS WITHOUT SECONDARY SIDE HEAT SINK WBN2 IPEEE DESIGN REPORT Page 59 ATTACHMENT 1

Watts Bar Nuclear Plant Unit 2 IPEEE Seismic Report No: WBNIPEEE-003 Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List Revision 0 Page: 24 TABLE 2.1 WATTS BAR IPEEE SEISMIC MARGINS INPUT DOCUMENTATION 1 Watts Bare Nuclear Plant IPEEE Seismic Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List, Report No.: WBNIPEEE-001, Revision 2, dated Dec. 15, 1997.2 WBN Final Safety Analysis Report (FSAR), Watts Bar Flow, Control, and Electrical Drawings.3 WBN Individual Plant Examination (IPE) Final Report 4 Review with Cognizant TVA Systems Engineers, and Operations.

5 System Descriptions and System Design Criteria 6 Sequoyah Nuclear Plant IPEEE Seismic Margins Evaluation Safe Shutdown Paths Report No. SCG-5M-0008, Revision I (B89 950620 010)7 Watts Bar Nuclear Plant Unit I Technical Specifications 8 Emergency Procedures E-0 Reactor Trip or Safety Injection Revision 28 E-1 Loss of Reactor Secondary Coolant Revision 15 ES-0. I Reactor Trip Response Revision 22 ES-1.3 Transfer to RHR Containment Sump Revision 17 ES-1.4 Transfer to Hot Leg Recirculation Revision 10 FR-H. 1 Loss of Secondary Heat Sink Revision 17 WBN2 IPEEE DESIGN REPORT Page 60 ATTACHMENT 1

Watts Bar Nuclear Plant Unit 2 IPEEE Seismic Report No: WBNIPEEE-003 Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List Revision 0 Page: 25 THIS PAGE INTENTIONALLY LEFT BLANK WBN2 IPEEE DESIGN REPORT Page 61 ATTACHMENT 1

Watts Bar Nuclear Plant Unit 2 IPEEE Seismic Report No: WBNIPEEE-003 Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List Revision 0 Page: 26 TABLE 2.2: WATTS BAR SUPPORT TO FRONTLINE DEPENDENCY MATRIX REQUIRED FRONTLINE SYSTEM OR FUNCTION SUPPORT SG 5G MSIV/ AFW REACT PZR RCP CVCS SIS RHR COLD LEG SUMP RHR HOT LEG CONTAINMENT AIR CONTAINMENT SYSTEM PORV BYPASS OR PORVS SEAL INJ SYSTEM RECIRCULATION SPRAY RECIRC ISOLATION RETURN SPRAY S VALVES SHUTD & FANS OWN COOLING 6.9 KVAC --4 '4 '4 '4 480V AC 4 4 4 4 4 4 4 4 4 4 4'125V DC 4 4 '4 '4 '4 __ _ ' '120VAC ' ', '4 '4 ESFAS 4 4 '4 '4 'HVAC -T ' 4 T CONTROL AIR , ' '4 '4-ESSENTIAL AIR PORTION ERCW --'COMPONENT

'4 '4 '4 '4 COOLING SYSTEM RWST 'CNTMTSUMP

'4 WBN2 IPEEE DESIGN REPORT Page 62 ATTACHMENT 1

Watts Bar Nuclear Plant Unit 2 IPEEE Seismic Report No: WBNIPEEE-003 Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List Revision 0 Page: 27 TABLE 2.3: WATTS BAR SUPPORT TO SUPPORT DEPENDENCY MATRIX REQUIRED SUPPORT SYSTEM SUPPORT 6.9k 480 VAC 480 VAC 480 VAC 480 VAC 125 VDC 120 VAC DIESEL FUEL SSPS / SHUTDOWN 480VAC 480VAC CCS / CONTROL ERC CCS SYSTEM VAC SO SO REACTO CNTL & AUX DIESEL VITAL VITAL INST GENERATO OIL ESFAS BD ROOM TRANS BD AFW AIR W BOARD BOARD R MOV BLDOG AUX BATTERY PWR RS VENT ROOM ROOM EQPMT SYSTEM S S BOARD VENT BOARDS BOARDS BOARDS VENT VENT COOLER BOARD S 6.9k VAC SD O 4 BOARDS 480 VAC SD BOARDS q 4 4 4 4 4 4 4 480 VAC C&A VEND 4 4 4 4 BD 480 VAC DIESEL AUX 4 BD 125 VDC VITAL BAT 4 4 4 4 4 4 4 BD 120 VAC VITAL INST / 4 4 PWR DIESEL 4 4/GENERATORS FUEL OIL SYSTEM 4 SSPS I ESFAS SHUTDOWN BD 4 4 VENT 480V TRANS RM.VENT 480V BD ROOM VENT CCS I AFW EQPMT 4 COOLERS ESSENTIAL AIR 4 SYSTEM ERCW 4 4 4 4 WBN2 IPEEE DESIGN REPORT Page 63 ATTACHMENT 1

Watts Bar Nuclear Plant Unit 2 IPEEE Seismic Report No: WBNIPEEE-003 Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List Revision 0 Page: 28 3.0 SYSTEMS DESCRIPTION The Seismic Margin Criteria specify that a success path or paths that will safely shut the plant down to either hot or cold stable shutdown, and maintain the plant in that condition for a minimum of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> must be demonstrated to survive and function properly during and after a Seismic Margin Event (SME). As discussed in Section 1.4, the two safe shutdown states used in the SME for WBN are: 1. For the intact RCS scenario using the preferred success path,, the safe shutdown condition is defined to be Hot Standby (Reactor coolant temperature 3500 F). Hot Standby is the licensed long term safe, stable shutdown condition for the Watts Bar units.2. For the intact RCS scenario using the alternate success path or the small LOCA scenario using either the preferred or alternate success paths, the safe shutdown condition is defined to be Cold Shutdown.Each of the mechanical systems at WBN was reviewed to assess their role in accomplishing the five long term safety functions listed in Section 2.3. A system by system listing is given in Table 3.1, which identifies those systems which perform an essential safety, function.

Summary descriptions of the identified essential systems are given in this section. For a detailed description of system operation, the reader is referred to the specific system description and design criteria.For schematic diagrams of systems that are presented in the following sections, the reader is referred to the system flow diagrams referenced in the SSEL. The Safe Shutdown Equipment List (SSEL), discussed in Section 4.0, consists of the specific components, associated with the essential systems, which must function properly to achieve the desired safe shutdown condition and to satisfactorily perform containment performance functions.

WBN2 IPEEE DESIGN REPORT Page 64 ATTACHMENT 1

Watts Bar Nuclear Plant Unit 2 IPEEE Seismic Report No: WBNIPEEE-003 Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List Revision 0 Page: 29 TABLE 3.1 WATTS BAR SEISMIC MARGINS STUDY SYSTEMS

SUMMARY

LIST System System Part of No. Description SSEL Comment 001 Main Steam YES Provides containment isolation and secondary side pressure control 002 Condensate NO Not required for safe shutdown or containment isolation.

003 Main Feedwater YES Provides containment isolation.

003B Auxiliary Feedwater YES Provides decay heat removal.005 Steam, Extraction NO Not required for safe shutdown or containment isolation.

006 Feedwater Heater Drains and Vents NO Not required for safe shutdown or containment isolation.

007 Turbine, Extraction Traps and NO Not required for safe shutdown or Drains containment isolation.

008 Turbine, Miscellaneous NO Not required for safe shutdown or Connections containment isolation.

009 Turbine, Miscellaneous Vents NO Not required for safe shutdown or containment isolation.

012 Auxiliary Boiler NO Not required for safe shutdown or containment isolation.

013 Fire Detection NO Not required for safe shutdown or containment isolation.

014 Condensate Demineralized NO Not required for safe shutdown or containment isolation.

015 Steam Generator Blowdown NO Not required for safe shutdown or containment isolation.

WBN2 IPEEE DESIGN REPORT Page 65 ATTACHMENT 1

Watts Bar Nuclear Plant Unit 2 IPEEE Seismic Report No: WBNIPEEE-003 Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List Revision 0 Page: 30 TABLE 3.1 WATTS BAR SEISMIC MARGINS STUDY SYSTEMS

SUMMARY

LIST (continued)

System System Part of No. Description SSEL Description 018 Fuel Oil YES Support system for emergency power generation.

019 Lighting -Off Oil and Air Piping NO Not required for safe shutdown or containment isolation.

020 Central Lubricating Oil NO Not required for safe shutdown or containment isolation.

024 Raw Cooling Water NO Not required for safe shutdown or containment isolation.

025 Raw Service Water NO Not required for safe shutdown or containment isolation.

026 High Pressure Fire Protection YES Required for containment isolation.

027 Condenser Circulating Water/ NO Not required for safe shutdown or Cooling Tower containment isolation.

028 Water Treatment NO Not required for safe shutdown or containment isolation.

029 Potable Water Distribution NO Not required for safe shutdown of containment isolation.

030 Ventilation YES Required for containment isolation and operation and cooling of essential equipment.

031 Air Conditioning (Cooling-Heating)

YES Required for cooling of essential equipment.

032 Control Air YES Required for operation of essential valves (Essential Air'Portion Only) and equipment.

033 Service Air NO Not required for safe shutdown or Containment isolation.

WBN2 IPEEE DESIGN REPORT Page 66 ATTACHMENT 1

Watts Bar Nuclear Plant Unit 2 IPEEE Seismic Report No: WBNIPEEE-003 Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List Revision 0 Page: 31 TABLE 3.1 WATTS BAR SEISMIC MARGINS STUDY SYSTEMS

SUMMARY

LIST (continued)

System System Part of No. Description SSEL Comment 035 Hydrogen, Generator Cooling NO Not required for safe shutdown or containment isolation.

036 Feedwater Secondary Treatment NO Not required for safe shutdown or containment isolation.

037 Gland Seal Water NO Not required for safe shutdown or containment isolation.

038 Insulating Oil NO Not required for safe shutdown or containment isolation.

039 CO 2 Storage, Fire Protection and NO Not required for safe shutdown or Purging containment isolation.

040 Drainage, Station NO Not required for safe shutdown or containment isolation.

041 Layup Water Treatment NO Not required for safe shutdown or containment isolation.

042 Chemical Cleaning NO Not required for safe shutdown or containment isolation.

043 Sampling and Water Quality YES Required for containment isolation.

System 044 Building Heating NO Not required for safe shutdown or containment isolation.

046 Feedwater Control System NO Not required for safe shutdown or (includes 046A through 046B) containment isolation.

047 Turbogenerator Controls NO Not required for safe shutdown or containment isolation.

WBN2 IPEEE DESIGN REPORT Page 67 ATTACHMENT I

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SUMMARY

LIST (continued)

System System Part of No. Description SSEL Comment 049 Air, Breathing NO Not required for safe shutdown or containment isolation.

050 Hypochlorite NO Not required for safe shutdown or containment isolation.

052 System Test Facility NO Not required for safe shutdown or (Seismic Instrumentation) containment isolation.

054 Injection Water NO Not required for safe shutdown or containment isolation.

055 Annunciator

& Sequential Events NO Not required for safe shutdown or Recording containment isolation.

056 Temperature Monitoring NO Not required for safe shutdown or containment isolation.

057 Generator Associated Electical NO Not required for safe shutdown or containment isolation.

058 Generator Bus Cooling NO Not required for safe shutdown or containment isolation.

059 Demineralized Water and Cask NO Not required for safe shutdown or Decontamination -containment isolation.

061 Ice Condenser YES Required for containment isolation and containment pressure control.062 Chemical Volume & Control YES Required for safe shutdown and containment isolation.

063 Safety Injection YES Required for safe shutdown.064 Ice Condenser Containment NO Not required for safe shutdown or containment isolation.

WBN2 IPEEE DESIGN REPORT Page 68 ATTACHMENT 1

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SUMMARY

LIST (continued)

System System Part of No. Description SSEL Comment 065 Emergency Gas Treatment NO Not required for safe shutdown or containment isolation.

067 Essential Raw Cooling Water YES Required for safe shutdown.(ERCW)068 Reactor Coolant System YES Required for safe shutdown.069 Plumbing NO Not required for safe shutdown or containment isolation.

070 Component Cooling Water YES Required for essential equipment operation during safe shutdown.072 Containment Spray YES Required for containment isolation and integrity.

074 Residual Heat Removal YES Required for safe shutdown &containment isolation 076 Volume Reduction and NO Not required for safe shutdown or Solidification System containment isolation.

077 Waste Disposal (Includes YES Required for containment isolation.

077A/Gas, 077B/Solid, 077C/Liquid) 078 Spent Fuel Pit Cooling YES Required for CCS pressure boundary 079 Fuel Handling and Storage NO Not required for safe shutdown or containment isolation.

080 Primary Containment Cooling NO Not required for safe shutdown or containment isolation.

WBN2 IPEEE DESIGN REPORT Page 69 ATTACHMENT 1

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SUMMARY

LIST (continued)

System System Part of No. Description SSEL Comment 081 Primary Makeup Water YES Required for containment isolation.

082 Standby Diesel Generators and YES Required for emergency power Diesel Starting generation.

083 Hydrogen Recombination NO Not required for safe shutdown or containment isolation.

084 Flood Mode Boration NO Not required for safe shutdown or containment isolation.

085 Control Rod Drive Mechanism YES Required for stopping Nuclear Reaction (Shutdown Rods).088 Containment Isolation YES Required for containment isolation.

Components listed under specific systems.090 Radiation Monitoring YES Required for containment isolation

&Containment Monitoring.

092 Neutron Monitoring System YES Required for reactivity monitoring.

094 Incore Flux Detectors NO Not required for safe shutdown or containment isolation.

99 Reactor Protection System YES Required for safe shutdown.200 161/6.9 KV Common Power NO Required for safe shutdown or containment isolation.

201 6.9 kV Unit Power NO Not required for safe shutdown or containment isolation.

202 6.9kV Reactor Cooling Pump NO Not required to power essential equipment Power for safe shutdown.WBN2 IPEEE DESIGN REPORT Page 70 ATTACHýENT 1

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SUMMARY

LIST (continued)

System System Part of No. Description SSEL Comment 203 480 V Unit Power NO Not required for safe shutdown or containment isolation.

204 480 V Switchyard Power NO Not required for safe shutdown or containment isolation.

205 480 V Turbine Building Common NO Not required for safe shutdown or Power containment isolation.

206 480 V Aux. Building Common NO Not required for safe shutdown or Motor Control containment isolation.

207 Turbine Building Common Motor NO Not required for safe shutdown or Control containment isolation.

208 Auxiliary Building Common NO Not required for safe shutdown or Motor Control containment isolation.

209 Turbine Building Motor Operated NO Not required for safe shutdown or Valve Power containment isolation.

210 Turbine Building Vent Power NO Not required for safe shutdown or containment isolation.

211 6.9kV Shutdown Power YES Required for safe shutdown 212 480 V Shutdown Power YES Required for safe shutdown 213 Reactor Motor Operated Valve YES Required for safe shutdown &Power containment isolation.

214 Control and Auxiliary Vent Power YES Required for safe shutdown &containment isolation.

215 Diesel Auxiliary Power YES Required for safe shutdown WBN2 IPEEE DESIGN REPORT Page 71 ATTACHMENT 1

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SUMMARY

LIST (continued)

System System Part of No. Description SSEL Comment 216 Fuel and Waste Handling Power NO Not required for safe shutdown or containment isolation.

217 Chemical and Volume Control NO Not required for safe shutdown or Power containment isolation.

218 Lube Oil Power NO Not required for safe shutdown or containment isolation.

219 Chlorination Building Power NO Not required for safe shutdown or containment isolation.

220 Makeup Water Power NO Not required for safe shutdown or containment isolation.

221 480V Service Building Power NO Not required for safe shutdown or containment isolation.

222 Service Building Vent Power NO Not required for safe shutdown or containment isolation.

223 Office Building Vent Power NO Not required for safe shutdown or containment isolation.

224 Gate House Power NO Not required for safe shutdown or containment isolation.

225 CCW Pumping Station Power NO Not required for safe shutdown or containment isolation.

226 Intake Pumping Station Power NO Not required for safe shutdown or containment isolation.

227 Turbine Building Lighting NO Not required for safe shutdown or containment isolation.

228 Auxiliary Building Lighting NO Not required for safe shutdown WBN2 IPEEE DESIGN REPORT Page 72 ATTACHMENT 1

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SUMMARY

LIST (continued)

System System Part of No. Description SSEL Comment 229 Service Building Lighting NO Not required for safe shutdown or containment isolation.

230 Office Building Lighting NO Not required for safe shutdown or containment isolation.

231 480 V Transformer Yard Power NO Not required for safe shutdown or containment isolation.

232 Reactor Vent Power YES Required for safe shutdown 233 Yard Lighting NO Not required for safe shutdown or containment isolation.

234 Heat Tracing Systems NO Not required for safe shutdown or containment isolation.

235 120 VAC Vital Power YES Required for safe shutdown 236 125 VDC Vital Power YES Required for safe shutdown.237 120 VAC Instrument Power YES Required for safe shutdown 238 120 VAC Preferred Power NO Not required for safe shutdown / Con. Iso.239 250 VDC Power NO Not required for safe shutdown or containment isolation.

240 48 VDC Power NO Not required for safe shutdown or containment isolation.

241 120 VAC Computer Power NO Not required for safe shutdown or containment isolation.

242 Rad Mon & Sampling PWR, NO Not required for safe shutdown or Process and Area Rad Mon Pwr containment isolation.

243 Recording Instrument Boards NO Not req'd For Safe Shutdown WBN2 IPEEE DESIGN REPORT Page 73 ATTACHMENT 1

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SUMMARY

LIST (continued)

System System Part of No. Description SSEL Comment 244 24 KV Power (Includes Main NO Not required for safe shutdown or Transformer).

containment isolation.

245 500 KV Switchyard Equipment

& NO Not required for safe shutdown or Cable Tunnel Cable Trays containment isolation.

246 Main Relay Boards NO Not required for safe shutdown or containment isolation.

247 24 VDC Communications Power NO Not required for safe shutdown or containment isolation.

248 Electrical Control and Recording NO Not required for safe shutdown or Instrument containment isolation.

249 Condensate Demineralized Motor NO Not required for safe shutdown or Control Center containment isolation.

250 Automatic, Manual & Public NO Not required for safe shutdown or Telephones containment isolation.

251 Sound-Powered Telephones NO Not required for safe shutdown or containment isolation.

252 Code Call, Paging, Intercom and NO Not required for safe shutdown or Evacuation Alarms containment isolation.

253 Microwave and VHF Radio NO Not required for safe shutdown or containment isolation.

254 Carrier Equipment NO Not required for safe shutdown or containment isolation.

255 Supervisory Control NO Not Req'd for Safe Shutdown 256 Sound Powered Communications NO Not required for safe shutdown 257 Closed-Circuit TV and Security NO Not required for safe shutdown WBN2 IPEEE DESIGN REPORT Page 74 ATTACHMENT 1

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SUMMARY

LIST (continued)

System System Part of No. Description SSEL Comment 258 Miscellaneous Audio NO Not required for safe shutdown or containment isolation.

259 Communications Room NO Not required for safe shutdown or containment isolation.

260 Data Logger, Data Acquisition NO Not required for safe shutdown or containment isolation.

261 Process Computer NO Not required for safe shutdown or containment isolation.

262 Load Shed Logic NO Not required for safe shutdown 263 Station Monitor Computer NO Not required for safe shutdown 264 Technical Support Center NO Not required for safe shutdown or containment isolation.

265 Computer Interface Equipment NO Not required for safe shutdown 268 Permanent Hydrogen Mitigation NO Not required for safe shutdown or Equipment containment isolation.

270 Handling Systems and Misc NO Not required for safe shutdown or Excluding Control & Aux bldgs. containment isolation.

WBN2 IPEEE DESIGN REPORT Page 75 ATTACHMENT 1

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SUMMARY

LIST (continued)

System System Part of No. Description SSEL Comment 271 Cranes YES Integrity required for safe shutdown 272 Water Treatment Plant Bldg. NO Not required for safe shutdown or Conduit & Cable Trays containment isolation.

275 Train Aux. Inst. Racks Only YES Required for Containment Iso.276 Local Instrument Control Racks YES Required for safe shutdown or containment isolation 278 Main and Auxiliary Control Panels YES Required for safe shutdown 280 Condenser Tube Cleaning NO Not required for safe shutdown or containment isolation.

281 Makeup Water Treatment Motor NO Not required for safe shutdown or Operated Valve Power containment isolation.

282 Field Service Facility Electrical NO Not required for safe shutdown or Equipment containment isolation.

283 Low Level Rad-waste Facility NO Not required for safe shutdown or containment isolation.

284 Volume Reduction Solidification NO Not required for safe shutdown or Facility containment isolation.

285 Spare Cables NO Not required for safe shutdown or containment isolation.

286 Security Power Backup Building NO Not required for safe shutdown or containment isolation.

290 Control Building Conduit and YES Required for safe shutdown Cable Trays 300 Miscellaneous Elect Equip NO Not required for safe shutdown WBN2 IPEEE DESIGN REPORT Page 76 ATTACHMENT 1

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SUMMARY

LIST (continued)

System System Part of No. Description SSEL Comment 301 Chemical Laboratory Equipment NO Not required for safe shutdown or containment isolation.

302 Health Physics Laboratory NO Not required for safe shutdown or Equipment containment isolation.

303 Prime Computer and Associated NO Not required for safe shutdown or Equipment containment isolation.

304 Penetrations and Sleeves NO No change of state or position required for (Mechanical) containment isolation.

510 Plant Equipment (non system NO Not required for safe shutdown or related) containment isolation.

900 Local Instrument Panels NO Not required for safe shutdown or containment isolation.

928 Makeup Water Treatment Plant NO Not required for safe shutdown or Equipment containment isolation.

959 Demineralized Water Storage & NO Not required for safe shutdown or Distribution containment isolation.

WBN2 IPEEE DESIGN REPORT Page 77 ATTACHMENT 1

Watts Bar Nuclear Plant Unit 2 IPEEE Seismic Report No: WBNIPEEE-003 Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List Revision 0 Page: 42 3.1 Main Steam System The main steam system (MSS) is designed to conduct steam from the steam generator outlets to the high pressure turbine or to the condenser turbine bypass (steam dump) system. This system also supplies steam to the feedwater pump turbines, auxiliary feedwater pump turbines, moisture separator reheater, and turbine seals. The MSS includes self-actuating safety valves (PORVs) to provide emergency pressure relief for steam generators and atmospheric relief valves to provide the means for plant cool down by steam discharge into the atmosphere if the turbine bypass system is not available.

The normal functions of the MSS can be summarized as follows: 1) Transport main steam from the steam generators to the final steam users, including the turbine generator and main feedwater pump turbines (MFPT), 2) Provide steam dump to the main condenser for control of nuclear steam supply system (NSSS) temperature and steam generator pressure during all phases of operation (startup, shutdown, and load rejection), and, 3) Drain condensate that accumulates in the main steam piping during initial heat up and normal operation.

For the SME, the MSS is required to provide isolation of the steam lines. This is accomplished with the closure of the main steam isolation valves (MSIVs), which are designated as WBN-2-FCV-00 1 -0004-T; WBN-2-FCV-00 1-0011 -T; WBN-2-FCV-00 1 -0022-T; WBN-2-FCV-00 1-0029-T. The MSIVs are 32-inch, failed-closed, air operated valves located in the main steam line downstream of the main steam safety valves and atmospheric relief valves. During normal plant operation, the MSIVs are held open by the pneumatic pressure supplied through the normally energized solenoid valves to the valve actuators.

The MSIV isolation signal will block the air supply to the MSIVs and open the air venting paths from the valve actuators.

On loss of pneumatic pressure, fast closure is achieved by a spring installed in the MSIV valve operator.Pressure transmitters:

WBN-2-PT-001-0002A-D; WBN-2-PT-001-0002B-E; WBN-2-PT-001-0009A-D; WBN-2-PT-001-0009B-E; WBN-2-PT-001-0020A-D; WBN-2-PT-001-0020B-E; WBN-2-PT-001-0027A-D; WBN-2-PT-001-0027B-E are located between the respective steam generator and its MSIV. MCR indication of the steam generator pressure is required for manual operation of the respective PORVs.WBN2 IPEEE DESIGN REPORT Page 78 ATTACHMENT 1

Watts Bar Nuclear Plant Unit 2 IPEEE Seismic Report No: WBNIPEEE-003 Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List Revision 0 Page: 43 3.2 Main Feedwater System The main feedwater (MFW) system is designed to supply feedwater to the steam generator secondary side during all normal operating conditions.

The MFW system pumps take suction from the main condenser hotwells and deliver water to the steam generators via the feedwater heaters.The main feedwater isolation valves, WBN-2-FCV-003-0033-A; WBN-2-FCV-003-0047-B; WBN-2-FCV-003-0087 -A; WBN-2-FCV

-003-0100-B are motor operated and receive power from a diesel backed Reactor MOV Boards. On receipt of a MFW isolation signal, the MFW system isolates so that main feedwater will not be delivered to the steam generators; this prevents reactor overcooling or possible over pressurization of containment due to feedwater line breaks inside containment.

For the SME evaluation in which a coincident loss of offsite power is assumed to occur, the MFW pumps lose power. The motor operated MFW isolation valves continue to receive power from the diesel backed Reactor MOV Boards and close upon receipt of a MFW isolation signal. The main and bypass regulation valves, located upstream of the MFW isolation valves, are not safety related and are installed in non-seismically qualified piping. Thus, only the MFW isolation valves are considered in the SSEL.3.3 Auxiliary Feedwater System The safety function of the auxiliary feedwater (AFW) system is to supply a sufficient feedwater flow to the steam generators to remove primary system stored and residual core energy in the event of a loss of main feedwater (MFW). The AFW may also be required to perform its safety function in other events, such as loss of off-site power, cool down after a loss of coolant accident for a small break LOCA, maintaining a water head in the steam generators following a loss of coolant accident, Main feed line, or main steam line breaks, and flood above plant grade.The system is designed to start automatically in the event that any of a number of events should occur which may result in, may be coincident with, or may be caused by a reactor trip. These include a loss of offsite electrical power, safety injection signal, low-low steam generator water level alarm, or a trip of both main feedwater pumps. It will supply sufficient feedwater to prevent the relief of primary coolant through the pressurizer safety valves and the uncovering of the core.It has adequate capacity to maintain the reactor at hot standby and then to cool the reactor coolant system to the temperature at which the residual heat removal system may be placed in operation, but it cannot supply sufficient feedwater for power generation.

WBN2 IPEEE DESIGN REPORT Page 79 ATTACHMENT I

Watts Bar Nuclear Plant Unit 2 IPEEE Seismic Report No: WBNIPEEE-003 Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List Revision 0 Page: 44 The AFW system consists of two motor-driven pumps (410-gpm rated flow each) and a turbine-driven pump (720-gpm rated flow). The turbine-driven pump and its associated valving, instrumentation, and piping was not considered in this seismic margins evaluation as it was determined that the motor-driven AFW pumps were the preferred equipment to be used by plant operators for the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> period of interest for the SMA. The AFW system has no normal power operation function.

However, it may be used during plant startup and normal plant cool down when the preferred supply from MFW is unavailable.

Motor-driven pump 2A-A (WBN-2-PMP

-003-0118-A), which delivers flow to steam generators 1 and 2, and pump 2B-B (WBN-2-PMP

-003-0128-B), which services steam generators 3 and 4, are aligned for auto start.The Unit 2 Condensate Storage Tank (CST) serves as the normal water supply for the AFW pumps with ERCW as an automatic backup source. The CST is not seismically qualified and is assumed to be unavailable following the SME. Thus, the suction pressure of the motor-driven pumps is expected to decrease as they begin operation.

The motor-operated ERCW supply valves servicing the motor-driven pumps (WBN-2-FCV-003-01 16A-A and WBN-2-FCV-003-0116B-A for pump 2A-A and WBN-2-FCV-003-0126A-B; WBN-2-FCV-003-0126B-B for pump 2B-B) open automatically when the pump is running and the suction pressure drops < 2.0 psig as sensed by two out of three pressure switches (WBN-2-PS-003-0139A-A, WBN-2-PS -003-0139B-A and WBN-2-PS-003-0139D-A for pump 2A-A, and WBN-2-PS-003-0144A-B, WBN-2-PS-003-0144B -B and WBN-2-PS-003-0144D-B for pump 2B-B) for 4 seconds.The essential raw cooling water (ERCW) system serves as an unlimited backup water supply to the AFW pumps. ERCW discharge header A supplies water to motor-driven pump 2A-A, and ERCW discharge header B supplies water to pump 2B-B. There are two motor-operated FCVs in series in each supply line from the main ERCW header to the pump suction (eight total). During normal operation, these valves are closed to isolate the AFW system from the low quality ERCW. Two motor operated valves each in train A and train B ERCW headers supply ERCW to the suction of the Turbine Driven Auxiliary Feedwater Pump. Two of these valves (WBN-2-FCV

-003-0136A

-A, WBN-2-FCV

-003-0179A-B) are listed in the SSEL for system isolation purposes since the turbine driven pump is not considered in the SSEL.The motor-driven pumps 2A-A and 2B-B start automatically in the event of a two out of three low-low level signal for any steam generator, a safety injection signal, a trip of both MFW pump turbines, or a loss of offsite power. The motor-driven pumps may also be manually started via hand switches in the main control room. Cooling for the motor-driven pump spaces is provided by WBN2 IPEEE DESIGN REPORT Page 80 ATTACHMENT 1

Watts Bar Nuclear Plant Unit 2 IPEEE Seismic Report No: WBNIPEEE-003 Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List Revision 0 Page: 45 the system 30 AB component cooling system (CCS)/AFW space coolers 2A-A and 2B-B (WBN-2-PMCL-030-0190 and WBN-2-PMCL-030-0191 respectively).

As the motor-driven pump breakers close to start the pumps, contacts off of the breakers open to allow the large LCVs in the steam generator inlet lines to open. These LCVs (WBN-2-LCV-003-0156 -A and WBN-2-LCV-003-0164 -A from pump 2A-A and WBN-2-LCV-003-0148-B and WBN-2-LCV-003-0171-B from pump 2B-B) are then modulated to a preset open position by their respective level control circuits.Portions of the Main feedwater System are isolated by containment isolation valves; WBN-2-FCV-003-0236, WBN-2-FCV-003-0239, WBN-2-FCV-003-0242, WBN-2-FCV-003-0245 and these valves are identified on the SSEL. The following check valves function as containment isolation valves, but are generically screened and are not included on the SSEL (Ref EPRI NP-6041-SL, R1, Table 2-4 & page 3-29): 2-CKV-3-0805, -0806.3.4 Chemical and Volume Control System The chemical and volume control system (CVCS) is designed to maintain the required water inventory and water chemistry control of the reactor coolant system (RCS). Specifically, the CVCS functions during normal operation to: 1) Maintain the programmed water level in the pressurizer.

2) Maintain seal water flow to the reactor coolant pumps.3) Control the reactor coolant water chemistry conditions, activity level, soluble chemical neutron absorber concentrations, and makeup.4) Degas the RCS.During emergency operation, the CVCS functions as part of the emergency core cooling system (ECCS) to provide high pressure safety injection from the refueling water storage tank (RWST)into the RCS. The CVCS is also used during certain events such as a faulted steam generator tube rupture (SGTR) or loss of sump recirculation to refill the refueling water storagetank.

The CVCS is a safety-related system designed to perform functions during normal operations and accident conditions.

The CVCS operates during all modes of operation:

During reactor startup, RCS pressure and temperature increase as the boron concentration is decreased from shutdown concentration so that criticality can be achieved.

During normal WBN2 IPEEE DESIGN REPORT Page 81 ATTACHMENT 1

Watts Bar Nuclear Plant Unit 2 IPEEE Seismic Report No: WBNIPEEE-003 Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List Revision 0 Page: 46 operation, the CVCS adjusts the RCS boron concentration to account for core burn up and the build up and decay of xenon during the core lifetime.

At hot shutdown, the RCS boron concentration is maintained such that the reactor is subcritical but can return to full power by withdrawing the control rods. During cold shutdown, the RCS boron concentration is increased to cold shutdown concentration, and the RCS pressure and temperature are decreased.

A portion of the CVCS is shared with the ECCS and is required for emergency shutdown following a LOCA.For the SME evaluation in which a coincident loss of offsite power is assumed to occur, the centrifugal charging pumps are used for normal makeup and for high pressure injection to achieve the alternate safe shutdown path of bleed and feed with an intact RCS. For the scenario with a small LOCA, the centrifugal charging pumps are used in both the preferred and alternate safe shutdown paths. For all SME shutdown paths, equipment necessary to maintain Reactor Coolant Pump seal flow is listed on the SSEL. Portions of the CVCS System are isolated by containment isolation valves; these valves are identified on the SSEL. The following check valves function as containment isolation valves, but are generically screened and are not included on the SSEL: 2-CKV-062-543, -0560,-061,-0562,-0563, and -0639.3.5 Safety Injection System The safety injection system (SIS) consists of two independent pump trains. The two pump trains discharge to a common header before splitting into four injection paths to provide flow to each of the four RCS cold legs. Separate injection paths are used for hot leg recirculation.

Miniflow recirculation is provided for the SIS pumps. The cold leg accumulators inject their contents of borated water into the RCS during intermediate or large break LOCA events. The SIS is used for safety injection during both the injection and sump recirculation phases.During normal plant operation, the SIS is in standby alignment for accident mitigation.

The SIS pumps are used to fill and top-off the cold leg accumulators (WBN-2-ACUM-063-0001, WBN-2-ACUM-063-0002, WBN-2-ACUM-063-0003, and WBN-2-ACUM-063-0004).

The normal standby lineup of the SIS is such that the pumps 2A-A and 2B-B (WBN-2-PMP

-063-0010-A, WBN-2-PMP-063-0015-B) are aligned to take suction from the RWST (WBN-2-TANK-063-0046) through the normally open inlet valve (WBN-2-FCV

-063-0005-B).

The train A path includes a normally open pump suction valve (WBN-2-FCV

-063-0047-A), safety injection pump 2A-A, a pump discharge check valve (WBN-2-CKV-063-0524-A), a locked-open pump discharge isolation valve (WBN-2-ISV-063-0525-A), and a normally open cold leg isolation valve (WBN-2-FCV -063-0152-A).

The train B path includes a normally open pump suction valve (WBN-2-FCV-063-0048-B), safety injection pump 2B-B, a pump discharge check valve (WBN-2-CKV

-063-WBN2 IPEEE DESIGN REPORT Page 82 ATTACHMENT 1

Watts Bar Nuclear Plant Unit 2 IPEEE Seismic Report No: WBNIPEEE-003 Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List Revision 0 Page: 47 0526-B), a locked-open pump discharge isolation valve (WBN-2-ISV-063-0527-B), and a normally open cold leg isolation valve (WBN-2-FCV-063-0153-B).

The two discharge paths join and flow through the normally open valve (WBN-2-FCV-063-0022-B), to the four RCS cold leg injection paths. Operation of SIS pumps requires cooling from the component cooling system (CCS) for the pump mechanical seals and cooling water supplied from the ERCW for the pump room coolers and the pump lube oil coolers. The pump room cooling fan is also required for successful operation of each SIS pump. Each SIS pump will start upon receipt of the "SI" signal.The injection phase of the SIS is initiated by an "SI" signal. The "SI" signal automatically starts the SIS pumps. The SIS pumps will recirculate water through the minimum flow line back to the RWST until RCS pressure falls below the shutoff head. As pressure decreases in the RCS, the SIS pumps will start injecting into the RCS. The SIS pump maximum injection flow rate is 650 gpm at 1808 feet of head.The recirculation phase is automatically initiated for ECCS when the RWST reaches low level with the containment sump high level and an "SI" signal present. In this phase, cooling water is pumped into the reactor vessel from the containment sump. Water collected in the sump is pumped by the RHR pumps, cooled by the RHR heat exchangers, and discharged to the RCS cold legs, the SIS and chemical and volume control system (CVCS) pump suction. The SIS and CVCS pumps inject the recirculated sump water into the reactor vessel when the RCS pressure remains above the shutoff head of the RHR pumps.Upon receipt of the required level signal, the containment sump isolation valves (WBN-2-FCV-063-0072 -A and WBN-2-FCV-063-0073-B) automatically open and the RHR normal suction valves (WBN-2-FCV-074-0003-A and WBN-2-FCV-074-0021-B) automatically close to isolate the RWST. The operator is instructed to complete the alignment of the SIS for the recirculation phase manually.

This is accomplished by closing the RWST isolation valve (WBN-2-FCV-063-0005-B), closing the SIS minimum flow valves (WBN-2-FCV-063-0003-A, WBN-2-FCV

-063-0004 -B and WBN-2-FCV-063-0175 -B), and opening the supply path from the RHR system to the SIS (WBN-2-FCV-063-001 1-B) and the flow path (WBN-2-FCV-063-0006-B, WBN-2-FCV-063-0007-A, WBN-2-FCV-063-0008-A).

Three hours after a LOCA starts, the operator is instructed to manually align the RHR and SIS for"hot leg" recirculation.

To initiate Hot Leg recirculation using the Safety Injection Pumps, the suction piping alignment remains the same as for cold leg injection, but each Safety Injection Pump discharge is aligned through separate headers that split to two branch lines and eventually to WBN2 IPEEE DESIGN REPORT Page 83 ATTACHMENT 1

Watts Bar Nuclear Plant Unit 2 IPEEE Seismic Report No: WBNIPEEE-003 Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List Revision 0 Page: 48 the hot legs. A throttling valve in each injection line is pre-set to limit Safety Injection Pump run-out and to equalize flow through the lines such that the amount of safety injection that spills to containment is minimized in the RCS loop that has ruptured.

The throttle valves are also adjusted to limit RHR pump run-out when one RHR pump is used to supply flow to the RHR Spray headers and to the Safety Injection Pumps suction.Placing the unit in the "hot leg" recirculation mode will provide long-term cooling for the reactor and prevent boric acid "plate out" in the core by reversing flow in the reactor and limiting the boron concentration.

This will prevent boric acid from blanketing the fuel rods, which may degrade the heat transfer from the core.The cold leg accumulators are pressurized to a minimum 585 psig by nitrogen gas and have a nominal volume of approximately 1355.59 ft 3 of borated water. Each accumulator injects its contents through a normally open, de-energized, motor-operated isolation valve and two check valves into the RCS cold leg through a 10-inch line. The four cold leg accumulators function independently of the rest of the SIS and inject into the RCS solely on the basis of pressure differential between the accumulators and the RCS.The accumulators are primarily designed to inject during LOCA conditions when RCS pressure rapidly decreases, and a large volume of water is needed to flood and cool the reactor in a relatively short period of time.Portions of the Safety-Injection System are isolated by containment isolation valves; these valves are identified on the SSEL. The following check valves function as containment isolation valves, but are generically screened and are not included on the SSEL: 2-CKV-063-0551,-0553,-0555,-

0557,-0581,-0632,-0633,-0634,-0635,-0640, & -0643.3.6 Essential Raw Cooling Water The primary function of the essential raw cooling water (ERCW) system (System 67) during normal operation is to provide cooling water to primary and secondary components such as component cooling system (CCS) heat exchangers, reactor coolant pump motors, control rod drive ventilation coolers, room coolers, and the air compressors.

During accident conditions, the ERCW system provides an ultimate heat sink function for dissipating heat from essential plant equipment, room ventilation systems, and the component cooling system. The ERCW system is also .the alternate water supply for the auxiliary feedwater system and the CCS surge tank.WBN2 IPEEE DESIGN REPORT Page 84 ATTACHMENT 1

Watts Bar Nuclear Plant Unit 2 IPEEE Seismic Report No: WBNIPEEE-003 Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List Revision 0 Page: 49 The ERCW system consists of an intake pumping station (IPS) structure, traveling screens, pumps, strainers, discharge overflow structure, valves and piping arranged in two trains, each of which has two supply headers to the various supplied equipment.

Those components of the ERCW System required to operate and/or maintain system integrity following a SME are included in the SSEL.3.7 Component Cooling System The component cooling system (CCS) acts as a barrier between the radioactive fluid flowing in the various coolers and the essential raw cooling water (ERCW) system to avoid release of radioactivity into the environment.

The function of the CCS is to serve as an intermediate heat conductor for the removal of heat from potentially radioactive heat loads during normal and accident conditions.

This function is accomplished through the use of a closed-loop system in which the CCS removes heat from the various component coolers (CCS loads) and transfers it to the CCS heat exchangers where the heat is transferred to the ERCW system. Those portions of the CCS required following a SME are included in the SSEL. Portions of this system are isolated following a LOCA; the isolation valves are included in the SSEL. Check valves 2-CKV-070-0679,-0687,-0698, & 0790 also function as containment isolation valves, but are generically screened and are not listed on the SSEL. Other components serviced by the CCS such as the Waste Gas Compressor and Spent Fuel Cooling Heat Exchangers are not specifically required for the SMA, but are listed on the SSEL for CCS pressure boundary integrity.

3.8 Residual

Heat Removal The RHR system is a safety-related system designed to perform functions during startup and cool down operations, shutdown operations, and during accident conditions.

The RHR consists of two independent pump trains in each unit. With the exception of the common piping described below, each loop is capable of performing the safety-related and normal operating functions of the system.Each loop consists of a pump, pump miniflow loop, a heat exchanger, and flow control and isolation valves. Both loops share a common heat exchanger bypass line, suction piping from the RCS, suction and discharge to the RWST.The normal functions of the RHR system are used during reactor startup, cool down, shutdown, and refueling.

These normal functions of the RHR are: 1) To transfer decay heat from the RCS to the component cooling system when the RCS pressure and temperature are below RHR system design conditions.

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Watts Bar Nuclear Plant Unit 2 IPEEE Seismic Report No: WBNIPEEE-003 Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List Revision 0 Page: 50 2) To maintain adequate RCS flow with the reactor coolant pumps off to ensure adequate chemical mixing, and, 3) To transfer refueling water between the RWST and the refueling cavity at the beginning and end of refueling operations.

The RHR system is designed to perform several safety functions during accident conditions:

1) Provide low pressure injection to the RCS, 2) Switch over the RHR suction from the RWST to the containment sump and provide suction to centrifugal charging pumps and SIS pumps, 3) Provide normal cool down for decay heat removal with suction from the RCS loop 4 hot leg, 4) Provide RHR spray as part of the CSS, and, 5) Provide hot leg recirculation.

On receipt of a safety injection signal, the RHR pumps start; the RWST to RHR pump flow control valve (WBN-2-FCV-074-0001-A), normally aligned open,. provides a suction path from the RWST; and normally open RHR heat exchanger outlet valves (WBN-2-FCV-074-0016 and WBN-2-FCV-074-0028) provide a discharge path to the four RCS cold-legs.

Miniflow valves (WBN-2-FCV-074-0012-A and WBN-2-FCV-074-0024-B) are opened or closed (depending on RHR injection flow into the RCS). When the RWST level is below the low level setpoint, the containment sump level is above the required setpoint and increasing, and the safety injection signal is present, the RHR supply valves automatically swap, and the recirculation mode begins.Switchover of the RHR pump suction to the containment sump from the RWST is required during an event when the RWST level is below the low level setpoint and the containment sump level is above the required setpoint.

The RHR recirculation mode begins with the automatic opening of the containment sump supply valves (WBN-2-FCV-063-0072-A and WBN-2-FCV-063-0073-B)

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Watts Bar Nuclear Plant Unit 2 IPEEE Seismic Report No: WBNIPEEE-003 Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List Revision 0 Page: 51 and closing the RWST supply valves (WBN-2-FCV-074-0003-A and WBN-2-FCV-074-0021-B).

If automatic switchover does not occur, the operators are instructed to complete the switchover manually.RHR supply to CVCS and safety injection valves (WBN-2-FCV-063-0008-A and WBN-2-FCV-063-001 1-B) are manually opened to establish a flow path from the containment sump through the RHR pumps and heat exchangers to the suction of the centrifugal charging pumps and safety injection pumps for high pressure recirculation.

The two RHR crosstie valves (WBN-2-FCV-074-0033-A and WBN-2-FCV-074-0035-B) are manually closed during the recirculation mode to separate the RHR train A and B flow paths for protection against a passive failure.The RHR system can be used in the recirculation mode to supply part of its flow to the one of two parallel RHR containment spray headers. One of the RHR spray header isolation valves (WBN-2-FCV-072-0040-A and WBN-2-FCV-072-0041-B) must be opened by the operators to establish the flow path from the RHR pumps to the headers.3.9 Reactor Protection System The reactor protection system (RPS) provides (1) automatic protection against unsafe reactor operation and (2) initiating signals to mitigate the possible consequences of faulted conditions.

The RPS is composed of two major systems: the reactor trip system (RTS) and the engineered safety features actuation system (ESFAS). The function of the RTS is to ensure that the reactor operates within established safe operating limits. The ESFAS is provided to sense accident situations and to initiate the operation of necessary engineered safety features (ESF).Reactor Trip System The RTS automatically trips the reactor whenever the safe operating limits are about to be breached.

The general safety limits relating to the necessity of tripping the reactor are;1) High power, 2) Excessive reactor coolant temperature, 3) Low coolant pressure, and 4) A combination of these parameters.

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Watts Bar Nuclear Plant Unit 2 IPEEE Seismic Report No: WBNIPEEE-003 Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List Revision 0 Page: 52 The RTS functions by interrupting power to the rod control system, causing the control rods to drop into the core. The two reactor trip breakers associated with each unit, RTA and RTB, are arranged in series between the control rod motor generator set switchgear and the control rod drive power cabinets.

A breaker can be tripped electrically by either of two methods;1) Operation of the under voltage trip attachment (UVTA), and/or, 2) The shunt trip attachment (STA).Engineered Safety Features Actuation System (ESFAS)ESFAS senses selected plant parameters, determines whether established safety limits are being approached and, if they are, combines the signals into logic matrices that are sensitive to combinations indicative of a primary or secondary system boundary rupture.The ESFAS is composed of the following systems: 1) Process and Control Instrumentation Protection Set Racks (EAGLE 21)2) Solid State Protection System (SSPS)3) ESF Test Cabinets 4) Manual Actuation Circuits The process protection system and the SSPS are designed to accomplish the following:

1) Generate all necessary process protection signals, combine them into logic matrices, and initiate a reactor trip or actuate necessary ESF equipment.
2) Maintain physical and electrical separation by providing four sets of process and control instrumentation protection system (EAGLE 21) cabinets and two sets of SSPS cabinets (racks), one for each protection train (A and B).Process Protection System (Eagle 21)The EAGLE 21 process protection system is a microprocessor-based system housed in 14 racks, in four cabinets which are divided into 4 protection channel sets. In each protection channel the process electronics power the sensors and perform signal conditioning, calculation, and isolation operations on the input signals. The analog input module of the system powers the field sensor(s)WBN2 IPEEE DESIGN REPORT Page 88 ATTACHMENT 1

and performs signal conditioning.

All calculations for the process channel functions are performed by the loop calculation processor (LCP), and channel trip signals are provided through the partial trip output boards to the protection logic circuits of the SSPS.Solid State Protection System The SSPS is a dual-train, redundant protection system housed in two 3-bay cabinets, one single bay control board demultiplexer cabinet, and a computer-monitored demultiplexer assembly.

Each 3-bay cabinet contains an input relay bay, a logic bay, and an output relay bay. The inputs are transmitted through AC-operated relays that separate SSPS logic circuits from the protection set inputs. The output relays consist of master and slave relays with the slave relays driven by the master relays.When a transient occurs, various signals are generated, depending on the event that initiated the transient.

These signals provide actuation for the equipment that is expected to operate automatically during the transient.

Manual actuation of the ESFAS signals is also available.

3.10 ELECTRIC POWER SYSTEM The electric power system at the Watts Bar Nuclear Plant consists of the unit main generator, three unit station service transformers (USST) per unit, four common station service transformers (CSST), four diesel generators, the station and vital batteries, and a two-train electrical distribution system. The electric power system can be broken into six subsystems:

offsite grid, 6.9-kV AC power and diesel generators, 480V AC power, 250V DC power, 125V DC, and 120V AC power.During normal plant operation, the unit main generators supply electric power through the USSTs to the non-safety auxiliary power system. Offsite electric power supplies Class IE circuits through the 1.61 -Kv system via the CSSTs. The normal power operation alignment of each subsystem is described below.Offsite Grid Two offsite power grids are connected to the Watts Bar Nuclear Plant: a 500-kV grid via a 500-kV switchyard and a 161-kV grid via the 161-kV switchyard.

The 500-kV grid is supplied power from the Unit 2 main generator during normal power operation.

When a unit trip occurs, the unit is separated from the 500-kV grid, and all offsite power is supplied by the 161-kV grid. The Watts WBN2 IPEEE DESIGN REPORT Page 89 ATTACHMENT 1

Watts Bar Nuclear Plant Unit 2 IPEEE Seismic Report No: WBNIPEEE-003 Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List Revision 0 Page: 54 Bar Hydro Plant switchyard supplies the Watts Bar Nuclear Plant 161-kV transformer yard. Two separate transmission lines provides 161 kV power to the 161 -6.9 kV CSSTs A, B, C, and D.6.9-kV AC Power and Diesel Generators During power operation, the 6.9-kV unit boards receive power from the unit main generators via the USSTs. When a unit trip occurs the unit boards will transfer to the start buses for power. The 6.9 kV start buses are powered by CSSTs A and.B via the 161 kV transformer yard.The 6.9 kV shutdown boards are powered by CSSTs C and D via the 161 kV transformer yard.Upon loss of offsite power (161 kV power from Watts Bar Hydro Plant), the 6.9 kV shutdown boards will be powered from the emergency diesel generators.

Following the postulated SME, offsite power is assumed to be lost and unavailable for the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> period of interest.

Therefore, electrical power for the operation of safeguards equipment must be supplied by the onsite emergency power source; the diesel generators.

Specifically, the diesel generators are needed to provide electrical power to the 6.9-kV shutdown boards for the operation of pumps, valves, and instrumentation in the fluid mechanical systems used to keep the plant in a safe stable shutdown condition.

The 250V DC system supplies control power to the 6.9-kV Unit Boards. Although these boards are required to re power the 6.9-kV Shutdown Boards from the grid (should the grid be regained)via the unit boards, they are not required for the plant to achieve or maintain a safe stable shutdown condition.

Thus, 250V DC system components are not included on the SSEL.WBN has four diesel generators, with each diesel generator unit supplying a single safeguards bus.Each diesel generator unit includes the following equipment:

1) A diesel generator, 2) Sequencer, 3) Dedicated DC control power, 4) Diesel room ventilation, 5) Essential raw cooling water (ERCW) cooling systems, 6) Output breaker, and, 7) Associated piping and valves.WBN2 IPEEE DESIGN REPORT Page 90 ATTACHMENT 1

Watts Bar Nuclear Plant Unit 2 IPEEE Seismic Report No: WBNIPEEE-003 Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List Revision 0 Page: 55 The Watts Bar diesel generators are each a single generator with a 16-cylinder engine mounted on each end connected to a single shaft. The diesel is started by an air start system unique to each diesel. At WBN, the diesel generators are skid-mounted package units with most essential auxiliary components furnished as part of the skid. This includes such components as the starting air receiver tanks and motors, fuel day tanks and fuel oil priming pumps, and cooling water heat exchangers.

The intake air supply silencers, air cleaners and the exhaust silencers are not part of the skid and are located in a separate room on the floor above the diesel generator rooms.Combustion air supply is hard piped from the air cleaner in the intake room through the generator room to the engine.The standby diesel generator (DG) system serves as the plant emergency standby alternating current power source. Each DG is capable of starting and accelerating to rated speed within 10 seconds to provide power to the needed engineering safety features and shutdown loads. The DG system is required to operate under each of, or, any combination of the following events: 1. Loss of Off-site Power (LOOP)2. Degraded voltage on the 6.9-kV shutdown boards 3. Safety Injection (SI) signal The DG starts automatically when a sustained (longer than 1.5 seconds) loss of voltage on the 6.9-kV Shut Down board occurs. After an additional time delay of 3 seconds at zero volts, all 6.9-kV Shutdown Boards loads (except the 6.9-kV to 480 volt transformers) and major 480 Vac loads are tripped. The DG is automatically connected to the 6.0-kV Shut down-Board after it reaches rated speed and voltage. The return of voltage to the board initiates logic to connect the required loads in sequence.

Such an automatic start signal operates a lock-out relay that removes all manually, electrically, and mechanically-operated stop signals except emergency stop, over-speed trip and generator differential trip. Wiring features a transformer, resistor and relay assembly used as a neutral grounding device to limit ground fault current and provide relay contacts for annunciation in the main and auxiliary control room.The fuel oil transfer system for the diesels consists of the following equipment:

1) A 62,000-gallon minimum fuel oil supply, 2) Fuel oil transfer pumps 1 and 2, and, 3) Associated piping, valves, and instrumentation.

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Watts Bar Nuclear Plant Unit 2 IPEEE Seismic Report No: WBNIPEEE-003 Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List Revision 0 Page: 56 The diesel generators consume 100 pounds of fuel in 2:28 minutes at a load of 5,000 kW. The fuel transfer pumps are started at different levels by level switches such that if the first pump does not start, the level will drop, and the second will start. Both fuel transfer pumps will stop as the day tank level reaches a "high level" and activates a level switch.Ventilation and cooling for the Unit 1 shutdown boards is supplied by air handling Unit lA-A and Unit lB-B; Unit 2 boards are cooled by air handling Unit 2A-A and Unit 2B-B (WBN-0-AHU

-031-0044, WBN-0-AHU

-031-0045, WBN-0-AHU,-031-0055, WBN-0-AHU

-031-0061).

During normal operation, one air handling unit is in operation, while the other is in standby.480V AC Power The 480V AC power subsystems receive power from their associated 6.9-kV AC power shutdown boards. 6.9-kV board lA-A (WBN-1-BD-211-A-A) supplies 480V boards 1A1-A and 1A2-A (WBN-1-BD-212-A001-A and WBN-1-BD-212-A002-A).

6.9-kV board 1B-B supplies 480V boards 1B1-B and 1B2-B. Similarly, Unit 2 480V AC power boards 2A1-A, 2A2-A, 2B1-B, and 2B2-B (WBN-2-BD-212-A00 l-A, WBN-2-BD-212-A002-A, WBN-2-BD-212-B00 1-B, WBN-2-BD-212-B002-B) receive power from their associated 6.9-kV shutdown boards. Namely, 2A1-A and 2A2-A receive power from 6.9-kV shutdown board 2A-A (WBN-2-BD-21 1-A-A), and 2B 1-B and 2B2-B receive power from 6.9-kV shutdown board 2B-B (WBN-2-BD-21 1-B-B). The 480V AC auxiliary building common boards are supplied from the 6.9-kV common boards.480V AC transformer room ventilation is supplied by four fans each for rooms 1A and 2B, and by three fans each for rooms 2A and lB. Each fan is activated by a temperature switch that is set differently so that each fan is started on a staggered basis as the room temperature rises. The ventilation system in each room includes two motor-operated inlet dampers. Each fan has a mechanical back draft damper to prevent reverse flow while the fan is not in operation.

125V DC/120V AC Power Each of the four 125V DC Vital Battery channels consists of the following:

1) One 125V DC Vital Battery, 2) One Vital Battery charger, and, 3) One Vital Battery Board with associated breakers and fuses.WBN2 IPEEE DESIGN REPORT Page 92 ATTACHMENT 1

Watts Bar Nuclear Plant Unit 2 IPEEE Seismic Report No: WBNIPEEE-003 Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List Revision 0 Page: 57 During normal operation, the battery chargers supply all required DC. Vital Battery chargers 1A, IB, 2A, and 2B (WBN-0-CHGR-236-0001-D, WBN-0-CHGR-236-0002-E, WBN-0-CHGR-236-0003-F and WBN-0-CHGR-236-0004-G) are normally powered from 480V shutdown boards 1A1-A, 1B2-B, 2A1-A, and 2B2-B, respectively.

The 125V DC battery is maintained on a float charge by virtue of its connection to the bus and acts as an emergency DC supply should the charger fail. There are two spare chargers (WBN-0-CHGR-236-0006-S and WBN-0-CHGR-236-0007-S), one in each transfer rack.There is a fifth Vital Battery (WBN-0-BAT-236-0005-S) that may be substituted for any of the other four Vital Batteries.

Each of the eight 120V Vital AC Instrument Power Systems consists of the following:

1) One Vital Inverter, 2) One Vital Instrument Power Board with associated breakers, and, 3) An alternate supply for the Vital Instrument Power Board from an Instrument Power Transformer supplied from a 480V Shutdown Board.During normal operation, the inverter provides power to the 120V AC vital instrument bus. The inverter is normally supplied by the associated 480V shutdown board. Should all AC power be lost, the inverter will be directly supplied from its associated 125V DC battery board.The 120V AC instrument power buses 1A consists of: 1) Instrument Power Transformer 1 A, 2) Instrument Power Distribution Panel 1 A with associated breakers, 3) A fused disconnect switch, and, 4) A 480V manual transfer switch.The transformer is normally powered from 480V shutdown board 1Al-A with a non automatic alternate supply from 480V shutdown board 1B1-B. The transformer reduces the voltage from 480V to 120V for use in instrumentation.

The 120V AC instrument power buses lB, 2A, and 2B are similarly configured.

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Watts Bar Nuclear Plant Unit 2 IPEEE Seismic Report No: WBNIPEEE-003 Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List Revision 0 Page: 58 The 120V AC instrument power system requires ventilation for operation.

The 480V board room ventilation system supplies rooms lB and 2B in which the 120V AC inverters reside. One air handling unit (AHU) supplies each room; and there is no cross connect of ducting.Accident/Transient Operation The following describes the expected operation of the electric power subsystems for the loss of offsite power (LOOP) assumed coincident with the SME.1) The offside grid is assumed to be lost. The equipment affected includes the 500-kV switchyard, the 161-kV transformer yard, the CSSTs (A, B, C and D), the start and shutdown buses, and their associated secondary-side breakers.2) With a loss of offsite power, the unit boards and the 6.9-kV shutdown boards fail to receive power from the CSSTs; the start buses are de-energized.

The diesel generators will start and supply power to their associated 6.9-kV shutdown board.On a loss of offsite power, after the diesel generators are started and loaded to the 6.9-kV shutdown boards, the 480V shutdown boards are then re-energized.

The ventilation system is then restarted to serve the shutdown board rooms.3) The 480V AC power subsystem is further divided into Class 1E power and balance-of-plant (BOP) power. Train A equipment includes 480V shutdown transformer 1A-A and 2A-A 480V shutdown board lA-A and 2A-A, reactor motor-operated valve board 1A-A and 2A-A,diesel auxiliary board lA-A and 2A-A, reactor ventilation board lA-A and 2A-A, control and auxiliary building ventilation board lA-A and 2A-A, and their associated loads. Alternate power can be supplied to 480V Shutdown Board lA-A and 2A-A via Shutdown Transformer 1Al-A and 2A1-A, but the transfer is manual. Train B equipment is similarly configured.

With the loss of offsite power (LOOP), the 480V Shutdown Boards continue to be powered from their associated 6.9-kV shutdown boards. Several Class 1E boards, however, are shed from the 480V shutdown boards on a loss of offsite power.These boards include all four reactor ventilation boards and control and auxiliary building ventilation boards 1A2-A, 1B2-B, 2A2-A, and 2B2-B.4) The 120V AC Vital Instrument Power boards 1-I, 1-I, 1-III, and 1 -IV are normally supplied from the 480V AC shutdown boards lA1-A, 1B2-B, 2A2-A and 2B2-B, WBN2 IPEEE DESIGN REPORT Page 94 ATTACHMENT I

respectively, via a vital inverter.

Following the LOOP, power continues to be supplied from the 480V AC bus (the diesel generators supply the 480V AC shutdown boards via the 6.9-kV AC boards). Also, for the LOOP, the AHUs in the 480V board rooms will restart after power is restored to the shutdown buses by the diesel generators.

Ventilation is required by the 120V AC instrument power system; the 480V AC board room ventilation system supplies rooms 1B and 2B;these rooms hold the 120V AC inverters.

One AHU supplies each room, and there is no cross connecting of duct work.3.11 MISCELLANEOUS SYSTEMS In addition to the systems described in Sections 3.1 through and including 3.10, equipment from several other systems are included in the SSEL. These systems are described briefly in this section.High Pressure Fire Protection (System 26)This system provides automatic fire suppression (water spray) to the reactor coolant pumps (RCPs)in the event of an RCP fire. There are two motor operated valves associated with fire suppression for the reactor coolant pumps; WBN-2-FCV-026-0240-A which feeds the spray headers positioned about the reactor coolant pumps, and WBN-2-FCV-026-0243-A which supplies the associated stand pipe. The valves are powered by 480 V Reactor MOV Board BD 2A1-A. This system is automatically isolated during containment isolation.

Thus, the valves required to perform the containment isolation safety function are included on the SSEL.Check valves 1-CKV-026-1296 and -1260 also function as isolation valves, but are generically screened and are not included on the SSEL.Ventilation (System 30)This system is required for containment isolation, containment lower compartment environmental circulation and to provide airflow to certain pumps and electrical boards that are required to operate following the SME. The ventilation equipment required to perform both its containment isolation function for the small LOCA scenario, long term cooling post accident and its support system function for both the intact RCS and the small LOCA scenarios are included in the SSEL.WBN2 IPEEE DESIGN REPORT Page 95 ATTACHMENT 1

Air Conditioning (System 31)The air conditioning system provides cooling for essential electrical equipment and pumps that are required to operate following the SME. Those components (air handlers, chillers, etc.) required to function to perform this task are included 'in the SSEL.Control Air (System 32)The Essential Air portion of the Control Air System provides pneumatic power to valves required to perform front line safety functions following a SME. Those components required to function to accomplish this task are included in the SSEL. Check valves 2-CKV-032-0323,-0333-

& -0343 function as containment isolation valves, but are generically screened and are not listed on the SSEL.Post Accident Sampling (System 43)The Post Accident Sampling System is isolated following receipt of a containment isolation signal.The valves that perform this function are included in the SSEL. Check valves 2-CKV-043-0834,-

0841,-0883, & -0884 function as containment isolation valves, but are generically screened and are not listed on the SSEL.Ice Condenser (System 61)The glycol lines to the chillers inside the ice condenser are isolated following receipt of a containment isolation signal. The valves that perform the containment isolation function are included in the SSEL. Check valves 2-CKV-061-0692,-0680,-0745, & -0533 function as containment isolation valves, but are generically screened and are not listed on the SSEL.Also, the ice condenser doors and ice condenser itself is required for cases when primary coolant is released to the containment and is therefore also included in the SSEL. For a small break LOCA, only the lower compartment doors are required to function and are included on the SSEL. Flow to the upper compartment bypasses the Intermediate deck doors through the vent curtain. Since the ice condenser baskets are passive components, they are not included on the SSEL.WBN2 IPEEE DESIGN REPORT Page 96 ATTACHMENT 1

Watts Bar Nuclear Plant Unit 2 IPEEE Seismic Report No: WBNIPEEE-003 Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List Revision 0 Page: 61 Reactor Coolant System (System 68)The reactor coolant system (RCS) components, i.e. reactor vessel, steam generators, pressurizer, and accumulators etc., are included in the SSEL for evaluation of their ability to perform their intended function.Containment Spray (System 72)The primary safety function of the containment spray system is to remove thermal energy from the containment in the event of a LOCA or a steam line break inside containment.

The heat removable capability of the spray system assists in maintaining containment integrity when steam generated in the core continues to enter containment by keeping the containment pressure below its design pressure after all the ice in the containment ice condenser has melted. Those components of the CS system required to accomplish this function are listed on the SSEL.Check valves 2-CKV-072-548,-0549,-0562, & -0563 function as containment isolation valves, but are generically screened and are not listed on the SSEL.Waste Disposal (System 77)During normal operation, one function of the Waste Disposal system is to collect contaminated liquid waste from inside containment and duct it to holding tanks located in the Auxiliary Building.In the event of containment isolation, the Waste Disposal system is isolated from containment.

Thus, those valves required to perform this isolation function are listed on the SSEL. The valves are closed on a Phase A Containment Isolation Signal. Power is supplied by 125 VDC Battery Boards.Check valves 2-CKV-077-0849 and 2-CKV-077-0862 function as containment isolation valves, but are generically screened and are not included on the SSEL.Spent Fuel Cooling (System 78)The spent Fuel Cooling components are listed for CCS pressure boundary integrity only. The spent fuel cooling function is not considered as part of the SMA.WBN2 IPEEE DESIGN REPORT Page 97 ATTACHMENT 1

Watts Bar Nuclear Plant Unit 2 IPEEE Seismic Report No: WBNIPEEE-003 Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List Revision 0 Page: 62 Primary Makeup Water (System 081)This system provides containment isolation of the reactor coolant pump standpipe seal water. The valve that performs the containment isolation function is included in the SSEL. Check valve 2-CKV-081-0502 functions as a containment isolation valve, but is generically screened and is not included on the SSEL.Containment Isolation (System 88)Containment isolation valves are not assigned system 88 identification number; they carry the component identification of the system in which they are located. Consequently, there are no system 88 identification numbers listed in the SSEL.Radiation Monitoring (System 90)The Containment Upper and Lower Compartment Area Radiation Monitors monitor radiation levels inside containment post accident; these monitors are listed in the SSEL.The Containment Upper and Lower Compartment Airborne Radiation Monitors monitor radiation levels inside containment during normal operation.

In the event of containment isolation, these monitors are isolated from containment.

The valves required to perform this isolation function are listed on the SSEL.Neutron Monitoring System (System 92)Specific components (instruments) of the Neutron Monitoring System required to be operable post-accident to monitor reactivity associated with the core are included on the SSEL. Those components not in the main control room are included in the SSEL.Incore Instrumentation System (System 94)The Incore Instrumentation System does not perform any safety function; therefore, it is not included in the SSEL. It should be noted that seismic restraints for the movable frame assembly above the seal table have been provided.

However, since these restraints are only removed during refueling, they are not required to be evaluated by this program.WBN2 IPEEE DESIGN REPORT Page 98 ATTACHMENT 1

Watts Bar Nuclear Plant Unit 2 IPEEE Seismic Report No: WBNIPEEE-003 Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List Revision 0 Page: 63 Cranes (System 271)The polar crane located inside containment is a large, heavy structure that, should it become dislodged from its track could fall and damage other critical equipment inside containment required to perform safety functions following a SME. Thus, the polar crane is included on the SSEL for evaluation of position retention capability only. For the same reasons, the crane located over the refueling floor in the Auxiliary Building is included in the SSEL.System 275 & 276 Racks and Panels The racks and panels listed were identified during the control-power interface review and are classified as Auxiliary Relay Racks (ARR) and trained BOP instrument racks. The relay racks (ARR) provide the necessary isolation and separation between the process signals and safety circuits.

The following criteria apply to these circuits: 1. A safety signal derived from the Solid-State Protection System overrides the process signal.2. The isolation relays have a coil to contact rating equal to or greater than the maximum credible ac or dc potential that could be applied to the non-lE circuit as its end points or intermediate routing.3. The isolation relays and racks designated as Train A or Train B be seismically qualified.

Each relay rack is included in the relay evaluation.

Both the 1E and non-lE racks were reviewed for the application of seismically susceptible relays in control circuits.Junction Boxes (Systems 290, 292 and 293)Junction boxes in which equipment, not readily identifiable to a process system, is mounted are included in the SSEL. The junction boxes have unique component identifiers with conduit system numbers 290, 292 and 293. This system number identifies the building in which the box is located. Since junction boxes are usually considered bulk commodities, they are generally not within the SMA scope. Accordingly, these systems have not been included in the Systems Summary List (Table 3.1). Each component is evaluated with its related SMA system.WBN2 IPEEE DESIGN REPORT Page 99 ATTACHMENT 1

Watts Bar Nuclear Plant Unit 2 IPEEE Seismic Report No: WBNIPEEE-003 Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List Revision 0 Page: 64 4.0 SAFE SHUTDOWN EQUIPMENT LIST This section describes the approach taken to develop the Safe Shutdown Equipment List (SSEL), provides a key to the reading of the SSEL, and provides the SSEL itself.4.1 Approach The approach taken was to minimize the amount of equipment to be included on the SSEL itself, yet maintain high reliability by listing redundant components.

To accomplish this, the following guidelines were employed: Equipment Included on SSEL 1) Reactor Coolant System Boundary Components, such as the reactor vessel, the steam generators, pressurizer and accumulators.

2) Active valves, those that are required to change position to successfully accomplish the Safe Shutdown Path(s). Examples of these would be containment isolation valves. Another example would be RHR and SI valves that may be required to change position as a result of realigning from cold leg recirculation to hot leg recirculation.
3) Active components such as those that are required to operate to successfully accomplish the Safe Shutdown Path(s). These include the ERCW pumps, RHR pumps, motor-driven charging pumps, air compressors and related equipment required for essential air, air handling units required to cool pump rooms, tanks, and heat exchangers.
4) Instrumentation required by the operators to monitor parameters associated with the key long term safety functions.
5) Electrical equipment required to power required pumps, valves, air handlers and instrumentation required to accomplish the Safe Shutdown Path(s).6) Major system passive components necessary to maintain system pressure boundary, such as tanks, filters, strainers, etc.Equipment Excluded from SSEL Similarly, the following guide lines were used to exclude equipment from the SSEL: 1) Check valves. Generically screened out due to seismic ruggedness.

WBN2 IPEEE DESIGN REPORT Page 100 ATTACHMENT 1

Watts Bar Nuclear Plant Unit 2 IPEEE Seismic Report No: WBNIPEEE-003 Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List Revision 0 Page: 65 2) Valves in a required system that do not change state (move). These include valves that are locked open, locked closed, open with power removed, close with power removed, open with air removed and close with air removed.3) Electrical systems not required either to achieve or maintain the plant in its safe, stable shutdown condition for the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> period of interest for the SME. An example of this is the 250 Volt DC battery system.4) Components downstream of their isolation valves for systems not required to operate in order to achieve or maintain the safe shutdown condition.

5) Similarly, non-essential components beyond the isolation valves when only a portion of the system is required to operate successfully to achieve the safe shutdown state.An example of this is the boric acid system associated with the CVCS.The guidelines listed above were applied to determine the equipment and components to be included on the SSEL.4.2 Definitions of Table Entries Table 4-1 lists the key for use in interpreting the critical equipment lists developed based on the SPLDs shown in Figures 2-1 and 2-2, the dependency matrices given in Tables 2.2 and 2.3 and the detailed list of references given in Table 2.1. This key provides the definitions for and explanations of the ten entries made for the critical equipment list developed under the Systems Evaluation portion of the SME.4.3 Safe Shutdown Equipment List The SSEL provided in Appendix A includes equipment unique to Unit 2, Unit 1 equipment required for Unit 2 operation, and equipment designated as common to Unit 1 and Unit 2.WBN2 IPEEE DESIGN REPORT Page 101 ATTACHMENT 1

Watts Bar Nuclear Plant Unit 2 IPEEE Seismic Report No: WBNIPEEE-003 Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List Revision 0 Page: 66 Table 4.1 Entry (Column)KEY FOR READING SAFE SHUTDOWN EQUIPMENT LIST Description of Entry Equipment ID = UNID: UJNnique Equipment IDentifier Plant Unit Function System Address/Sub Train Where Plant = WBN (Watts Bar Nuclear) (Three Characters)

Unit = 0 -Common (One Character) 1 Unit I 2 Unit 2 Function Function Code Defined in Table 4.1 Column 8 (Four Characters).

System = is the three digit System Number as identified in Table 3.1(Three Characters).

Address / Sub address = are two fields of (four character) given to number each type of component.

Train = A single alpha character given to safety related components.

Equipment Description

-Description of Equipment Flow diagram -Drawing depicting equipment functional arrangement 2 3 WBN2 IPEEE DESIGN REPORT Page 102 ATTACHMENT 1

Watts Bar Nuclear Plant Unit 2 IPEEE Seismic Report No: WBNIPEEE-003 Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List Revision 0 Page: 67 (Column) Description of Entry 4 Path -Success Path to which the Component is associate; 1 -Intact RCS, Aux Feed Available 2 -Intact RCS, Aux Feed unavailable (alternate success path uses RCS bleed and feed)3 -Small LOCA, Aux Feed available and use Safety Injection 4 -Small LOCA, Aux Feed unavailable (alternate success path uses RCS bleed and feed)5 -Component is used in all (preferred and alternate) success paths.5 Func -Success Path Function Performed By Equipment A -Support System B -Reactivity Control C -RCS Pressure Control D -RCS Inventory Control E -Decay Heat Removal F Containment Isolation G -System Isolation H -Room or Area Cooling (HVAC)I -Containment Pressure / Temperature Control J -Combination of Above 6 Op Nor -The operational status of the component during normal operation.

WBN2 IPEEE DESIGN REPORT Page 103 ATTACHMENT 1

(Column) Description of Entry 7 Op Des -The desired status of the component following the SME event Entry Definition On -On Off -Off O -Open C -Closed V -Variable NR -Not Required T -Throttled-- Non-active, i.e. tank WBN2 IPEEE DESIGN REPORT Page 104 ATTACHMENT 1

Watts Bar Nuclear Plant Unit 2 IPEEE Seismic Report No: WBNIPEEE-003 Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List Revision 0 Page: 69 (Column) Description of Entry 8 Function Code Definition ACUM ACCUMULATOR AHU AIR HANDLING UNIT ARB BOARD, RELAY, AUXILIARY BAT BATTERY BD SWITCHBOARD BKR BREAKER, CIRCUIT CDPL PANEL, DISTRIBUTION, CONTROL CHGR CHARGER, BATTERY CHR CHILLER CLR COOLER COMP COMPRESSOR COND CONDENSER CRN CRANE DEMN DEMINERALIZER DIEG ENGINE, DIESEL DOOR DOOR DPL PANEL, DISTRIBUTION DRYR DRYER, AIR DXF TRANSFORMER, DRY-TYPE FAN FAN FCO VALVE OPERATOR, FLOW CONTROL FCV VALVE, CONTROL, FLOW FE ELEMENT, FLOW FIS SWITCH, INDICATING, FLOW FLTR FILTER FS SWITCH, FLOW FSV VALVE, SOLENOID, FLOW FT TRANSMITTER, FLOW GEN GENERATOR HIC CONTROLLER, INDICATING, HAND HS SWITCH, HAND HTX HEAT EXHANGER IACL CLEANER, AIR INV INVERTER ISV VALVE, ISOLATION WBN2 IPEEE DESIGN REPORT Page 105 ATTIACHMENT 1

Watts Bar Nuclear Plant Unit 2 IPEEE Seismic Report No: WBNIPEEE-003 Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List Revision 0 Page: 70 (Column) Description of Entry 8 (Cont.) Function Code Definitions JB BOX, JUNCTION, TVA LCV VALVE, CONTROL, LEVEL LSV VALVE, SOLENOID, LEVEL LT TRANSMITTER, LEVEL MCC CONTROL CENTER, MOTOR NM MODIFIER, NEUTRON OXF TRANSFORMER, OIL-FILLED PCV VALVE, CONTROL, PRESSURE SWITCH, INDICATING, DIFFERENTIAL, PDIS PRESSURE TRANSMITTER, DIFFERENTIAL, PDT PRESSURE PMCL COOLER, PUMP PMP PUMP PNL PANEL PRES PRESSURIZER PS SWITCH, PRESSURE PSV VALVE, SOLENOID, PRESSURE PT TRANSMITTER, PRESSURE RCVR RECEIVER RE ELEMENT, RADIATION RFV VALVE, RELIEF ROD ROD, CONTROL RPV VESSEL, REACTOR PRESSURE SEP SEPARATOR SFV VALVE, SAFETY SGEN GENERATOR, STEAM SILN SILENCER STN STRAINER TANK TANK TC CONTROLLER, TEMPERATURE TCV VALVE, CONTROL, TEMPERATURE TE ELEMENT, TEMPERATURE TIS SWITCH, INDICATING, TEMPERATURE TS SWITCH, TEMPERATURE TSV VALVE, SOLENOID, TEMPERATURE TT TRANSMITTER, TEMPERATURE WBN2 IPEEE DESIGN REPORT Page 106 AI-IACHMENT 1

Watts Bar Nuclear Plant Unit 2 IPEEE Seismic Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List Revision 0 Report No: WBNIPEEE-003 Page: 71 TWS SCREEN, TRAVELING WATER Descrintion of Entry (Column)8 (Cont.) Function Code Definitions xSW SWITCH, TRANSFER 9 Pwr -Power (either electrical or air) required to achieve the desired status: Entry Y N or NA Description

-Yes-No 10 11 Seism Cat Seismic Category Floor elevation and etc.) of equipment Flr Elev/Room

-location (room, azimuth, column lines, WBN2 IPEEE DESIGN REPORT Page 107 ATTACHMENT 1

Watts Bar Nuclear Plant Unit 2 IPEEE Seismic Report No: WBNIPEEE-003 Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List Revision 0 Page: 72 5.0 Peer Review To be performed as part of verification

/ validation.

WBN2 IPEEE DESIGN REPORT Page !108 ATTACHMENT 1

6.0

SUMMARY

This report documents the following:

1) The objectives of the systems evaluation portion of the seismic margins assessment being performed for Watts Bar Nuclear Plant Unit 2.2) The basic assumptions associated with the system evaluation.
3) The resources utilized in the development of the shutdown paths 4) The Safe Shutdown Paths themselves as shown in the SPLDs of Figure 2-1 and 2-2, along with a statement of the key feature(s) used by the path to obtain and maintain the safe shutdown condition.
5) The Development of dependency matrices, showing the interrelationship between support systems and front line safety systems and functions (Table 2.2) and between support systems and other support systems (Table 2.3).6) The development of a Safe Shutdown Equipment List (Appendix A) identifying the critical front line and support equipment required for the safe shutdown paths shown in the SPLDs of Figures 2-1 and 2-2.The SSEL is used as input to the seismic evaluation of structures and mechanical

/ electrical components required during or after the seismic event, including the containment isolation function, and input to the selection and evaluation of relays for effects of low ruggedness relay chatter on performance of equipment required to achieve the Safe Shutdown Path functions.

WBN2 IPEEE DESIGN REPORT'Page 109 ATTACHMENT 1

Watts Bar Nuclear Plant Unit 2 IPEEE Seismic Report No: WBNIPEEE-003 Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List Revision 0 Page: 74

7.0 REFERENCES

1. Report No: WBNIPEEE-001, Rev. 2, Dated December 15, 1997 and entitled: "Watts Bar Nuclear Plant IPEEE Seismic Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List." 2. US NRC Generic Letter 88-20 Supplement 4, "Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities," June 1991.3. NUREG-1407, "Procedural and Submittal Guidance for the Individual Plant Examinations of External Events (IPEEE) for Severe Accident Vulnerabilities, US Nuclear Regulatory Commission, June 1991.4. EPRI NP-6041M, "A Methodology for Assessment of Nuclear Power Plant Seismic Margin," Revision 1, Electric Power Research Institute, Palo Alto, CA, August 1991.5. EG/CR-4826, Volume 2, "Seismic Margin Review of the Maine Yankee Atomic power Station," US Nuclear Regulatory Commission, March 1987.WBN2 IPEEE DESIGN REPORT Page 110 ATTACHMENT 1

Watts Bar Nuclear Plant Unit 2 IPEEE Seismic Report No: WBNIPEEE-003 Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List Revision 0 Page: 75 Appendix A SAFE SHUTDOWN EQUIPMENT LIST (SSEL)WBN2 IPEEE DESIGN REPORT Page 111 ATTACHMENT 1

TENNESSEE VALLEY AUTHORITY Watts Bar Unit 2 Nuclear Plant IPEEE Seismic Margins Evaluation Report -Safe Shutdown Paths and Safe Shutdown Equipment List Page Number Al Op Op atPW Floor Room EquiplD Description FlowDiag Path Func Or OsCat R Fleo Nm Nor Des R Elev Num WBN-2-FCV

-001-0004 -T WBN-2-FCV

-001-0007 -B WBN-2-FCV

-001-0011 -T WBN-2-FCV

-001-0014 -A WBN-2-FCV

-001-0015 -A WBN-2-FCV

-001-0016 -A WBN-2-FCV

-001-0022 -T WBN-2-FCV

-001-0025 -B WBN-2-FCV

-001-0029 -T WBN-2-FCV

-001-0032 -A WBN-2-FCV

-001-0051 -S WBN-2-FCV

-001-0052 WBN-2-FCV

-001-0147 -A WBN-2-FCV

-001-0148 -B WBN-2-FCV

-001-0149 -A WBN-2-FCV

-001-0150 -B WBN-2-FCV

-001-0181 -A WBN-2-FCV

-001-0182 -B SG 2 MAIN STM HDR ISOLATION VALVE BLOWDOWN ISOLATION VALVE SG-2 SG 2 MAIN STM HDR ISOLATION VALVE BLOWDOWN HDR FLOW CONTROL VALVE, SG-2 TO AUX FW PMP STM SUPPLY FROM SG NO 2 AUX FW PMP TURB STM SUPPLY FROM SG NO 4 SG 3 MAIN STM HDR ISOLATION VALVE BLOWDOWN HDR ISOLATION VALVE, SG-3 STEAM GENERATOR 4 MAIN STEAM ISOL VLV BLOWDOWN HDR FLOW CONTROL VALVE, SG-4 TD AUX FEEDWATER PMP TRIP & THROTTLE VALVE TD AUX FEEDWATER PMP GOVERNOR VALVE MAIN STEAM ISOL VLV LOOP 2 BYP WARMING VLV MAIN STEAM ISOL VLV LOOP 2 BYP WARMING VLV MAIN STEAM ISOL VLV LOOP 3 BYP WARMING VLV MAIN STEAM ISOL VLV LOOP 4 BYP WARMING VLV BLOWDOWN ISOL VLV INSIDE CNTMT, SG-1 BLOWDOWN ISOL VLV INSIDE CNTMT, SG-2 2-47W801-1 2-47W801-2 2-47W801-1 2-47W801-2 2-47W803-2 2-47W803-2 2-47W801-1 2-47W801-2 2-47W801-1 2-47W801-1 2-47W803-2 2-47W803-2 2-47W801-1 2-47W801-1 2-47W801-1 2-47W801-1 2-47W801-2 2-47W801-2 5 F,G 0 5 F,G 0 5 F,G 0 S F,G C C C C C AOV Yes 0 0 SOV Yes 0 0 AOV Yes AUX/757 0 SOV Yes AUX/713 All 5 E, G,F 0 O/C 5 E, G,F C O/C 5 5 5 5 2,3 2,3 5 5 5 6 5 5 F,G 0 C F,G 0 C/o F,G 0 F,G 0 E C E 0 F,G C F,G C F,G C F,G C G 0 G 0 C C O/C 0 C C C C C C MOV Yes 0 /ALVE VAULT ROOI MOV Yes 0 /ALVE VAULT ROOf AOV Yes AUX/713 SOY Yes AUX/713 All AOV Yes RXB/716 SOV Yes AUX/741 All MOV Yes AUX/737 A25 EHA Yes AUX/713 A25 AOV Yes AUX/713 A29 AOV Yes AUX/713 AOV Yes AUX/713 AOV Yes AUX/713 A14 SOV Yes AUX/713 0 SOV Yes AUX/737 A9 WBN2 IPEEE DESIGN REPORT Page 112 ATTACHMENT 1

TENNESSEE VALLEY AUTHORITY Watts Bar Unit 2 Nuclear Plant IPEEE Seismic Margins Evaluation Report -Safe Shutdown Paths and Safe Shutdown Equipment List Page Number A2 WBN-2-FCV

-001-0183 -A BLOWDOWN ISOL VLV INSIDE CNTMT, SG-3 WBN-2-FCV

-001-0184 -B BLOWDOWN ISOL VLV INSIDE CNTMT, SG-4 WBN-2-FSV-001-0004A -A SG 1 MAIN STM HDR ISOLATION VLV WBN-2-FSV

-001-0004B -B SG 1 MAIN STM HDR ISOLATION VLV WBN-2-FSV

-001-0004D -A SG 1 MAIN STM HDR ISOLATION VLV WBN-2-FSV

-001-0004E -A SG 1 MAIN STM HDR ISOLATION VLV WBN-2-FSV

-001-0004G -B SG 1 MAIN STM HDR ISOLATION VLV WBN-2-FSV

-001-0004H -B SG 1 MAIN STM HDR ISOLATION VLV WBN-2-FSV

-001-0011A -A SG 2 MAIN STM HDR ISOLATION VLV WBN-2-FSV

-001-0011B -B SG 2 MAIN STM HDR ISOLATION VLV WBN-2-FSV

-001-0011D -A SG 2 MAIN STM HDR ISOLATION VLV WBN-2-FSV

-001-0011E -A SG 2 MAIN STM HDR ISOLATION VLV WBN-2-FSV

-001-O011G -B SG 2 MAIN STM HDR ISOLATION VLV WBN-2-FSV

-001-0011H -B SG 2 MAIN STM HDR ISOLATION VLV WBN-2-FSV

-001-0022A -A SG 3 MAIN STM HDR ISOLATION VLV WBN-2-FSV

-001-0022B -B SG 3 MAIN STM HDR ISOLATION VLV WBN-2-FSV

-001-0022D -A SG 3 MAIN STM HDR ISOLATION VLV WBN-2-FSV

-001-0022E -A SG 3 MAIN STM HDR ISOLATION VLV WBN-2-FSV

-001-0022G -B SG 3 MAIN STM HDR ISOLATION VLV WBN-2-FSV

-001-0022H -B SG 3 MAIN STM HDR ISOLATION VLV 2-47W801-2 2-47W801-2 2-47W610-1-1 2-47W610-1-1 2-47W610-1-1 2-47W610-1-1 2-47W610-1-1 2-47W610-1-1 2-47W610-1-1 2-47W610-1-1 2-47W610-1-1 2-47W610-1-1 2-47W610-1-1 2-47W610-1-1 2-47W610-1-1 2-47W610-1-1 2-47W610-1-1 2-47W610-1-1 2-47W610-1-1 5 G 0 5 G 0 5 F,G 0 5 F,G 0 5 F,G C 5 F,G C 5 F,G C 5 F,G C 5 F,G 0 5 F,G 0 5 F,G C 5 F,G C 5 F,G C 5 F,G C 5 F,G 0 5 F,G 0 5 F,G C 5 F,G C 5 F,G C SOV Yes AUX/676 0 SOV Yes AUX/676 0 SOV Yes AUX/708 All SOV Yes CTL/708 All SOV Yes CTL/708 All SOV Yes AUX/708 All SOV Yes CTL/708 All SOV Yes CTL/708 All SOV Yes CTL/708 A10 SOV Yes cr1/708 A10 SOY Yes CTL/708 A10 SOV Yes CTL/708 A10 SOY Yes CTL/708 A10 SOV Yes CTL/708 A10 SOV Yes CTL/708 A10 SOV Yes CTL/708 A10 SOV Yes CTL/708 A10 SOV Yes CTL/708 A10 SOV Yes CTL/708 A10 2-47W610-1-1 5 F,G C 0 SOV Yes CTL/708 A10 WBN2 IPEEE DESIGN REPORT Page 113 ATTACHMENT I

TENNESSEE VALLEY AUTHORITY Watts Bar Unit 2 Nuclear Plant IPEEE Seismic Margins Evaluation Report -Safe Shutdown Paths and Safe Shutdown Equipment List WBN-2-FSV

-001-0029A -A WBN-2-FSV

-001-0029B -B WBN-2-FSV

-001-0029D -A WBN-2-FSV

-001-0029E -A WBN-2-FSV

-001-0029G -B WBN-2-FSV

-001-0029H -B WBN-2-PCV

-001-0005 -T WBN-2-PCV

-001-0012 -T WBN-2-PCV

-001-0023 -T WBN-2-PCV

-001-0030 -T WBN-2-PSV

-001-0006A -A WBN-2-PSV

-001-0006B -A WBN-2-PSV

-001-0006C -B WBN-2-PSV

-001-0013A -B WBN-2-PSV

-001-0013B -B WBN-2-PSV

-001-0013C -A WBN-2-PSV

-001-0024A -A WBN-2-PSV

-001-0024B -A WBN-2-PSV

-001-0024C -B WBN-2-PSV

-001-0031A -B SG 4 MAIN STM HDR ISOLATION VLV SG 4 MAIN STM HDR ISOLATION VLV SG 4 MAIN STM HDR ISOLATION VLV SG 4 MAIN STM HDR ISOLATION VLV SG 4 MAIN STM HDR ISOLATION VLV SG 4 MAIN STM HDR ISOLATION VLV SG 2 MAIN STM HDR PWR RELIEF CONTROL VLV SG 2 MAIN STM HDR PWR RELIEF CONTROLVLV SG 3 MAIN STM HDR PWR RELIEF CONTROL VLV SG 4 MAIN STM HDR PWR RELIEF CONTROL VLV SG1 MAIN PWR RELIEF CONT VLV SG1 MAIN PWR RELIEF CONTVLV SG1 MAIN PWR RELIEF CONT VLV SG2 MAIN PWR RELIEF CONT VLV SG2 MAIN PWR RELIEF CONT VLV SG2 MAIN PWR RELIEF CONT VLV SG3 MAIN PWR RELIEF CONT VLV SG3 MAIN PWR RELIEF CONT VLV SG3 MAIN PWR RELIEF CONT VLV SG4 MAIN PWR RELIEF CONT VLV 2-47W610-1-1 2-47W610-1-1 2-47W610-1-1 2-47W610-1-1 2-47W610-1-1 2-47W610-1-1 2-47W801-2 2-47WB01-2 2-47W801-2 2-47W801-2 2-47W610-1-1 2-47W610-1-1 2-47W610-1-1 2-47W610-1-1 2-47W610-1-1 2-47W610-1-1 2-47W610-1-2 2-47W610-1-2 2-47W610-1-2 2-47W610-1-2 Page Number A3 5 F,G 0 C SOV Yes CTL/708 5 F,G 0 C SOV Yes CTL/708 5 F,G C 0 SOV Yes CTL/708 5 F,G C 0 SOV Yes. CTL/708 5 F,G C 0 SOV Yes CTL/708 5 F,G C 0 SOV Yes CTL/708 5 E.FFG C O/C AOV Yes 729 All All 0 All All All 0 E,F,G E,F,G E,F,G E,F,G E,F,G E,F,G E,F,G E,F,G E,F,G E,F,G E,F,G E,F,G E,F,G C O/C AOV Yes C O/C AOV Yes C O/C AOV Yes ON ON/OFF SOY YES ON ON/OFF SOV YES ON ON/OFF SOV YES OFF ON/OFF SOV YES OFF ON/OFF SOV YES OFF ON/OFF SOV YES OFF ON/OFF SOV YES OFF ON/OFF SOV YES OFF ON/OFF SOV YES OFF ON/OFF SOV YES 729 729 729 729 729 729 729 729 729 729 729 729 729 0 0 All 0 0 A5 0 0 A16 0 0 A512 0 WBN2 IPEEE DESIGN REPORT Page 114 AI-TACHMENT 1

TENNESSEE VALLEY AUTHORITY Watts Bar Unit 2 Nuclear Plant IPEEE Seismic Margins Evaluation Report -Safe Shutdown Paths and Safe Shutdown Equipment List Page Number A4 WBN-2-PSV

-001-0031B -B WBN-2-PSV

-001-0031C -A WBN-2-PT -001-0002A -D WBN-2-PT -001-0002B -E WBN-2-PT -001-0009A -D WBN-2-PT -001-0009B -E WBN-2-PT -001-0020A -D WBN-2-PT -001-0020B -E WBN-2-PT -001-0027A -D WBN-2-PT -001-0027B -E WBN-2-SFV

-001-0512 WBN-2-SFV

-001-0513 WBN-2-SFV

-001-0514 WBN-2-SFV

-001-0515 WBN-2-SFV

-001-0516 WBN-2-SFV

-001-0517 WBN-2-SFV

-001-0518 WBN-2-SFV

-001-0519 WBN-2-SFV

-001-0520 WBN-2-SFV

-001-0521 SG4 MAIN PWR RELIEF CONTVLV SG4 MAIN PWR RELIEF CONT VLV MAIN STEAM LOOP 1 PRESSURE MAIN STEAM LOOP 1 PRESSURE MAIN STEAM LOOP 2 PRESSURE MAIN STEAM LOOP 2 PRESSURE MAIN STEAM LOOP 3 PRESSURE MAIN STEAM LOOP 3 PRESSURE MAIN STEAM LOOP 4 PRESSURE MAIN STEAM LOOP 4 PRESSURE MAIN STEAM SAFETY VALVES, SG-3 MAIN STEAM SAFETY VALVES, SG-3 MAIN STEAM SAFETY VALVES, SG-3 MAIN STEAM SAFEN VALVES, SG-3 MAIN STEAM SAFETY VALVES, SG-3 MAIN STEAM SAFETY VALVES, SG-2 MAIN STEAM SAFETY VALVES, SG-2 MAIN STEAM SAFETY VALVES, SG-2 MAIN STEAM SAFETY VALVES, SG-2 MAIN STEAM SAFETY VALVES, SG-2 2-47W610-1-2 5 EFG OFF ON/OFF SOV YES 729 0 2-47W610-1-2 2-47W801-1 2-47W801-1 2-47W801-1 2-47W801-1 2-47W801-1 2-47W801-1 2-47W801-1 2-47W801-1 2-47W801-1 2-47W801-1 2-47W801-1 2-47W801-1 2-47W801-1 2-47W801-1 2-47W801-1 2-47W801-1 2-47W801-1 E,F,G OFF ON/OFF SOV YES A,E ON A,E ON A,E ON A,E ON A,E ON A,E

  • ON A,E ON ON PT YES ON PT YES ON PT YES 729 713 713 729 A,E E E E E E E E E E ON C C C C C C C C C ON ON ON ON ON O/C O/C O/C O/c O/C O/C O/C O/C O/C PT YES PT YES PT YES SFV NO PT YES 729 PT YES 729 729 713 713 729 A705 0 0 0 0 0 0 0 0 A10 AI0 AI0 A10 A10 A10 AI0 Ai0 Ai0 SFV NO 729 SFV NO .729 SFV NO 729 SFV NO 729 SFV NO 729 SFV NO 729 SFV NO 729 SFV NO 729 2-47W801-1 1,3 E C O/C SFV NO 729 A10 WBN2 IPEEE DESIGN REPORT Page 115 ATTACHMENT 1

TENNESSEE VALLEY AUTHORITY Watts Bar Unit 2 Nuclear Plant IPEEE Seismic Margins Evaluation Report -Safe Shutdown Paths and Safe Shutdown Equipment List WBN-2-SFV

-001-0522 WBN-2-SFV

-001-0523 WBN-2-SFV

-001-0524 WBN-2-SFV

-001-0525 WBN-2-SFV

-001-0526 WBN-2-SFV

-001-0527 WBN-2-SFV

-001-0528 WBN-2-SFV

-001-0529 WBN-2-SFV

-001-0530 WBN-2-SFV

-001-0531 WBN-2-FCV

-003-0033 -A WBN-2-FCV

-003-0047 -B WBN-2-FCV

-003-0087 -A WBN-2-FCV

-003-0100 -B WBN-2-FCV

-003-0116A -A WBN-2-FCV

-003-0116B -A WBN-2-FCV

-003-0126A -B WBN-2-FCV

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TENNESSEE VALLEY AUTHORITY Watts Bar Unit 2 Nuclear Plant IPEEE Seismic Margins Evaluation Report -Safe Shutdown Paths and Safe Shutdown Equipment List TENNESSEE VALLEY AUTHORITY WBN-2-FCV

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TENNESSEE VALLEY AUTHORITY Watts Bar Unit 2 Nuclear Plant IPEEE Seismic Margins Evaluation Report -Safe Shutdown Paths and Safe Shutdown Equipment List WBN-2-PMP-018-0054/3 -A DAY TNK XFER PMP WBN-2-PMP

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TENNESSEE VALLEY AUTHORITY Watts Bar Unit 2 Nuclear Plant IPEEE Seismic Margins Evaluation Report -Safe Shutdown Paths and Safe Shutdown Equipment List WBN-1-FAN

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TENNESSEE VALLEY AUTHORITY Watts Bar Unit 2 Nuclear Plant IPEEE Seismic Margins Evaluation Report -Safe Shutdown Paths and Safe Shutdown Equipment List WBN-1-FCO

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TENNESSEE VALLEY AUTHORITY Watts Bar Unit 2 Nuclear Plant IPEEE Seismic Margins Evaluation Report -Safe Shutdown Paths and Safe Shutdown Equipment List Page Number A12 WBN-1-TS -030-0244D -A WBN-1-TS -030-0244E WBN-1-TS -030-0248A -B WBN-1-TS -030-0248B -B WBN-1-TS -030-0248D -B WBN-1-TS -030-0447A -A WBN-1-TS -030-0447B -A WBN-1-TS -030-0449A -B WBN-1-TS -030-0449B -B WBN-1-TS -030-0451A -A WBN-1-TS -030-0451B -A WBN-1-TS -030-0453A -B WBN-1-TS -030-04531 -B WBN-2-CLR

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TENNESSEE VALLEY AUTHORITY Watts Bar Unit 2 Nuclear Plant IPEEE Seismic Margins Evaluation Report -Safe Shutdown Paths and Safe Shutdown Equipment List WBN-2-CLR

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TENNESSEE VALLEY AUTHORITY Watts Bar Unit 2 Nuclear Plant IPEEE Seismic Margins Evaluation Report -Safe Shutdown Paths and Safe Shutdown Equipment List WBN-2-FAN

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TENNESSEE VALLEY AUTHORITY Watts Bar Unit 2 Nuclear Plant IPEEE Seismic Margins Evaluation Report -Safe Shutdown Paths and Safe Shutdown Equipment List WBN-2-FCO

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TENNESSEE VALLEY AUTHORITY Watts Bar Unit 2 Nuclear Plant IPEEE Seismic Margins Evaluation Report -Safe Shutdown Paths and Safe Shutdown Equipment List WBN-2-FCO

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TENNESSEE VALLEY AUTHORITY Watts Bar Unit 2 Nuclear Plant IPEEE Seismic Margins Evaluation Report -Safe Shutdown Paths and Safe Shutdown Equipment List WBN-2-FS -030-0447 -A WBN-2-FS -030-0448 -A WBN-2-FS -030-0449 -B WBN-2-FS -030-0450 -B WBN-2-FS -030-0451 -A WBN-2-FS -030-0452 -A WBN-2-FS -030-0453 -B WBN-2-FS -030-0454 -B WBN-2-FSV

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TENNESSEE VALLEY AUTHORITY Watts Bar Unit 2 Nuclear Plant IPEEE Seismic Margins Evaluation Report -Safe Shutdown Paths and Safe Shutdown Equipment List WBN-2-PMCL-030-0182 -B WBN-2-PMCL-030-0183 -A WBN-2-PMCL-030-0184 -A WBN-2-PMCL-030-0185 -B WBN-2-PMCL-030-0190 WBN-2-PMCL-030-0191 WBN-2-PT -030-0310 -A WBN-2-PT -030-0311 -B WBN-2-TS -030-0175 -A WBN-2-TS -030-0176 -B WBN-2-TS -030-0177 -A WBN-2-TS -030-0178 -B WBN-2-TS -030-0179 -B WBN-2-TS -030-0180 -A WBN-2-TS -030-0182 -B WBN-2-TS -030-0183 -A WBN-2-TS -030-0184A -A WBN-2-TS -030-0184B -A WBN-2-TS -030-0185A -B WBN-2-TS -030-0185B -B CENT CHARGING PUMP 2B-B COOLER CENT CHARGING PUMP 2A-A COOLER AFW/BA XFER PUMP SPACE COOLER 2A-A AFW/BA XFER PUMP SPACE COOLER 2B-B CCS & AUX FEEDWATER PUMP AREA COOLER CCS & AUX FEEDWATER PUMP AREA COOLER CONTAINMENT PRESS TRANSMITTER CONTAINMENT PRESS TRANSMITTER RHR PUMP 2A-A ROOM CLR TEMP RHR PUMP 2A-A ROOM CLR TEMP CS PUMP 2A-A ROOM CLR TEMP CS PUMP 28-B ROOM CLR TEMP SIS PUMP 2B-B ROOM CLR TEMP SIS PUMP 2A-A ROOM CLR TEMP CENT CHARG PUMP 2B-B RM CLR TEMP CENT CHARG PUMP 2A-A RM CLR TEMP AFW & BA TRANSFER PUMP CLR 2A-A AFW & BA TRANSFER PUMP CLR 2A-A AFW & BA TRANSFER PUMP CLR 28-B AFW & BA TRANSFER PUMP CLR 28-B Page Number A18 2-47W866-8 5 H V ON CLR Yes 692 2-47W866-8 5 H V ON CLR Yes 692 0 0 1-47W866-8 1-47W866-8 2-47W866-8 2-47W866-8 2-47W610-30-5 2-47W610-30-5 2-47W610-30-5 2-47W610-30-5 2-47W610-30-5 2-47W610-30-5 2-47W610-30-5 2-47W610-30-5 1-47W610-30-5, 1-47W610-30-5, 1-47W610-30-5, 1-47W610-30-5i 5 H V ON CLR Yes 713 5 H V ON CLR Yes 713 5 H V ON CLR Yes 723 5 H V ON CLR Yes 723 0 5 I ON ON 0 Y 0 0 5 I ON ON 0 Y 0 2,3,4 H O/C O/C TS No 676 2,3,4 H O/C O/C TS No 692 2,3,4 H O/C O/C TS No 676 2,3,4 H O/C O/C TS No 676 2,3,4 H O/C O/C TS No 692 2,3,4 H O/C O/C TS No 692 5 H O/C O/C TS No 692 5 H O/C O/C TS No 692 A 5 H O/C O/C TS No 713 A 5 H O/C O/C TS No 713 A 5 H O/C O/C TS No 713 A 5 H O/C O/C TS No 713 WBN2 IPEEE DESIGN REPORT Page 129 ATTACHMENT 1

TENNESSEE VALLEY AUTHORITY Watts Bar Unit 2 Nuclear Plant IPEEE Seismic Margins Evaluation Report -Safe Shutdown Paths and Safe Shutdown Equipment List WBN-2-TS -030-0186A -A WBN-2-TS -030-0186B -A WBN-2-TS -030-0187A -B WBN-2-TS -030-0187B -B WBN-2-TS -030-0190A -A WBN-2-TS -030-0190B -A WBN-2-TS -030-0191A -B WBN-2-TS -030-0191B -B WBN-2-TS -030-0196A -A WBN-2-TS -030-0196B -A WBN-2-TS -030-0197A -B WBN-2-TS -030-0197B -B WBN-2-TS -030-0200A -A WBN-2-TS -030-0201A -A WBN-2-TS -030-0201B -A WBN-2-TS -030-0202A -B WBN-2-TS -030-0202B -B WBN-2-TS -030-0207A -B WBN-2-TS -030-0244A -A WBN-2-TS -030-0244B -A PENETRATION RM EL 692 CLR 2A-A TEMP PENETRATION RM EL 692 CLR 2A-A TEMP PENETRATION RM EL 692 CLR 2B-B TEMP PENETRATION RM EL 692 CLR 2B-B TEMP CCS & AFW PUMP SPACE CLR 2A-A TEMP CCS & AFW PUMP SPACE CLR 2A-A TEMP CCS & AFW PUMP SPACE CLR 2B-B TEMP CCW & AFW PUMP SPACE CLR 2B-B TEMP PENETRATION RM EL 713 CLR 2A-A TEMP PENETRATION RM EL 713 CLR 2A-A TEMP PENETRATION RM EL 713 CLR 2B-B TEMP PENETRATION RM EL 713 CLR 2B-B TEMP EGTS ROOM COOLER 2A-A TEMP AB EL 692 PIPE CHASE CLR 2A-A TEMP ABEL 692 PIPE CHASE CLR 2A-A TEMP AB EL 692 PIPE CHASE CLR 26-B TEMP'AB EL 692 PIPE CHASE CLR 26-B TEMP EGTS ROOM COOLER 26-B TEMP TRANSFORMER RM 2A EXH FAN 2A2-A TEMP TRANSFORMER RM 2A EXH FAN 2A2-A TEMP 2-47W610-30-5 2-47W610-30-5 2-47W610-30-5 2-47W610-30-5 2-47W610-30-6 2-47W610-30-6 2-47W610-30-6 2-47W610-30-6 2-47W610-30-6 2-47W610-30-6 2-47W610-30-6 2-47W610-30-6 1-47W610-30-6A 2-47W610-30-6 2-47W610-30-6 2-47W610-30-6 2-47W610-30-6 1-47W610-30-6A 1-47W610-30-8 1-47W610-30-8 H H H H H H H H H H H 0/C 0/C 0/C 0/C 0/C 0/C 0/C 0/C 0/C 0/C 0/C 0/C 0/C 0/C 0/C 0/C 0/C 0/C 0/C 0/C 0/C 0/C Page Number A19 TS No 692 A325 TS No 692 0 TS No 692 A325 TS No 692 A25 TS No 713 Al TS No 713 Al TS No 713 Al TS No 713 Al TS No 713 A19 TS No 713 A19 TS No 713 A19 TS No 713 0 TS No 757 A16 Z--1 s H O/C O/C 5 H O/C O/C 5 H O/C o/C 5 H O/C O/C S H O/C O/C 5 H O/C O/C s H O/C O/C S H O/C O/C 5 .H O/C O/C TS No 692 TS No 692 TS No 692 TS No 692 TS No 757 TS No 772 TS No 772 A324 0 A324 0 A16 A6 A6 WBN2 IPEEE DESIGN REPORT Page 130 ATTrACHMENT 1

TENNESSEE VALLEY AUTHORITY Watts Bar Unit 2 Nuclear Plant IPEEE Seismic Margins Evaluation Report -Safe Shutdown Paths and Safe Shutdown Equipment List WBN-2-TS -030-0244D -A WBN-2-TS -030-0244E WBN-2-TS -030-0246A -B WBN-2-TS -030-0246B -B WBN-2-TS -030-0246D -B WBN-2-TS -030-0248A -B WBN-2-TS -030-0248B -B WBN-2-TS -030-0248D -B WBN-2-TS -030-0250A -A WBN-2-TS -030-0250B -A WBN-2-TS -030-0250D -A WBN-2-TS -030-0447A -A WBN-2-TS -030-0447B -A WBN-2-TS -030-0448A -A WBN-2-TS -030-0448B -A WBN-2-TS -030-0449A -B WBN-2-TS -030-0449B -B WBN-2-TS -030-0450A -B WBN-2-TS -030-0450B -B WBN-2-TS -030-0452A -A TRANSFORMER RM 2A EXH FAN 2A3-A TEMP TRANSFORMER RM 2A EXH FAN 2A4-A TEMP TRANSFORMER RM 2B EXH FAN 2B2-B TRANSFORMER RM 2B EXH FAN 2B2-B TRANSFORMER RM 2B EXH FAN 2B3-B TRANSFORMER RM 2B EXH FAN 2B2-B TEMP TRANSFORMER RM 2B EXH FAN 2B2-B TEMP TRANSFORMER RM 2B EXH FAN 2B2-B TEMP TRANSFORMER RM 2A EXH FAN 2A2-A TEMP TRANSFORMER RM 2A EXH FAN 2A2-A TRANSFORMER RM 2A EXH FAN 2A3-A DG 2A-A RM EXH FAN 2A HI TEMP DG 2A-A RM EXH FAN 2B HI TEMP DG 2A-A RM EXH FAN 2A HI TEMP EG 2A-A RM EXH FAN 2A LOW TEMP DG 2B-B RM EXH FAN 2B HI TEMP DG 2B-B RM EXH FAN 2B LO TEMP DG 2B-B RM EXH FAN 2B LOW TEMP DG 2B-B RM EXH FAN 2B LOW TEMP DG 2A-A RM EXH FAN 2A HI TEMP 1-47W610-30-8 1-47W610-30-8 1-47W610-30-8 1-47W610-30-8 1-47W610-30-8 1-47W610-30-8 1-47W610-30-8 1-47W610-30-8 l-47W610-30-8A l-47W610-30-8A l-47W610-30-8A 1-47W866-9 1-47WB66-9 1-47W866-9 1-47WS66-9 S H O/C O/C 5 H O/C O/C 5 H O/C O/C 5 H O/C O/C 5 H O/C O/C Page Number A20 TS No 772 TS No 772 TS No 772 TS No 772 TS No 772 A6 A6 All All All 5 5 5 5 5 5 5 5 5 5 o/C o/C 0/C 0/C 0/C 0/C 0/C 0/C 0/C 0/C 0/C 0/C 0/C 0/C 0/C 0/C 0/C 0/C 0/C 0/C 0/C 0/C 0/C 0/C TS No 772 TS No 772 TS No 772 TS No 772 TS No 772 TS No 772 TS No 760 TS No 760 TS No 760 TS No 760 TS No 760 TS No 760 AS AS AS A12 A12 A12 D3 D3 D6 D6 D9 D9 D14 D14 1-47W866-9 5 1-47W866-9 S 1-47W866-9 S 1-47W866-9 5 1-47W866-9 5 O/C O/C TS No 760 O/C O/C TS No 760 O/C O/C TS No 760 D6 WBN2 IPEEE DESIGN REPORT Page 131 ATTACHMENT 1

TENNESSEE VALLEY AUTHORITY Watts Bar Unit 2 Nuclear Plant IPEEE Seismic Margins Evaluation Report -Safe Shutdown Paths and Safe Shutdown Equipment List WBN-2-TS -030-0452B -A WBN-2-TS -030-0453A -B WBN-2-TS -030-0453B -B WBN-2-TS -030-0454A -B WBN-2-TS -030-0454B -B WBN-0-AHU

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TENNESSEE VALLEY AUTHORITY Watts Bar Unit 2 Nuclear Plant IPEEE Seismic Margins Evaluation Report -Safe Shutdown Paths and Safe Shutdown Equipment List WBN-0-CHR

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TENNESSEE VALLEY AUTHORITY Watts Bar Unit 2 Nuclear Plant IPEEE Seismic Margins Evaluation Report -Safe Shutdown Paths and Safe Shutdown Equipment List WBN-0-FCO

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TENNESSEE VALLEY AUTHORITY Watts Bar Unit 2 Nuclear Plant IPEEE Seismic Margins Evaluation Report -Safe Shutdown Paths and Safe Shutdown Equipment List WBN-0-FS -031-0154 -B WBN-0-FS -031-0401 -A WBN-0-FS -031-0402 -B WBN-0-PDIS-031-0101 -A WBN-0-PDIS-031-0131 -B WBN-0-PDIS-031-0161 -A WBN-0-PDIS-031-0186 -B WBN-0-PDIS-031-0211 -A WBN-0-PDIS-031-0241 -B WBN-0-PMP

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-031-0128 ELEC BD RM AHU D-B AIR FLOW BATTERY ROOM EL 692 EXH FAN B- B FLOW BATTERY ROOM EL 692 EXH FAN A-A FLOW SHUTDOWN BD RM CW PUMP A-A DIFF PRESS SHUTDOWN BD RM CW PUMP B-B DIFF PRESS MAIN CONTROL ROOM CW PUMP A-A DIFF PRE MAIN CONTROL ROOM CW PUMP B-B DIFF PRE ELEC BD ROOM CW PUMP A-A DIFF PRESS ELEC BD ROOM CW PUMP B-B DIFF PRESS SDBR CHLR PKG A-A CW CIRC PUMP A-A SDBR CHLR PKG B-B CW CIRC PUMP B-B CW CIRC PMP A-A (MCR)CW CIRC PMP B-B (MCR)ELEC BD RM CHILLER PKG A-A CIR PMP A-A ELEC BD RM CHILLER PKG B-B CIR PMP B-B SDBR CHLR PKG A-A CW AIR SEPARATOR A SDBR CHLR PKG B-B CW AIR SEPARATOR B AIR SEPARATOR-A-(MCR)

AIR SEPARATOR-B-(MCR)

AIR SEPARATOR A-(EBR)1-47W610-31-3 1-47W610-31-1 1-47W610-31-1 1-47W865-8 1-47W865-8 1-47W865-3 1-47W865-3 1-47W865-7 1-47W865-7 1-47W865-8 1-47W865-8 1-47W865-3 1-47W865-3 1-47W865-7 1-47W865-7 1-47W865-8 1-47W865-8 1-47W865-3 1-47W865-3 1-47W865-7 Page Number A24 5 H O/C O/C FS 'Yes 692 S H O/C O/C FS Yes 692 5 H O/C O/C FS *Yes 692 5 H O/C O/C DPIS No 737 5 H O/C O/C DPIS No 737 5 H O/C O/C DPIS No 755 5 H O/C O/C DPIS No 737 5 H O/C O/C DPIS No 692 5 H O/C O/C DPIS No 692 5 5 5 5 5 5 5 5 5 5 5 H ON ON PMP Yes 737 H ON ON PMP Yes 737 H ON ON PMP Yes 755 H ON ON PMP Yes 755 H ON ON PMP Yes 692 H ON ON PMP Yes 692 H 0 0 SEP Yes 737 H 0 0 SEP Yes 737 H 0 0 SEP No 755 H 0 0 SEP No 755 H 0 0 SEP No 692 C2 C3 C3 AOl A01 C12 Al ClO ClI Al Al Cl Cl CIO C10 Al Al Cl Cl C10 WBN2 IPEEE DESIGN REPORT Page 135 ATTACHMENT 1

TENNESSEE VALLEY AUTHORITY Watts Bar Unit 2 Nuclear Plant IPEEE Seismic Margins Evaluation Report -Safe Shutdown Paths and Safe Shutdown Equipment List WBN-O-SEP

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-031-0500 AIR SEPARATOR B-(EBR)ELECT BD RM AHU A-A STEAM GENERATOR MAIN CONT RM AHU A-A STEAM GENERATOR STRAINER (SDBR CHR A-A)STRAINER (SDBR CHR B-B)STRAINER (MCR CHR A-A)STRAINER (MCR CHR B-B)STRAINER (EBR CHR A-A)STRAINER (EBR CHR B-B)SDBR CHLR PKG A-A CW CHEM TREATMENT TK SDBR CHILR PKG B-B CW CHEM TREATMENT TK SDBR CHLR PKG A-A CW COMPRESSION TK A SDBR CHLR PKG B-B CW COMPRESSION TK B COMPRESSION TANK A-(MCR)COMPRESSION TANK B-(MCR)COMPRESSION TANK A-(EBR)MCR AHU A-A MODULATING DAMPER -CONTRO MCR AHU B-B MODULATING DAMPER -CONTRO SUPPLEMENTAL AHU 2 CW TCV SUPPLEMENTAL AHU 2 CW TCV 1-47WB65-7 1-47W865-7 1-47W865-3 1-47W865-8 Page Number A25 5 H 0 0 SEP No 692 5 H ON ON S STGR Yes 692 5 H ON ON STGR Yes 755 5 H 0 0 STN No 737 1-47W865-8 5 H 0 0 STN No 737 1-47W865-3 5 H 0 0 STN No 755 1-47W865-3 5 H 0 0 STN No 755 1-47W865-7 5 H 0 0 STN No 692 1-47W865-7 5 H 0 0 STN No 692 1-47W865-8 5 H 0 0 TANK Yes 737 1-47W865-8 5 H 0 0 TANK Yes 737 1-47W865-8 5 H 0 0 TANK Yes 737 1-47W865-8 5 H 0 0 TANK Yes 737 1-47W865-3 5 H 0 0 TANK No 755 1-47W865-3 5 H 0 0 TANK No 755 1-47W865-7 5 H 0 0 TANK No 692 1-47W610-31-2 5 H V V TC Yes 755 1-47W610-31-2 5 H V V TC Yes 755 1-47W865-12 5 H 0 0 TCV No 708 1-47W865-12 5 H 0 0 TCV No 708 C10 CIO C1 Al Al Cl Cl C10'C10 Al Al Al Al Cl Cl CIO C1 Cl C2 C2 WBN2 IPEEE DESIGN REPORT Page 136 ATTACHMENT 1

TENNESSEE VALLEY AUTHORITY Watts Bar Unit 2 Nuclear Plant IPEEE Seismic Margins Evaluation Report -Safe Shutdown Paths and Safe Shutdown Equipment List WBN-0-TS -031-0040B -A WBN-0-TS -031-0048B -A WBN-0-TS -031-0052B -B WBN-0-TS -031-0060B -B WBN-0-TS -031-0088B -A WBN-0-TS -031-0089B -B WBN-0-TS -031-0150B -A WBN-0-TT -031-0082 WBN-0-TT -031-0091 WBN-1-AHU

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-031-0285 -A S BD ROOM AHU A-A INLET TEMP S BD ROOM AHU B-A INLET TEMP SD BD ROOM AHU C-B INLET TEMP SB BD ROOM AHU D-B INLET TEMP, MAIN CONTROL RM AHU A-A SUPPLY TEMP MAIN CONTROL RM AHU B-B SUPPLY TEMP ELEC BD RM AHU CB/DB SUPPLY TEMP MCR AHU A-A MODULATING DAMPER -THERMO MCR AHU B-B MODULATING DAMPER -THERMO 480V BD ROOM 2A SUPPLY AHU 1A-A 480V BOARD ROOM 1B SUPPLY AHU 1B-B 480V BD RM 1B RECIP COMP 1B-B 480V BD RM 1A RECIP COMP lA-A 480V ELECT BD RM 1A AIR COOLED COND 480V ELECT BD RM 18 AIR COOLED COND BATTERY ROOM II EXHAUST FAN 182-A BATTERY ROOM II EXHAUST FAN 282-B BATTERY ROOM I EXHAUST FAN 2A2-A BATTERY ROOM I EXHAUST FAN 2A2-B BATTERY ROOM II DAMPER FOR FAN 181-A 1-47W610-31-4 1-47W610-31-4 1-47W610-31-4 1-47W610-31-4 1-47W610-31-2 1-47W610-31-2 1-47W610-31-3 1-47W610-31-2 1-47W610-31-2 1-47W866-3 1-47W866-3 1-47W865-6 1-47W865-6 1-47W866-3 1-47W866-3 1-47W866-3 1-47W866-3 1-47W866-3 1-47W866-3 1-47W866-3 5 5 5 5 5 5 5 5 5 H O/C H O/C H O/C H O/C H O/C H O/C H O/C H V H V Page Number A26 O/C TS No 757 O/C TS No 757 O/C TS No 757 O/C TS No 757 O/C TS No 755 O/C TS No .755 O/C TS No 692 V TT Yes 755 V TT Yes 755 A9 A817 A9 A817 C401 C401 C101 C12 C12 H ON ON H ON ON H ON ON H ON ON H ON ON H H H H H H ON ON ON ON ON 0 ON ON ON ON ON 0 AHU Yes 772 AHU Yes 786 COM Yes 786 COM Yes 772 COND Y 786 COND Y 772 FAN Yes 772 FAN Yes 772 FAN Yes 772 FAN Yes 772 DMP Yes 772 A7 A3 A3 A7 A7 A3 A3 A3 A4 A4 A3 WBN2 IPEEE DESIGN REPORT Page 137 ATTACHMENT 1

TENNESSEE VALLEY AUTHORITY Watts Bar Unit 2 Nuclear Plant IPEEE Seismic Margins Evaluation Report -Safe Shutdown Paths and Safe Shutdown Equipment List WBN-1-FCO

-031-0286 -B WBN-1-FCO

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-031-0475 -B WBN-2-COMP-031-0447 -B BATTERY ROOM II DAMPER FOR FAN 1B2-B BATTERY ROOM I DAMPER FOR FAN Al-A BATTERY ROOM I DAMPER FOR FAN 1A2-B BOARD ROOM CONDENSER 1B-B DAMPER BOARD ROOM CONDENSER 1A-A DAMPER 480V BD RM 2A COND 1A-A DAMPER 480V BD RM 1B FILTER DRIER 480V BD RM 1A FILTER DRIER 480V BD RM 1A REFRIGERANT LINE FSV 480V 8D RM 1+C796B REFRIGERANT LINE FSV 480V BD RM 1B R-22 GAS PRESS 480V 8D RM 1A R-22 GAS PRESS 480V BD RM 1B OIL SEPARATOR 480V BD RM 1A OIL SEPARATOR 480V BD RM 1A AHU 1A-A TEMP A/C 480V BD RM B TEMP SWITCH 480V BD RM 1B AHU 18-B TEMP 480V BD RM 2A AHU 2A-A 480V BD RM 2B AHU 2B-B 480V BD RM 2B RECIP COMP 2B-B 1-47W866-3 1-47W866-3 1-47W866-3 1-47W866-3 1-47W866-3 1-47W866-3 1-47W865-6 1-47W865-6 1-47W865-6 1-47W865-6 1-47W610-31-8 0 0 0 0 0 0 0 0 0 0 O/C O/C 0 0 O/C O/C O/C ON ON Page Number A27 0 DMP Yes 772 0 DMP Yes 772 0 DMP Yes 772 0 DMP Yes 786 0 DMP Yes 772 0 DMP Yes 772 0 DRIE No RXB/726 0 DRIE No RXB/726 0 FSV Yes AUX/692 0 FSV Yes AUX/692 O/C PS No 772 A3 A4 A4 A3 A7 A7 A3 A7 A7 A3 IC ANN A3 A7 A2 A2 A2 A1O A4 A4 1-47W610-31-7 5 1-47W865-6 5 1-47W865-6 1-47W610-31-7 1-47W610-31-7 1-47W610-31-8 1-47W866-3 1-47W866-3 1-47W865-6A 5 5 5 5 5 5 o/C 0 0 O/C O/C O/C ON ON PS No 722 SEP No 786 SEP No 772 TS No 772 TS No 772 TS No 772 AHU Yes 772 AHU Yes 786 5 H ON ON COM Yes AUX/786 WBN2 IPEEE DESIGN REPORT Page 138 ATTACHMENT 1

TENNESSEE VALLEY AUTHORITY Watts Bar Unit 2 Nuclear Plant IPEEE Seismic Margins Evaluation Report -Safe Shutdown Paths and Safe Shutdown Equipment List WBN-2-COMP-031-0465 -A 480V BD RM 2A RECIP COMP 2A-A WBN-2-COND-031-0289 -B 480V BD RM 2B AIR COOLED COND WBN-2-COND-031-0290 -A 480V BD RM 2A AIR COOLED COND WBN-2-FAN

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-031-0327 -B BTRY RM III EXHAUST FAN 2B2-A BTRY RM III EXHAUST FAN 282-B BTRY RM IV EXHAUST FAN 2A2-A BTRY RM IV EXHAUST FAN 2A2-B BTRY RM III DAMPER FOR FAN 2B2-A BTRY RM III DAMPER FOR FAN 282-B BTRY RM IV DAMPER FOR FAN 2A2-A BTRY RM IV DAMPER FOR FAN 2A2-B 480V BD RM COND UNIT 2B-B DAMPER 480V BD RM 2A COND 2A-A DAMPER 480V BD RM 2A COND 2A-A DAMPER ANNULUS ISOLATION VALVE INSTRUMENT ROOM ISOLATION VALVE INSTRUMENT ROOM ISOLATION VALVE ANNULUS ISOLATION VALVE ANNULUS ISOLATION VALVE INSTRUMENT ROOM ISOLATION VALVE 1-47W865-6A 1-47W866-3 1-47W866-3 1-47W866-3 1-47W866-3 1-47W866-3 1-47W866-3 1-47W866-3 1-47W866-3 1-47W866-3 1-47W866-3 1-47W866-3 1-47W866-3 1-47W866-3 2-47W865-5 2-47W865-5 2-47W865-5 2-47W865-5 2-47W865-5 2-47W865-5 Page Number A28 0 COM Yes AUX/772 A4 ON CON Yes AUX/786 A4 ON CON Yes AUX/772 Ai0 ON FAN Yes 772 A4 ON FAN Yes 772 A4 ON FAN Yes 772 A13 ON FAN Yes 772 A13 0 DMP Yes 772 A4 0 DMP Yes 772 A4 O DMP Yes AUX/772 A4 0 DMP Yes AUX/772 A13 0 DMP Yes 772 A13 0 DMP Yes AUX/772 A1O O DMP Yes AUX/772 A10 0 0 N 0 0 0 0 N 0 0 0 0 N 0 0 0 0 N AUX/676 0 0 0 N AUX/676 0 0 0 N 0 0 WBN2 IPEEE DESIGN REPORT Page 139 ATTACHMENT 1

TENNESSEE VALLEY AUTHORITY Watts Bar Unit 2 Nuclear Plant IPEEE Seismic Margins Evaluation Report -Safe Shutdown Paths and Safe Shutdown Equipment List WBN-2-FCV

-031-0329 -B WBN-2-FCV

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-031-0465 WBN-2-TS -031-0441A -A WBN-2-TS -031-0441C -A WBN-2-TS -031-0447A -B WBN-0-ACU M-032-0060 WBN-0-ACUM-032-0086 WBN-0-COM P-032-0060 WBN-0-COM P-032-0086 WBN-0-DRYR-032-0074 WBN-0-DRYR-032-0075 WBN-0-DRYR-032-0099 INSTRUMENT ROOM ISOLATION VALVE ANNULUS ISOLATION VALVE 480V BD RM 2B FILTER DRIER 480V BD RM 2A FILTER DRIER 480V BD RM 2A REFRIGERANT LINE FSV 480V BD RM 2B REFRIGERANT LINE FSV 480V BD RM 2B R-22 GAS PRESS 480V BD RM 2A R-22 GAS PRESS 480V BD RM 2B OIL SEPARATOR 480V BD RM 2A OIL SEPARATOR 480V BD RM 2A AHU 2A-A TEMP 480V BD RM 2A AHU 2A-A TEMP 480V ELEC BD RM 28 AHU 2B-B TEMP AUX CNT AIR COMP ACC TNK AUX CNT AIR COMP ACC TNK AUX CONTROL AIR COMPRESSOR A-A AUX CONTROL AIR COMPRESSOR B-B ESSNT CON AIR TRAN A DRYR 2 ESSNT CON AIR TRAN A DRYR 1 ESSNT CON AIR TRAN B DRYR 2 2-47W865-5 2-47W865-5 1-47W865-6A 1-47W865-6A 5 5 5 5 1-47W865-6A 5 1-47W865-6A 5 1-47W610-31-8 5 1-47W610-31-7 5 1-47W865-6A 5 1-47W865-6A 5 47W610-31-7A 5 47W610-31-7A 5 Page Number A29 H 0 0 0 N 0 H 0 0 0 N 0 H 0 0 DRIE No 786 H 0 0 DRIE No 772 H 0 0 FSV Yes 772 H 0 0 FSV Yes 786 H O/C O/C PS No 772 H O/C O/C PS No 722 H 0 0 SEP No 786 H 0 0 SEP No 772 H O/C O/C TS No 772 H O/C O/C TS No 772 H O/C O/C TS No 772 A -OF 0 TANK Yes 757 A A OF 0 TANK Yes 757 A A OFF ON COM Yes 757 A OFF ON COM Yes 757 A A OF 0 DRYR Yes 757 A A OF 0 DRYR Yes 757 A A OF 0 DRYR Yes 757 A 0 0 A4 A10 A10 A4 NA NA A4 A10 A15 A15 A15/A6U/A9U A6U A10U/A7U/A7U/A7U 47W610-31-8A 47W848-01 47W848-01 47W848-02 47W848-02 47W848-01 47W848-01 47W848-01 5 5 5 5 5 5 5 S WBN2 IPEEE DESIGN REPORT Page 140 ATTACHMENT 1

TENNESSEE VALLEY AUTHORITY Watts Bar Unit 2 Nuclear Plant IPEEE Seismic Margins Evaluation Report -Safe Shutdown Paths and Safe Shutdown Equipment List WBN-0-DRYR-032-0100 WBN-0-FCV

-032-0082 -A WBN-0-FCV

-032-0085 -B WBN-0-FLTR-032-0060 WBN-0-FLTR-032-0076 WBN-0-FLTR-032-0086 WBN-0-FLTR-032-0101 WBN-0-FSV

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-032-0086 WBN-0-PS -032-0062 -A WBN-0-PS -032-0062A -A WBN-0-PS -032-0062B -A WBN-0-PS -032-0082 -A WBN-0-PS -032-0085 -B WBN-0-PS -032-0088 -B WBN-0-PS -032-0088A -B WBN-0-PS -032-0088B -B WBN-0-RCVR-032-0062 WBN-0-RCVR-032-0088 ESSNT CON AIR TRAN B DRYR 1 ACAS COMPRESSOR A-A, ACAS ISOLATION ACAS COMPRESSOR B-B, ACAS ISOLATION AUX CNTL AIR COMP INTK FLTR ESSNT CON AIR TRN A AFT-FLTR AUX CNTL AIR COMP INTK FLTR ESSNT CON AIR TRN B AFT-FLTR ACAS COMPRESSOR A-A, ACAS ISOLATION AUX CNTL AIR COMP AFTRCOOL AUX CNTL AIR COMP AFTRCOOL AUXILIARY CONTROL AIR RECEIVER A LPS AUXILIARY AIR COMPRESSOR A-A UNLOAD SW AUXILIARY AIR COMPRESSOR A-A LOAD CTL SW CAS LOW PRESSURE ACAS ISOLATION CAS LOW PRESSURE ACAS ISOLATION AUXILIARY CONTROL AIR RECEIVER B LPS AUXILIARY AIR COMPRESSOR B-B UNLOAD SWIT AUXILIARY AIR COMPRESSOR B-B ESSNT CON AIR TR A AIR RECVR ESSNT CON AIR TR B AIR RECVR 47W848-01 47W848-1 47W848-1 47W848-01 47W848-01 47W848-01 47W848-01 47W848-1 47W848-01 47W848-01 47W848-1 47W848-1 47W848-1 47W848-1 47W848-1 47W848-1 47W848-1 47W848-1 47W848-01 47W848-01 A OF 0 FLTR Yes 757 A OF 0 FLTR Yes 757 A OF 0 FLTR Yes 757 A OF 0 FLTR Yes 757 A 0 C SOV Yes 757 A OF 0 HTX No 757 A OF 0 HTX No 757 A O/C O/C PS Yes 757 A O/C O/C PS Yes 757 A O/C O/C PS Yes 757 A C 0 PS Yes 757 A C 0 PS Yes 757 A O/C O/C PS Yes 757 A O/C O/C PS Yes 757 A O/C O/C PS Yes 757 A NA NA RCVR No 757 A NA NA RCVR No 757 A/A5U A/A7U A/A9U A/A9U A7U A13 A13 A6U A6U A6U A13 A13 A13 A9V A9V A/A8V A/A8V Page Number A30 5 A OF 0 DRYR Yes 757 A/A7U 5 A 0 C -AOV Yes 757 A7 5 A 0 C AOV Yes 757 A13 WBN2 IPEEE DESIGN REPORT Page 141 AI-TACHMENT I

TENNESSEE VALLEY AUTHORITY Watts Bar Unit 2 Nuclear Plant IPEEE Seismic Margins Evaluation Report -Safe Shutdown Paths and Safe Shutdown Equipment List WBN-2-FCV

-032-0081 -A WBN-2-FCV

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-043-0202 -A REACTOR BUILDING UNIT 2 TRAIN B ISOL VLV REACTOR BUILDING UNIT 2 TRAIN B ISOL VLV REACTOR BLDG UNIT 2 NONESNTL CNTL AIR PRSZR GAS SAMPL ISOL PRSZR GAS SAM PL ISOL PRSZR LIQUID SAMPL ISO HOT LEGS 2/3 SPL ISO ACCUM TANK SMPL HDR ISOL ACCUM TANK SMPL HDR ISOL STGR 2 DRM SPL ISO STEAM GEN 2 DRM SPL ISO STGR 2 DRM SPL ISO STEAM GEN 2 DRM SPL ISO STGR 3 DRM SPL ISO STEAM GEN 3 DRM SPL ISO STGR 4 DRM SPL ISO STEAM GEN 4 DRM SPL ISO DNSTR EXCESS LTDN HT EXCH-ISOL VLV LOCA H2 CNTMT MN IS LOCA H2 CNTMT MN OT 0 0 0 47W625-02 47W625-01 47W625-01 47W625-01 47W625-02 47W625-02 2-47W625-02 47W625-02 2-47W625-02 47W625-02 2-47W625-02 5 5 5 5 5 5 5 5 5 5 5 5.5 5 Page Number A31 C AOV N 0 C AOV N AUX/729 C AOV N AUX/729 C/V SOV Yes AUY/729 C/V AOV Yes AUX/729 C/V SOY Yes AUX/729 O/C AOV Yes AUX/729 C/V SOV Yes AUX/729 C/V SOV Yes AUX/729 C AOV Yes AUX/729 C AOV No AUX/729 C AOV Yes AUX/729 C AOV No AUX/729 C AOV Yes AUX/729 0 All All All All All All A10 A10 A10 A10 A10 A10 A10 A10 A10 A10 A10 A10 All 47W625-02 S 2-47W625-02 5 47W625-02 47W625-07 47W625-11 47W625-11 5 5 5 5 C C C C/V 0 0 AOV No AUX/729 AOV Yes AUX/729 AOV No AUX/729 SOV Yes AUX/729 SOV Yes AUX/729 SOV Yes AUX/729 WBN2 IPEEE DESIGN REPORT Page 142 ATTACHMENT 1

TENNESSEE VALLEY AUTHORITY Watts Bar Unit 2 Nuclear Plant IPEEE Seismic Margins Evaluation Report -Safe Shutdown Paths and Safe Shutdown Equipment List Page Number A32 WBN-2-FCV

-043-0207 -B WBN-2-FCV

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-043-0342 -A LOCA H2 CNTMT MONITOR ISOLATION VLV LOCA H2 CNTMT MN OT LOCA H2 CNTMT MN IS LOCA H2 CNTMT MN IS LOCA H2 CNTMT MN IS LOCA H2 CNTMT MN IS DNSTR EXCSW ITDN HT EXC ISO EXCESS ITDN HX SMPL ISOL PAS HOT LEG 2 SAMPLE ISOL PAS HOT LEG 2 SAMPLE ISOL PAS CONT AIR SUPPLY ISOL PAS CONT AIR SUPPLY ISOL PAS CONTAIR RETURN ISOL PAS HOT LEG 3 SAMPLE ISOL PAS HOT LEG 3 SAMPLE ISOL PAS CONT AIR SPL ISO PAS CONT AIR SPL ISO PAS CNTMT AIR RETRN ISOL PAS WASTE TO CONT SUMP ISOL PAS WASTE TO CONT SMP ISOL 47W625-11 47W625-11 47W625-11 47W625-11 47W625-11 47W625-11 47W625-07 47W625-07 47W625-15 47W625-15 47W625-15 47W625-15 47W625-15 47W625-15 47W625-15 47W625-15 47W625-15 47W625-15 47W625-15 47W625-15 0 0 0 0 0 0 C/V C/V C/v C/v c/v C/v C/v C/v C/V c/V C/v C/v c/v C/v SOV Yes AUX/729 SOV Yes 0 SOY Yes AUX/729 SOV Yes AUX/729 SOV Yes AUX/729 SOY Yes RXB/726 SOV Yes CrL/705 AOV Yes CTL/708 SOV Yes CTL/708 SOV Yes CTL/708 SOV Yes CTL/708 SOY Yes CTL/708 SOV Yes 0 SOV Yes AUX/757 SOV Yes AUX/757 SOV Yes RXB/703 SOV Yes 0 SOV Yes 0 SOV Yes 0 SOV Yes 0 All 0 All All All AC4 0 0 0 0 0 0 0 0 0 0 0.0 0 0 WBN2 IPEEE DESIGN REPORT Page 143 ATTiACHMENT 1

TENNESSEE VALLEY AUTHORITY Watts Bar Unit 2 Nuclear Plant IPEEE Seismic Margins Evaluation Report -Safe Shutdown Paths and Safe Shutdown Equipment List WBN-2-FCV

-061-0096 -A WBN-2-FCV

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WBN-2-FCV

-062-0009 GLYCOL FLOOR SUPPLY ISOLATION VALVE INLET ISOLATION VALVE REACTOR BLDG GLYCOL FLOOR RETURN ISOLATION VALVE OUTLET ISOLATION VALVE REACTOR BLDG GLYCOL AIR HANDLER SUPPLY ISO VALVE GLYCOL SUPPLY ISOLATION VALVE GLYCOL AIR HANDLER RETURN ISO VALVE GLYCOL RETURN ISOLATION VALVE INLET ISOLATION VALVE AUX BLDG GLYCOL SUPPLY HDR ISO VLV GLYCOL RETURN HDR ISO VLV GLYCOL RETURN HDR ISO VLV GLYCOL SUPPLY CONT ISO VLV GLYCOL SUPPLY CONT ISO VLV GLYCOL FLOOR RETURN ISO VLV GLYCOL FLOOR RETURN ISO VLV HOLDUP TANK A CVCS MIXED BED DEMINERALIZER 2B ICE COND LOWER INLET DOOR (PAIR) 2 THRU 24 RCP 1 SEAL LEAKOFF FLOW CONTROL 2-47W824-2 2-47W824-2 2-47W824-2 1,2-47W814-2 2-47W824-2 2-47W824-2 2-47W824-2 1,2-47W814-2 2-47W620 S F 5 F 5 F 5 F 5 F 5 F 5 F 5. F 5 F,G 0 0 0 0 0 0 0 0 ON C C C C C C C C OFF Page Number A33 AOV Y RXB/702 AOV Y 0 AOV Y 0 AOV Y 0 AOV Y 0 AOV Y 0 AOV Y 0 AOV Y 0 SOV Yes AUX/713 ANN 0 0 0 0 0 0 0 0 2-47W620 5 F,G 2-47W620 5 F,G 2-47W620 5 F,G 2-47W620 5 F,G 2-47W620 5 F,G 2-47W620-61-5 5 F,G 2-47W620-61-5 5 F,G 147W809-3 2,2 D ON OFF SOV Yes 0 0 ON OFF SOV Yes /729 0 ON OFF SOV Yes /729 0 ON OFF SOV, Yes AUX/737 A5 ON OFF SOV Yes /729 0 ON OFF SOV Yes /729 0 ON OFF SOV Yes AUX/729 A16.. .. TANK No 676 A/A5R 2-47W809-2 47W824-2 2 D 2,3,4 E ON C ON DEMN No AUX/692 0 0 DOOR RXB OFF Yes 0. 0 2-47W620-62-2 5 0 OFF WBN2 IPEEE DESIGN REPORT Page 144 ATTACHMENT 1

TENNESSEE VALLEY AUTHORITY Watts Bar Unit 2 Nuclear Plant IPEEE Seismic Margins Evaluation Report -Safe Shutdown Paths and Safe Shutdown Equipment List WBN-2-FCV

-062-0022 WBN-2-FCV

-062-0035 WBN-2-FCV

-062-0048 RCP 2 SEAL LEAKOFF FLOW CONTROL RCP 3 SEAL LEAKOFF FLOW CONTROL RCP 4 SEAL LEAKOFF FLOW CONTROL 2-47W620-62-2 1-47W610-62-1 1-47W610-62-1 2-47W809-1 2-47W809-1 Page Number A34 5 D OFF OFF Yes 0 5 D OFF OFF Yes 0 5 D OFF OFF Yes 0 0 D 0 C MOV Y 0 0 D 0 C MOV Y 0 0 0 0 0#N/A WBN-2-FCV

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-062-1228 -A BORIC ACID ISOLATION VLV 2-47W809-1 0 F 0 C AOV FC 0 2-47W809-1 0 F 0 C AOV FC#N/A 2-47W809-1 0 F 0 C AOV FC 0 0 2,2-47W809-1 2,2 D/F 0/0 O/C AOV N AUX/757 0 2,2-47W809-1 2-47W809-1 2,2-47W809-1 2-47W809-1 2-47W809-1 2-47W809-1 2-47W809-1 2-47W809-1 2-47W809-1 2-47W809-1 2-47W809-1 2,2 D/F 0/0 O/C AOV N 0 5 F,D C C/O AOV Yes 0 2,2 0 0/0 O/C 0 N/Y 0 2 C C 0 AOV Yes 0 0 D,G 0 O/C AOV N 0 5 D,F T T AOV Yes 0 0 D 0 0 0 N 0 0 D 0 0 0 N 0 5 D T T

  • AOV Yes 0 0 F C C 0 N 0 5 G 0 C AOV Yes 0 0 WBN2 IPEEE DESIGN REPORT Page 145 ATTACHMENT I

TENNESSEE VALLEY AUTHORITY Watts Bar Unit 2 Nuclear Plant IPEEE Seismic Margins Evaluation Report -Safe Shutdown Paths and Safe Shutdown Equipment List Page Number A35 WBN-2-FCV

-062-1229 -B WBN-2-FLTR-062-0065 WBN-2-FLTR-062-0096 WBN-2-FLTR-062-0097 WBN-2-FLTR-062-0117 WBN-2-FSV

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-062-0085 -B WBN-2-FT -062-0001 WBN-2-FT -062-0014 BORIC ACID ISOLATION VLV SEAL WATER RETURN FILTER SEAL WATER INJECTION FILTER B SEAL WATER INJECTION FILTER A REACTOR COOLANT FILTER RCP 2 SEAL INJECT ISO VLV SOLENOID RCP 2 SEAL INJECT ISO VLV SOLENOID RCP 3 SEAL INJECT ISO VLV SOLENOID RCP 4 SEAL INJECT ISO VLV SOLENOID RC LOOP 3 LETDOWN ISOLATION VALVE RC LOOP 3 LETDOWN ISOLATION VALVE REGEN HT EXCH LETDOWN ORIFICE A ISO VL REGEN HT EXCH LETDOWN ORIFICE B ISO VL REGEN HT EXCH LETDOWN ORIFICE C ISO VL LETDOWN ORIFICE ISOLATION LETDOWN ORIFICE ISOLATION AUX PRZR SPRAY NORMAL CHARGING ISOLATION VALVE RCP NO.2 SEAL WATER FLOW RCP NO.2 SEAL WATER FLOW 2-47W809-1 5 G 0 C AOV Yes 0 0 2-47W809-2 2-47W809-2 2-47W809-2 2-47W809-2 0 0 5 0 2-47W620-62-1 5 2-47W620-62-1 5 2-47W620-62-1 2-47W620-62-1 2-47W620-62-2 2-47W620-62-2 2-47W620-62-2 2-47W620-62-2 2-47W620-62-2 2-47W620-62-2 2-47W620-62-2 2-47W620-62-2 2-47W809-2 2-47W809-1 2-47W809-1 D ON ON 0 N AUX/737 A ON ON 0 N AUX/737 D ON ON FLTR No #N/A A ON ON 0 N #N/A D OFF OFF SOV Yes /729 D OFF OFF SOV Yes /729 D OFF OFF SOV Yes AUX/729 D OFF OFF SOV Yes /729 F,D ON ON SOV Yes /729 F,D ON ON SOV Yes AUX/737 F OFF OFF SOV Yes AUX/713 F,D ON ON SOV Yes AUX/713 F,D OFF OFF SOY Yes 0 F,D OFF OFF SOV Yes 0 F,D ON ON SOV Yes 0 C OFF ON/OFF SOV Yes 0 D,G 0 O/C AOV Yes AUX/713 D ON ON LITS Y D ON ON LITS Y 0 0#N/A#N/A 0 0 A512 0 0 A705 0 0 0 0 0 0 0 0 0 WBN2 IPEEE DESIGN REPORT Page 146 ATTACHMENT 1

TENNESSEE VALLEY AUTHORITY Watts Bar Unit 2 Nuclear Plant IPEEE Seismic Margins Evaluation Report -Safe Shutdown Paths and Safe Shutdown Equipment List Page Number A36 WBN-2-FT -062-0027 WBN-2-FT -062-0040 WBN-2-FT -062-0093A WBN-2-HIC

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-062-0118B -A WBN-2-LT -062-0129A WBN-2-LT -062-0129C WBN-2-LT -062-0130A WBN-2-LT -062-0130C RCP NO.3 SEAL WATER FLOW RCP NO.4 SEAL WATER FLOW CHARGING HEADER FLOW CONTROL CHARGING FLOW CONTVLV CHARGING HEADER FLOW CONTROLLER SEAL WATER HEAT EXCHANGER REGENERATIVE HEAT EXCHANGER EXCESS LETDOWN HEAT EXCHANGER LETDOWN HEAT EXCHANGER HOLDUP TANK VALVE VCT OUTLET ISOLATION VALVE LEVEL CONTROL VCT OUTLET ISOLATION VALVE LEVEL CONTROL RWST CVCS SUPPLY HDR ISOLATION RWST CVCS SUPPLY HDR ISOLATION DIVERSION FLOW TO HOLDUP TANKS HOLDUP TANK -DIVERSION VALVE VOLUME CONTROL TANK LEVEL VOLUME CONTROL TANK LEVEL VOLUME CONTROL TANK LEVEL VOLUME CONTROL TANK LEVEL 2-47W809-1 2-47W809-1 2-47W809-1 47W610-62-2 2-47W610-62-2 2-47W809-1 2-47W809-1 2-47W809-1 2-47W809-1 2-47W809-1 2-47W809-1 2-47W809-1 2-47W809-1 2-47W809-1 2-47W610-62-3 2-47W610-62-3 2-47W809-1 2-47W809-1 2-47W809-1 2-47W809-1 D D D D D D D D D D,G D,F D,F G,D G,D D,G D,G D D D D ON ON LITS Y 0 ON ON LITS Y 0 ON ON FT Yes 0 ON ON HIC Yes 0 ON ON HIC Yes 0 ON ON FSST N AUX/729 A10 ON NR HSCT N AUX/729 AiD OFF NR 0 N AUX/729 A10 ON NR HSCT N AUX/729 AIO O C/O AOV Yes 0 0 O C MOV Y 0 0 O C MOV Y 0 0 C 0 MOV Y 0 0 C 0 MOV Y 0 0 OFF ON/OFF SOV Yes 0 0 OFF ON/OFF SOV Yes 0 0 ON ON LT Yes AUX/692 0 ON ON LT Yes AUX/692 A25 ON ON LT Yes AUX/713 Al ON ON LT Yes AUX/713 Al WBN2 IPEEE DESIGN REPORT Page 147 ATTACHMENT 1

TENNESSEE VALLEY AUTHORITY Watts Bar Unit 2 Nuclear Plant IPEEE Seismic Margins Evaluation Report -Safe Shutdown Paths and Safe Shutdown Equipment List WBN-2-PCV

-062-0081 WBN-2-PMP

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-063-0007 -A LETDOWN HEAT EXCH PRESS CONT CENTRIFUGAL CHARGING PUMP 2B-B CENTRIFUGAL CHARGING PUMP 2A-A CVCS SEAL WTR RETURN HEADER RELIEF HOLDUP TANK B HOLDUP TANK B VOLUME CONTROL TANK LETDOWN FLOW TEMP DIVERSION CONTVLV LETDOWN FLOW TEMP SWITCH LETDOWN FLOW TEMP DIVERSION CONTROL SIS ACCUMULATOR TANK NO 1 SIS ACCUMULATOR TANK NO 2 SIS ACCUMULATOR TANK NO 3 SIS ACCUMULATOR TANK NO 4 RWSTTO RHR PMP FLOW CNTL VLV SIS PUMP DISCHARGE TO RWST SHUTOFF VALVE SIS PUMP A-A DISCH TO RWST SHUTOFF VALVE RWST TO SIS PUMP FLOW CONTROL VALVE SIS PUMP INLET TO CVCS CHARGING PUMP SIS PUMP INLETTO CVCS CHARGING PUMP 2-47W809-1 2-47W809-1 2-47W809-1 147W809-3 147W809-3 147W809-1 2-47W610-62-3 2-47W610-62-3 1-47W811-1 1-47W811-1 1-47W811-1 1-47W811-1 2-47W811-1 Page Number A37 0 2,2 D ON ON 0 Y 0 0 0 D ON ON HP Y 692 A22 0 D ON ON HP Y 692 A23 5 D C 0 RFV No 716 0 2,2 D .. .. TANK No 676 A/A5R 2,2 D .. .. TANK No 676 A/A8R 1,2 D -- -TANK No 713 0 0 0 0 OC O,C 0. Y 690 0 1,2 D C C TIS Yes 737 0 1,2 D :)N/OFION/OFF SOV Yes 713 A/A4T 3 D OFF ON FSST N RXB/702 ANN 3 D OFF ON FSST N RXB/702 ANN 3 D OFF ON FSST N RXB/702 ANN 3,4 D OFF ON FSST N AUX/770 A16 2,3,4 8,D,E 0 O/C MOV Y 0 0 ZI 2-47W811-1 2,3,4 D,E 0 C MOV Y 0 2-47W811-1 2,3,4 D,E 0 C MOV Y 0 0 0 2-47W811-1 2-47W811-1 2-47W811-1 2,3,4 B,D,E O/C 0 MOV Y AUX/723 0 2,3,4 D,E C 0 MOV Y 0 0 2,3,4 D,E C 0 MOV Y 0 0 WBN2 IPEEE DESIGN REPORT Page 148 ATTACHMENT 1

TENNESSEE VALLEY AUTHORITY Watts Bar Unit 2 Nuclear Plant IPEEE Seismic Margins Evaluation Report -Safe Shutdown Paths and Safe Shutdown Equipment List Page Number A38 WBN-2-FCV

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-063-0111 RHR HTX A TO CVCS CHARGING PUMPS RHR HTX B TO SIS PUMPS SIS PUMP OUTLET TO SIS TEST LINE SIS PUMPS COLD LEG INJECTION CONTROL VLV SIS ACCUM FILL LINE ISOLATION VALVE SIS CCP INJ TANK SHUTOFF VALVE SIS BORON INJ TANK SHUTOFF VALVE SIS PUMP A-A INLET VLV SIS PUMP B-B INLET VLV SIS ACCUM TANK N2 HDR INLET VALVE SIS ACCUMULATOR TNK 4 FLOW ISOL VALVE SIS CHECK VALVE FLOW ISOLATION VALVE RHR CONTAINMENT SUMP FLOW ISOL VLV RHR CONTAINMENT SUMP FLOW ISOL VLV SIS ACCUMULATOR TNK 3 FLOW ISOL VALVE SIS CHK VLV ISOL HDR FLOW ISOLATION VLV RHR PUMP A-A DISCHARGE TO CL 2&3 RHR PUMP B-B DISCHARGE TO CL 2&4 SIS ACCUMULATOR TNK 2 FLOW ISOL VALVE SIS CHECK VLV LEAK TEST ISOLATION VALVE 2-47W811-1 2-47W811-1 2-47W811-1 2-47W811-1 2-47W811-1 2-47W811-1 2-47W811-1 2-47W811-1 2-47W811-1 2-47W830-6 2-47W811-1 2-47W811-1 2-47W811-1 2-47W811-1 2-47W811-1 2-47W811-1 2-47W811-1 2-47W811-1 2-47W811-1 2-47W811-1 2,3,4 D,E C 2,3,4 D,E C 5 G,F C 2,3,4 D,E 0 2,3,4 G,F C 2,3,4 B,D,E C 2,3,4 B,D,E C 2,3,4 B,D,E 0 2,3,4 B,D,E D 5 G,F C 3,4 D 0 2,3,4 G,F C 2,3,4 E,D C 2,3,4 E,D C 3,4 D 0 2,3,4 G,F C 2,3,4 B,D,E 0 2,3,4 B,D,E 0 3,4 D D 2,3,4 G,F C 0 MOV Y 0 0 MOV Y 0 C AOV Y 0 C MOV Y 0 C AOV Y 0 0 0 0 0#N/A 0 MOV Y AUX/723 0 0 MOV Y 0 0 O/C MOV Y 0 0 O/C MOV Y 0 0 C AOV Y C MOV Y C AOV Y 0 MOV Y 0 MOV Y 0 MOV Y C AOV Y O/C MOV Y O/C MOV Y C MOV Y C AOV Y 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0#N/A WBN2 IPEEE DESIGN REPORT Page 149 ATTACHMENT 1

TENNESSEE VALLEY AUTHORITY Watts Bar Unit 2 Nuclear Plant IPEEE Seismic Margins Evaluation Report -Safe Shutdown Paths and Safe Shutdown Equipment List Page Number A39 WBN-2-FCV

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-063-0185 -A SIS CHECK VLV LEAK TEST ISOLATION VALVE SIS ACCUMULATOR TNK 2 FLOW ISOL VALVE SIS FLOW TO CL CHECK VALVE TEST SIS PUMP A-A OUTLET FLOW CONTROL VALVE SIS PUMP B-B OUTLET FLOW CONTROL VALVE SIS PUMP OUTLET TO RCS LP 2&3 HL SIS PUMP OUTLET TO RCS LP 2&4 HL CONT ISO SIS PUMP OUTLET TEST LINE RHR INJECTION OR RECIRC AFTER LOCA SIS TEST LINE CHECK VALVE TEST SIS PUMP B-B DISCH TO RWST SHUTOFF VLV SUMP RECIRC FROM RHR TO SIP CONT ISO SiS PUMP OUTLET TO SIS TEST LINE SIS ACUM FILL LINE ISO VLV SIS ACUM TANK N2 HDR INLET VLV SIS CHECK VLV FLOW ISO VLV SIS CHECK VLV HDR FLOW ISO VLV RHR SUPPLY TEST LINE VLV 2-47W811-1 2-47W811-1 2-47W811-1 2-47W811-1 2-47W811-1 2-47W811-1 2-47W811-1 2-47W811-1 2-47W811-1 2-47W811-1 2-47W811-1 2-47W811-1 2-47W811-1 2-47W811-1 2-47W620-63-1 2-47W620-63-1 2-47W620-63-2 2-47W620-63-1 2-47W620-63-1 2-47W620-63-2 2,3,4 G,F C 2,3,4 D 0 2,3,4 G C 2,3,4 B,D,E C/O 2,3,4 BD,E C/O 2,3,4 B,D,E C 2,3,4 B,D,E C 5 G,F C 5 G,F C 2,3,4 D,E C 5 F,G C 2,3,4 D,E 0 2,3,4 D,E 0 5 G,F C 2,3,4 D,E,F OF 2,3,4 D,E,F OF 5 G,F OF 2,3,4 G,F OF 2,3,4 G,F OF 5 G,F OF C AOV Y 0 0 MOV Y 0 C 0 Y 0 C MOV Y 0 C MOV Y 0.0 MOV Y 0 0 MOV Y 0 C AOV Yes 0 C AOV Y O/C MOV Y C AOV Y 0 0 0 0 0.0 Al Al C MOV Y 0 0 O MOV Yes AUX/676 0 C AOV Yes AUX/676 0 OFF SOV Yes AUX/713 0 OFF SOV Yes RXB/728 0 OFF SOV Yes RXB/728 0 OFF SOV Yes 0 0 OFF SOV Yes 0 0 OFF SOV Yes RXB/716 0 WBN2 IPEEE DESIGN REPORT Page 150 ATTACHMENT 1

TENNESSEE VALLEY AUTHORITY Watts Bar Unit 2 Nuclear Plant IPEEE Seismic Margins Evaluation Report -Safe Shutdown Paths and Safe Shutdown Equipment List Page Number A40 WBN-2-ISV

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-067-0032 -A SIS PUMP DISCH ISLN SI PUMP 28-B DISCHARGE ISOLATION SIS RWST LEVEL TRANSMITTER SIS RWST LEVEL TRANSMITTER RWST LEVEL RWST LEVEL CNTMT SUMP LEVEL CNTMT SUMP LEVEL CONTAINMENT SUMP LEVEL CONTAINMENT SUMP LEVEL SIS PUMP 2A-A SIS PUMP 2B-B CONTAINMENT SUMP STRAINER BORON INJECTION TANK 2A (CCP iNJ TNK)REFUELING WATER STORAGE TANK CCS HTX OUTLET CCS HX C ERCW HDR 2A OUT FLOW CCS HX C ERCW HDR 28 OUT FLOW ESSENTIAL RAW COOLING WATER PUMP A-A ESSENTIAL RAW COOLING WATER PUMP B-A 2-47W811-1 2-47W811-1 2-47W811-1 2-47W811-1 2-47W811-1 2-47W811-1 2-47W610-2 2-47W610-2 2-47W610-2 2-47W610-2 2-47W811-1 2-47W811-1 2-47W811-1 2-47W811-1 1-47W845-2 1-47W845-2 1-47W845-2 1-47W859-1 1-47W859-1 2,3,4 2,3,4 2,3,4 2,3,4 2,3,4 2,3,4 2,3,4 2,3,4 2,3,4 2,3,4 2,3,4 2,3,4 2,4 2,3,4 2,3,4 5 5 5 5 5 B,D,E B,D,E A A A A A A A A B,D,E B,D,E A B,D,E B,D,E A A A A A 0 0 0 0 0 0 0 0 0 0 OFF OFF NR OFF OFF 0 C C ON ON o/c O/O/0 0 0 0 0 0 0 0 ON ON 0 ON ON 0 C 0/c ON ON ISV ISV LT LT LT LT LT LT LT LT PMP PMP STR TAN K TANK MOV MOV MOV PMP PMP No No Yes Yes Yes Yes Yes Yes Yes Yes Y Y NO N N Yes Yes Yes Yes Yes AUX/729 RXB/703 AUX/713 AUX/713 AUX/713 AUX/692 AUX/692 AUX/692 AUX/692 CTL/7SS 669 669 690 705 737 737 737 742 742 A01 0 A19 A19 0 A324 0 A324 0 0 A19 A20 0 0 A01 A02 A02 IPS IPS WBN2 IPEEE DESIGN REPORT Page 151 AI-TACHMENT 1

TENNESSEE VALLEY AUTHORITY Watts Bar Unit 2 Nuclear Plant IPEEE Seismic Margins Evaluation Report -Safe Shutdown Paths and Safe Shutdown Equipment List WBN-0-PMP

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-067-0215 -B ESSENTIAL RAW COOLING WATER PUMP C-A ESSENTIAL RAW COOLING WATER PUMP D-A ESSENTIAL RAW COOLING WATER PUMP E-B ESSENTIAL RAW COOLING WATER PUMP F-B ESSENTIAL RAW COOLING WATER PUMP G-B ESSENTIAL RAW COOLING WATER PUMP H-B AUTO WATER REGULATING VALVE EBR CHILLER A-A AUTO WATER REGULATING VALVE MCR CHILLER A-A AUTO WATER REGULATING VALVE EBR CHILLER B-B AUTO WATER REGULATING VALVE MCR CHILLER B-B ERCW STRAINER lA-A BACKWASH ERCW STRAINER 1A-A FLUSH CONTROL ERCW STRAINER 1B-B BACKWASH ERCW STRAINER 1B-B FLUSH CONTROL EMERG DSL HTXS B1&B1 SUP VLV FROM HDR A EMERG DSL HTXS A1&A1 SUP VLV FROM HDR B CCS PMP & AUX FW PMP AREA CLR A-A CCS PMP & AUX FW PMP AREA CLR B-B SFP PIT PMP & TB BOOSTER PMP AREA CLR A-SFP PIT PMP & TB BOOSTER PMP AREA CLR B-1-47W859-1 1-47W859-1 1-47W859-1 1-47W859-1 1-47W859-1 1-47W859-1 1-47W845-4 1-47W845-2 1-47WB45-4 1-47W845-2 2-47W845-2 2-47W845-2 2-47W845-2 2-47W845-2 1,2-47W845-1 1,2-47W845-1 1-47W845-4 1-47W845-4 1-47W845-4 1-47WB45-4 5 5 5 5 5 5 5 5 5 5 5 5 5 5 5 5 5 5 5 5 ON ON ON ON ON ON 0 0 0 0 C C C C C C 0 O/C 0 O/C Page Number A41 ON PMP Yes 742 ON PMP Yes 742 ON PMP Yes 742 ON PMP Yes 742 ON PMP Yes 742 ON PMP Yes 742 0 TCV No 692 0 T(V No 737 0 TCV No 692 0 TCV No 737 0 MOV Yes 722 O MOV Yes 722 O MOV Yes 722 0 MOV Yes 722 0 MOV Y 722 0 MOV Y 722 0 AOV Yes 713 O AOV Yes 713 O AOV Yes 737 0 AOV Yes 737 IPS IPS IPS IPS IPS IPS C20 A02 C20 A09 IPS IPS IPS IPS DGB DGB Al Al Al Al WBN2 IPEEE DESIGN REPORT Page 152 ATTACHMENT 1

TENNESSEE VALLEY AUTHORITY Watts Bar Unit 2 Nuclear Plant IPEEE Seismic Margins Evaluation Report -Safe Shutdown Paths and Safe Shutdown Equipment List WBN-1-FCV

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-067-0068 -A EL 737 PEN RM CLR 1A-A EL 737 PEN RM CLR 1B-B ERCW HDR 1A SUPPLY ERCW HDR 1B SUPPLY SCRN WASH PUMP 1A-A HAND SWITCH SCRN WASH PUMP 1B-B HAND SWITCH ERCW SCREEN WASH PUMP 1A-A ERCW SCREEN WASH PUMP 1B-B TRAV SCRN 1A-A IN PRESS TRAV SCRN 1B-B IN PRESS ESSENTIAL RAW COOLING WATER STRAINER 1A-A ESSENTIAL RAW COOLING WATER STRAINER 18-B TRAVELING SCREEN 1A-A TRAVELING SCREEN 18-B ERCW STRAINER 2A-A BACKWASH VLV ERCW STRAINER 2A-A FLUSH VLV ERCW STRAINER 2B-B BACKWASH VLV ERCW STRAINER 2B-B FLUSH VLV EMERG DSL HTXS B2 & B2 SUP VLV HDR A EMERG DSL HTXS A2 & A2 SUP VLV HDR B 1-47WB45-4 1-47W845-4 1-47W600-144 1-47W600-144 1-47W610-67-4 1-47W610-67-4 1-47W845-1 1-47W845-1 147W610-4 147W610-4 1-47W845-1 1-47W845-1 1-47WB45-1 1-47W845-1 1-47W845-1 1-47W845: 1 1-47W845-1 1-47W845-1 1-47W845-1 1-47W845-1 Page Number A42 H 0 0 AOV Yes 737 H O/C 0 AOV Yes 737 A ON ON FT Yes AUX/692 A ON ON FT Yes AUX/692 A OFF ON/OFF HS Yes RX1 A OFF ON/OFF HS Yes AUX/723 A ON ON PMP Yes /742 A ON ON PMP Yes /742 A ON/OFF ON PS Yes 722 A ON/OFF ON PS Yes 722 A ON ON STN No 722 A ON ON STN No 722 A ON/OFF ON TWS Yes 741 A ON/OFF ON TWS Yes 741 A C 0 MOV Yes 722 A C 0 MOV Yes 722 A C 0 MOV Yes 722 A C 0 MOV Yes 722 A C 0 MOV Yes 742 A C 0 MOV Yes 742 AS AS YARD YARD IPS IPS IPS IPS IPS IPS IC 0 IPS IPS IPS IPS IPS IPS IPS IPS WBN2 IPEEE DESIGN REPORT Page 153 ATTACHMENT 1

TENNESSEE VALLEY AUTHORITY Watts Bar Unit 2 Nuclear Plant IPEEE Seismic Margins Evaluation Report -Safe Shutdown Paths and Safe Shutdown Equipment List Page Number A43 A,F 0 O/C MOV Yes 0 0 WBN-2-FCV

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-067-0130 -A LOWER CNTMT A COOLERS SUPPLY ISOL VALVE LWR CNTMT A CLRS DISCH ISOL VLV LOWER CNTMT A COOLERS DISCH ISOL VALVE LWR CNTMT D COOLERS SUPPLY ISOL VLV LWR CNTMT C CLRS DISCH ISOL VLV LOWER CNTMT C COOLERS DISCH ISOL VALVE LWR CNTMT C COOLERS SUPPLY ISOL VLV LOWER CNTMT B COOLERS SUPPLY ISOL VALVE LWR CNTMT B CLRS DISCH ISOL VLV LOWER CNTMT B COOLERS DISCH ISOL VALVE LWR CNTMT C COOLERS SUPPLY ISOL VLV LOWER CNTMT A COOLERS SUPPLY ISOL VALVE LWR CNTMT D CLRS DISCH ISOL VLV LOWER CNTMT D COOLERS DISCH ISOL VALVE LWR CNTMT D CLRS SUPPLY ISOL VLV CNTMT SPRAY HTX B SUPPLY CONTROL VALVE CONTAINMENT SPRAY HTX B DISCHARGE VALVE CNTMT SPRAY HTX A SUPPLY CONTROL VALVE CONTAINMENT SPRAY HTX A DISCHARGE VALVE UPPER CNTMT VENT CLR A SUPPLY ISOL VLV 2,2-47W845-3 2-47W945-3 2,2-47WB45-3 5 5 2-47W845-3 5 2-47W845-3 5 2,2-47W845-3 5 2-47W845-3 5 2,2-47W845-3 5 2-47W845-3 5 2,2-47W845-3 5 2-47W845-3 5 A,F A,F A,F A,F A,F A,F A,F A,F A,F A,F 0 0 0 0 0 0 0 0 0 0 O/C MOV Yes AUX/737 O/C MOV Yes AUX/737 O/C MOV Yes AUX/723 O/C MOV Yes AUX/723 O/C MOV Yes AUX/676 O/C MOV Yes AUX/676 O/C MOV Yes #N/A O/C MOV Yes AUX/692 O/C MOV Yes AUX/692 O/C MOV Yes 0 O/C MOV Yes 0 O/C MOV Yes AUX/737 O/C MOV Yes AUX/737 O/C 'MOV Yes AUX/737 0 MOV Y 0 0 MOV Y AUX/737 0 MOV Y AUX/737 0 MOV Y AUX/737 0 0 0 0 0 0#N/A 0 0 2,2-47W845-3 2-47W845-3 2,2-47W845-3 2-47W845-3 2,2-47WB45-2 2,2-47WB45-2 2,2-47W845-2 2,2-47W845-2 5 A,F 5 A,F 5 A,F 5 A,F 2,3,4 I 2,3,4 I 2,3,4 I 2,3,4 I 0 0 0 0 C C C C 0 0 0 0 0 0 0.0 2-47W845-3 5 A,F 0 O/C MOV Y AUX/737 WBN2 IPEEE DESIGN REPORT Page 154 ATTACHMENT 1

TENNESSEE VALLEY AUTHORITY Watts Bar Unit 2 Nuclear Plant IPEEE Seismic Margins Evaluation Report -Safe Shutdown Paths and Safe Shutdown Equipment List WBN-2-FCV

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TENNESSEE VALLEY AUTHORITY Watts Bar Unit 2 Nuclear Plant IPEEE Seismic Margins Evaluation Report -Safe Shutdown Paths and Safe Shutdown Equipment List WBN-2-FCV

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TENNESSEE VALLEY AUTHORITY Watts Bar Unit 2 Nuclear Plant IPEEE Seismic Margins Evaluation Report -Safe Shutdown Paths and Safe Shutdown Equipment List Page Number A46 WBN-2-HS -067-0447A -B WBN-2-HS -067-04478 -B WBN-2-PMP

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-068-0305 -A SCRN WASH PUMP 2B-B HAND SWITCH SCRN WASH PUMP 2B-B HAND SWITCH ERCW SCREEN WASH PUMP 2A-A ERCW SCREEN WASH PUMP 2B-B TRAV SCRN 2A-A IN PRESS TRAVEL SCRN 2A-A INLET PRESS TRAV SCRN 28-B IN PRESS TRAVEL SCRN 28-B INLET PRESS ERCW STRAINER 2A-A ERCW STRAINER 2B-B TRAVELING SCREEN 2A-A TRAVELING SCREEN 2A-A TRAVELING SCREEN 28-B TRAVELING SCREEN 2B-B RCS FLOW CNTLVLV WDS N2 MAN TO PRT RCSFLOW CNTL VLV WDS GA TO PRT RCS FLOW CNTL VLV WDS GA TO PRT RCS PRESSURIZER RELIEF FLOW CTRL VALVE RCS PRESSURIZER RELIEF FLOW CTRL VALVE RCS FLOW CNTL VLV WDS N2 TO PRT 1-47W620-67-4 1-47W620-67-4 1-47W845-1 1-47W845-1 2-47W620-4 1-47W620-67-4 2-47W620-4 1-47W610-67-4 1-47W845-1 1-47W845-1 1-47W845-1 1-47W845-1 1-47W845-1 1-47W845-1 2-47W830-6 2-47W813-1 2-47W813-1 1,2-47W813-1 1,2-47W813-1 2-47W620-68-6 OFF ON/OFF HS Yes OFF ON/OFF HS Yes 741 0 ON ON ON ON ON/OFF ON ON/OFF ON ON/OFF ON ON/OFF ON ON ON ON ON DN/OFI ON:)N/OFI ON JN/OFI ON)N/OFI ON C C 0 C C C O 0 0 0 PMP Yes PMP Yes PS Yes PS Yes PS Yes PS Yes STN No STN No TWS Yes TWS Yes TWS Yes TWS Yes AOV N AOV Y AOV N MOV N MOV N 0 742 742 722 722 722 722 722 722 742 742 742 742 0 0 0 0 o 0 IPS IPS IPS IPS IPS IPS IPS IPS IPS IPS IPS IPS 0 0 F ON OFF SOV Yes 0 WBN2 IPEEE DESIGN REPORT Page 157 ATTACHMENT 1

TENNESSEE VALLEY AUTHORITY Watts Bar Unit 2 Nuclear Plant IPEEE Seismic Margins Evaluation Report -Safe Shutdown Paths and Safe Shutdown Equipment List Page Number A47 WBN-2-FSV

-068-0307 -A WBN-2-FSV

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-068-0008 RCS FLOW CNTL VLV WDS GA TO PRT RCS FLOW CNTL VLV WDS GATO PRT REACTOR HEAD VENT ISOLATION VALVE REACTOR HEAD VENT ISOLATION VALVE REACTOR HEAD VENT ISOLATION VALVE REACTOR HEAD VENT ISOLATION VALVE RCS PRESSURIZER LEVEL TRANSMITTER RCS PRESSURIZER LEVEL TRANSMITTER RCS PRESSURIZER LEVEL TRANSMITTER RVLIS DYNAMIC HEAD RANGE (RCPS ON)RVLIS LOWER RANGE (RCPS OFF)RVLIS UPPER RANGE (RCPS OFF)RVLIS DYNAMIC HEAD RANGE (RCPS ON)RVLIS LOWER RANGE (RCPS OFF)RVLIS UPPER RANGE (RCPS OFF)PZR BACKUP HTR GRP 2A-A TRANSFORMER PZR BACKUP HTR GRP 2B-B TRANSFORMER PRESSURIZER POWER OPERATED RELIEF VALVE PRESSURIZER POWER OPERATED RELIEF VALVE REACTOR COOLANT PUMP 2 2-47W620-68-6 2-47W620-68-6 2-47W813-2 2-47W813-2 2-47W813-2 2-47W813-2 2-47W813-1 2-47W813-1 2-47W813-1 2-47W600-287 5 5 5 5 5 5 5 5 5 2, ON ON C/o C/o C/O C/O ON ON ON ON ON ON ON ON ON ON ON C C ON OFF SOV Yes 0 OFF SOV Yes AUX/713 C SOV N 0 C SOV N 0 C SOV N 0 C SOV N 0 ON LT Y 0 ON LT Y 0 ON LT Y 0 ON LT Y 0 ON LT Y 0 ON LT Y 0 ON LT Y 0 ON LT Y 0 ON LT Y 0 ON XFMR Yes 782 ON XFMR Yes 782 O/C SOV Yes 783'20 O/C SOV Yes 783'20 OFF VP N 695 2-47W600-287 2, 2-47W600-287 2, 2-47W600-287 2, 2-47W600-287 2, 2-47W600-287 2-25E500-2 2-25E500-2 2-47W813-1 2-47W813-1 1,2-47W813-1 2, 5 5 5 5 5 WBN2 IPEEE DESIGN REPORT Page 158 ATTACHMENT 1

TENNESSEE VALLEY AUTHORITY Watts Bar Unit 2 Nuclear Plant IPEEE Seismic Margins Evaluation Report -Safe Shutdown Paths and Safe Shutdown Equipment List Page Number A48 WBN-2-PMP

-068-0031 WBN-2-PMP

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-068-0073 WBN-2-PRES-068-PR WBN-2-PT -068-0070 -G WBN-2-PT -068-0322 -G WBN-2-PT -068-0323 -F WBN-2-PT -068-0334 -E WBN-2-PT -068-0340 -D WBN-2-RFV

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-068-U2 WBN-2-SGEN-068-SG1 WBN-2-SGEN-068-SG2 WBN-2-SGEN-068-SG3 WBN-2-SGEN-068-SG4 WBN-2-TANK-068-PRT WBN-2-TE -068-0001 -D WBN-2-TE -068-0018 -D REACTOR COOLANT PUMP 2 REACTOR COOLANT PUMP 3 REACTOR COOLANT PUMP 4 UNIT 2 PRESSURIZER RCS WIDE RANGE PRESSURE LOOP 4 HOT LEG RCS PRESSURIZER PRESSURE TRANSMITTER RCS PRESSURIZER PRESSURE TRANSMITTER RCS PRESSURIZER PRESSURE TRANSMITTER RCS PRESSURIZER PRESSURE TRANSMITTER PRESS. SAFETY VALVE PRESS SAFETY VALVE PRESS SAFETY VALVE REACTOR VESSEL STEAM GENERATOR 1 STEAM GENERATOR 2 STEAM GENERATOR 3 STEAM GENERATOR 4 PRESSURIZER RELIEF TANK RCS LOOP 1 HOT LEG TEMP RCS LOOP 1 COLD LEG TEM P 1,2-47W813-1 1,2-47W813-1 1,2-47W813-1 1-47W813-1 2-47W813-1 2-47W813-1 2-47W813-1 2-47W813-1 2-47W813-1 2-47W813-1 2-47W813-1 2-47WB13-1 1-47W813-1 2-47W813-1 2-47W813-1 2-47W813-1 2-47W813-1 2-47W813-1 2-47W813-1 2-45N 16 16-8 C,D C C C C C C,E C,E C,E ON OFF VP N 585 ON OFF VP N 685 ON OFF VP N 685 ON ON PRES No 703 ON ON PT Y 0 ON ON PT Yes 716 ON ON PT Y 0 ON ON PT Y 0 ON ON PT Y 0 C C RFV No 780 C C RFV No 780 C C RFV No 780.. .. RPV N 0.. .. SGEN N 0.. .. SGEN N 0.. .. SGEN *N 0.. .. SGEN N 0.. .. TNK No 703 ON ON TE Y 0 ON ON TE Y 0 0 0 0 0 0 0 0 0 0 IC IC IC 5 C,E 5 C,E 0 ' C,E WBN2 IPEEE DESIGN REPORT Page 159 AI-FACHMENT 1

TENNESSEE VALLEY AUTHORITY Watts Bar Unit 2 Nuclear Plant IPEEE Seismic Margins Evaluation Report -Safe Shutdown Paths and Safe Shutdown Equipment List Page Number A49 WBN-2-TE -068-0024 -D WBN-2-TE -068-0041 -D WBN-2-TE -068-0043 -E WBN-2-TE -068-0060 -E WBN-2-TE -068-0065 -E WBN-2-TE -068-0083 -E WBN-2-TE -068-0373 -D WBN-2-TE -068-0376 -D WBN-2-TE -068-0377 -D WBN-2-TE -068-0378 -D WBN-2-TE -068-0379 -D WBN-2-TE -068-0380 -E WBN-2-TE -068-0383 -E WBN-2-TE -068-0384 -E WBN-2-TE -068-0385 -E WBN-2-TE -068-0386 -E WBN-2-TE -068-0393 -E WBN-0-HTX

-070-0186 WBN-0-PMP

-070-0051 -S WBN-0-PT -070-0221 RCS LOOP 2 HOT LEG TEMP RCS LOOP 2 COLD LEG TEM P RCS LOOP 3 HOT LEG TEMP RCS LOOP 3 COLD LEG TEM P RCS LOOP 4 HOT LEG TEMP RCS LOOP 4 COLD LEG TEM P REACTOR LEVEL CAPILLARY TUBE TEMP COMP REACTOR LEVELTEMP COMP GUIDE TUBE REACTOR LEVEL CAPILLARY TUBE TEMP COMP REACTOR LEVEL CAPILLARY TUBE TEMP COMP REACTOR LEVEL CAPILLARY TUBE TEMP COMP REACTOR LEVEL CAPILLARY TUBE TEMP COMP REACTOR LEVEL TEMP COMP GUIDE TUBE REACTOR LEVEL CAPILLARY TUBE TEM COMP REACTOR LEVEL CAPILLARY TUBE TEMP COMP REACTOR LEVEL CAP TUBE TEMP COMP (HEAD)REACTOR LEVEL CAPILLARY TUBE TEMP COMP COMPONENT COOLING HX C CCS PUMP C-S CCS HTX C INLET PRESSURE 2-47W813-1 5 C.E ON ON TE Y 0 0 2-45N1616-8 2-47W813-1 2-45N1616-8 2-47W813-1 2-47B601-068 2-47W610-68-7 2-47W610-68-7 2-47W610-68-7 2-47W610-68-7 2-47W610-68-7 2-47W610-68-7 2-47W610-68-7 2-47W610-68-7 2-47W610-68-7 2-47W610-7 2-47W610-68-7 1-47W859-1 1-47W859-1 1-47W859-1 0 C,E 5 C,E 0 C,E 5 C,E 0 C,E 5 A ON ON ON ON ON ON ON ON ON ON ON ON ON ON ON ON TE Y 0 TE Y 0 TE Y 0 TE Y 0 TE Y 0 TE Yes 702 TE Yes 716 TE Yes 702 5 5 A A 5 A ON ON TE Yes 702'9" 5 A ON ON TE Yes 702 5 A ON ON TE Yes 702 5 A ON ON TE Yes 674'6" 5 A ON ON TE Yes 702 5 A ON ON TE Yes 702'9" 5 A ON ON TE Yes 702 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 Al A3S Al 5 5 5 S A ON ON A ON ON A ON ON A ON ON TE Yes 702 HTX No 737 PMP Yes 690 PT Yes 723 WBN2 IPEEE DESIGN REPORT Page 160 ATTACHMENT 1

TENNESSEE VALLEY AUTHORITY Watts Bar Unit 2 Nuclear Plant IPEEE Seismic Margins Evaluation Report -Safe Shutdown Paths and Safe Shutdown Equipment List WBN-0-TE -070-0162 -B WBN-2-FCV

-070-0066 WBN-2-FCV

-070-0085 -B WBN-2-FCV

-070-0087 -B WBN-2-FCV

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-070-0081 -B WBN-2-FT -070-0159A -A WBN-2-FT -070-0165A -B WBN-2-FT -070-0215A -A CCS HEAT EXCHANGER C OUTLET TEMP CCS SURGE TANK VENT VALVE EXCESS LETDOWN HTX OUTLET VALVE RC PUMP THERM BARRIER RET CNTNMT ISOL RC PUMP OIL CLR RET CNTNMT ISOL VALVE RC PUMP THERM BARRIER RET CNTNMT ISOL RC PUMP OIL CLR RET CNTNMT ISOL VALVE RC PUMP OIL CLR HDR CONT ISOL VALVE RC PUMP THERM BARRIER CONT ISOL VALVE RC PUMP THERM BARRIER CONT ISOL VALVE RC PUMP OIL CLR HDR CONT ISOL VALVE EXCESS LETDOWN HTX CONT INLET ISOL VLV RHR HTX B OUTLET VALVE RHR HTX A OUTLET VALVE SAMPLE HTX HDR OUTLET VALVE SAMP HTX INLET THERMAL BARRIER CCS HDR FLOW 2A ESF EQUIPMENT CCS SUPPLY HEADER FLOW 2B ESF EQUIPMENT CCS SUPPLY HDR FLOW SAMPLE HTX HDR INLET FLOW 1-47W610-70-1 2,3,4 A ON 1,2-47W859-3 5 A 0 Page Number A50 ON TE Yes 737 0 AOV N 0 Al 1,2-47W859-3 1,2-47W859-3 1,2-47WB59-3 1,2-47W859-3 1,2-47W859-3 2-47W859-2 1,2-47W859-2 1,2-47W859-2 1,2-47W859-2 2-47W859-2 5 F 5 A,F 5 A,F 5 A,F 5 A,F 5 A,F 5 A 5 A,F 5 A,F 5 F C C AOV N YRD/729 0 O O/C MOV Y 0 0 0 O/C MOV Y 0 0 O O/C MOV Y AUX/713 A28 0 O/C MOV Y AUX/713 A19 O O/C MOV Yes RXB/716 :UMULATOR

  1. 4 O 0 MOV N AUX/741 0 O O/C MOV Y AUX/737 A9 O O/C MOV Y AUX/713 A29 C C MOV Y AUX/713 A2 0 0 MOV Y AUX/713 Al C 0 MOV Y AUX/713 Al 1,2-47W859-4 2,3,4 A,E 1,2-47W859-4 2,3,4 A,E 1,2-47W859-4 5 A 0 2-47W859-2 5 A 0 2-47W620-70-3 5 A ON 2-47W859-2 5 A ON 2-47W859-2 5 A ON 2-47W610-70-2 5 A ON 0 MOV N AUX/713 A14 O/C MOV Yes AUX/713 0 ON FIS Yes RXB/703 0 ON FT Yes 0 ON FT Yes ON FT Yes 0 0 WBN2 IPEEE DESIGN REPORT Page 161 ATTACHMENT 1

TENNESSEE VALLEY AUTHORITY Watts Bar Unit 2 Nuclear Plant IPEEE Seismic Margins Evaluation Report -Safe Shutdown Paths and Safe Shutdown Equipment List Page Number A51 WBN-2-FT -070-0215B -A WBN-2-HTX

-070-0185 WBN-2-LCV

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-072-0045 -B SAMPLE HTX HDR OUTLET FLOW COMPONENT COOLING HX A CCS SURGE TANK DEMIN WATER INLET VALVE UNIT 2 CCS SURGE TANK 2A SIDE LEVEL UNIT 2 CCS SURGE TANK 2B SIDE LEVEL CCS PUMP 2B-B CCS PUMP 2A-A CCS THERMAL BARRIER BOOSTER PUMP 2B-B CCS THERMAL BARRIER BOOSTER PUMP 2A-A CCS HTX A INLET PRESSURE COMPONENT COOLING WATER SURGE TANK CCS HEAT EXCHANGER A OUTLET TEMP CONTAINMENT SPRAY HDR B ISOLATION VALVE RWST TO SPRAY HDR B FLOW CONTROL VALVE RWST TO SPRAY HDR A FLOW CONTROL VALVE CNTMT SPRAY HDR A ISOLATION VLV RHR SPRAY HEADER A ISOLATION VALVE RHR SPRAY HEADER B ISOLATION VALVE CNTMT SUMP SPRAY HDR A FLOW CONTROL VLV CNTMT SUMP SPRAY HDR B FLOW CONTROL VLV 2-47W610-70-2 2-47W859-1 1,2-47W859-1 2-47W859-1 2-47W859-1 1,2-47W859-1 1,2-47W859-1 1,2-47W859-2 1,2-47W859-2 2-47W859-1 2-47W859-1 2-47W610-70-1 1,2-47W812-1 1,2-47W813-1 1,2-47W813-1 1,2-47W813-1 1,2-47W813-1 1,2-47W813-1 2-47W812-1 2-47W812-1 5 5 5 5 5 5 5 5 5 5 5 2,3,4 2,3,4 2,3,4 2,3,4 2,3,4 2,3,4 2,3,4 5 ON FT Yes 0 ON HTX No AUX/729 AI0 O/C AOV N 0 0 ON LT Yes 737 A2 ON LT Yes 737 A2 ON PMP Y 690 CORRIDOR ON PMP Y 690 CORRIDOR ON PMP Y 724 0 ON PMP Y 724 0 ON P1 Yes 713 Al ON TANK No 757 0 ON TE Yes 737 A12 O/C MOV Y AUX/737 A9 C MOV Y AUX/676 0 O/C MOV Y AUX/676 0 O/C MOV Y AUX/737 A9 O/C MOV Y AUX/713 A29 0/C MOV Y AUX/713 A29 O/C MOV Yes AUX/676 A13 5. I,F C O/C MOV Yes AUX/676 A12 WBN2 IPEEE DESIGN REPORT Page 162 ATTACHMENT 1

TENNESSEE VALLEY AUTHORITY Watts Bar Unit 2 Nuclear Plant IPEEE Seismic Margins Evaluation Report -Safe Shutdown Paths and Safe Shutdown Equipment List Page Number A52 WBN-2-FT -072-0013 -G WBN-2-FT -072-0034 -F WBN-2-HTX

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-074-0032 WBN-2-FCV

-074-0033 -A WBN-2-FCV

-074-0035 -B WBN-2-FE -074-0012 CONTAINMENT SPRAY HEADER B MINI FLOW CTL CONTAINMENT SPRAY HEADER A MINI FLOW CTL CONTAINMENT SPRAY HEAT EXCHANGER 2A CONTAINMENT SPRAY HEAT EXCHANGER 2B CONTAINMENT SPRAY PUMP 2B-B CONTAINMENT SPRAY PUMP 2A-A RHR SYSTEM ISOLATION VALVE RHR SYSTEM ISOLATION VALVE RHR PUMP 2A-A INLET FLOW CONTROL VALVE RHR SYSTEM ISOLATION VALVE RHR SYSTEM ISOLATION VALVE RHR PUMP 2A-A MINIMUM FLOW VALVE RHR HT EX A OUTLET FLOW CONTROL VALVE RHR PUMP 2B-B INLET FLOW CONTROL VALVE RHR PUMP 2B-B MINIMUM FLOW VALVE RHR HT EX B OUT FLOW CONTROL VALVE RHR HT EX A/B BYPASS FCV RHR HT EX A BYPASS RHR HT EX B BYPASS RHR FLOW INDICATOR, TRAIN A 2-47W812-1 2-47WB12-1 2-47W812-1 2-47W812-1 1,2-47W812-1 1,2-47WB12-1 2-47W810-1 2,3,4 1 ON 2,3,4 I ON 2,3,4 --2,3,4 --2,3,4 1 OFF 2,3,4 1 OFF 2,3,4 G,E C 2-47W810-1 2,3,4 G,E,F C 2-47W810-1 2,3,4 B,D,E 0 2-47W810-1 2-47W810-1 2-47W810-1 1,2-47W810-1 2-47WB10-1 2-47W810-1 1,2-47W810-1 2-47W810-1 2-47W810-1 2-47W810-1 2-47W820-2 2,3,4 G,E,F C 2,3,4 G,E,F C 2,3,4 B,D,E 0 2,3,4 B,D,E 0 2,3,4 B,D,E 0 2,3,4 B,D,E 0 2,3,4 B,D,E 0 2,3,4 D,E C 2,3,4 D,E 0 2,3,4 D,E 0 2,3,4 D,E ON ON FT Yes 0 ON FT Yes 0-- HTX No AUX/729 A10-HTX No AUX/729 A10 ON HP Y 653 #N/A ON HP Y 653 #N/A C MOV N #N/A #N/A C MOV N #N/A #N/A 0 MOV N AUX/782 A12 C/O MOV Yes AUX/782 0 C/O MOV Yes AUX/782 0 O/C MOV Y AUX/782 0 O/C AOV Y 0 0 0 MOV N 0 A13 O/C MOV Y /729 0 O/C AOV

  • Y AUX/729 All O/C AOV Yes AUX/723 0 O MOV N 0 0 0 MOV N AUX/737 0 ON FE Y AUX/692 A20 WBN2 IPEEE DESIGN REPORT Page 163 Ai-FACHMENT 1

TENNESSEE VALLEY AUTHORITY Watts Bar Unit 2 Nuclear Plant IPEEE Seismic Margins Evaluation Report -Safe Shutdown Paths and Safe Shutdown Equipment List WBN-2-FE -074-0024 WBN-2-FIS

-074-0012 -A WBN-2-FIS

-074-0024 -B WBN-2-HTX

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-074-0010 -A WBN-2-PMP

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-077-0009 -B RHR FLOW INDICATOR, TRAIN B RHR PMP 2A-A MIN FLOW VALV SW RHR PMP 2B-B MIN FLOW VALV SW RHR PUMP 2A-A SEAL HEAT EXCHANGER RHR PUMP 2B-B SEAL HEAT EXCHANGER RHR HEAT EXCHANGER 2A RHR HEAT EXCHANGER 2B RHR PUMP 2A-A RHR PUMP 2B-B WASTE GAS COMPRESSOR A HEAT EXCHANGER WASTE GAS COMPRESSOR B HEAT EXCHANGER R C DRAIN TANK FLOW CNTL VALVE R C DRN TNK GAS ANALY FLW CON R C DRN TNK GAS ANALY FLW CON R C DRN TNKTOVENT HDR ISOL R C DRN TNK TO VENT HDR ISOL REACT COOLANT DRAIN TANK N2 SUPPLY FLOW REACT BLDG SUMP DISCH FLOW CONTROL VA REACT BLDG SUMP DISCH FLOW CONTROL VA R C DRAIN TANK FLOW CNTL VALVE 2-47W820-2 47W620-74-2 47W620-74-2 2-47W859-4 2-47W810-1 2-47W859-4 2-47W810-1 2-47W810-1 2-47W810-1 1-47W830-4 1-47W830-4 2-47W830-1 2-47W830-1" 2-47W830-1 2-47W830-1 2-47W830-1 2,3,4 D,E ON ON 2,3,4 B,D,E ON ON 2,3,4 B,D,E ON ON 2,3,4 B,D,E --2,3,4 B,D,E --2,3,4 B,D,E --2,3,4 B,D,E -- -2,3,4 B,D,E OFF ON 2,3,4 B,D,E OFF ON 5 A .. ..5 A .. ..5 F 0 C 5 F. 0 C 5 F 0 C S F 0 C 5 F 0 C Page Number A53 FE Y RXB/703 0 FIS Yes RXB/703 0 FIS Yes RXB/703 0 HTX N AUX/729 A10 HTX N AUX/729 A1O HTX N AUX/729 A1O HTX N AUX/729 A1O PMP Y 653 A12 PMP Y 653 A13 HTX No 723 A26 HTX No 723 A26 AOV Yes , 0 0 AOV Yes 0 ,0 AOV Yes 0 A29 AOV Yes 0 .0 AOV Yes /728 0 0 N AUX/676 A29 AOV Yes AUX/676 0 1,2-47W830-2 5 A C C 1-47W852-1 5 F 0 C 1-47W852-1 5 F 0 C AOV Yes AUX/676 A29 2-47W620-77-4 5 F ON OFF SOV Yes 0 WBN2 IPEEE DESIGN REPORT Page 164 ATTACHMENT 1

TENNESSEE VALLEY AUTHORITY Watts Bar Unit 2 Nuclear Plant IPEEE Seismic Margins Evaluation Report -Safe Shutdown Paths and Safe Shutdown Equipment List WBN-2-FSV

-077-0016 -B WBN-2-FSV

-077-0017" -A WBN-2-FSV

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-078-0032 WBN-0-PMP

-078-0009 -B WBN-0-PMP

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-081-0012 -A WBN-2-FSV

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/F -A WBN-1-CDPL-082-B

/F -B WBN-1-DIEG-082-A1 -A WBN-1-DIEG-082-A2 -A WBN-1-DIEG-082-B1 -B WBN-1-DIEG-082-B2 -B WBN-1-DPL

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-082-0001A -A RCDT TO GAS ANALYZER FLOW SOL VALVE R C DRAIN TK GAS ANALYZ FLOW R C DRAIN TK TO VENT HDR R C DRAIN TANK FLOW CNTL VALVE SPENT FUEL PIT HEAT EXCHANGER A SPENT FUEL PIT HEAT EXCHANGER B SPENT FUEL PMP SPENT FUEL PMP PW RCS PRESS RELF TNK & RCP STANDPIPES PW RCS PRESS RELF TNK & RCP STANDPIPES DG 1A-A PROTECT RELAY PNL 2 DIESEL GENERATOR lA-A CNTL DISTRIBUTION PNL DIESEL GENERATOR 1B-B CNTL DISTRIBUTION PNL DIESEL ENGINE 1A1 DIESEL GENERATOR ENGINE 1A2 DIESEL GENERATOR ENGINE 181 DIESEL GENERATOR ENGINE 182 DG 1A-A 125V DC DISTRIBUTION PANEL DG 1B-B 125V DC DISTRIBUTION PANEL DIESEL GENERATOR 1A-A 2-47W620-77-4 2-47W620-77-4 2-47W620-77-4 2-47W620-77-4 1-47W855-1 1-47W855-1 1-47W855-1 1-47W855-1 1,2-47W819-1 2-47W819-1 NA NA NA 1-47W839-1 1-47W839-1 1-47W839-1A 1-47W839-1A NA NA 1-47W839-18 Page Number A54 5 F ON OFF SOV Yes RXB/780 IC 5 F ON OFF SOV Yes /780 IC 5 F ON OFF SOV Yes RXB/780 IC 5 F ON OFF SOV Yes 0 5 A -- HTX No 737 AW6 5 A -- HTX No 737 AW6 5 A ON ON PMP Yes 737 A02 5 A ON ON PMP Yes 737 A01 5 F O/C C AOV Y AUX/723 A4W 5 F OFF/ON OFF SOV Yes 0 5 A ON ON PNL Yes 742 DGB 5 A ON ON PNL Yes 760 DGB 5 A ON ON PNL Yes 760 DGB 5 A OFF ON DIEG Yes 742 1A-A 5 A OFF ON DIEG Yes 742 1A-A 5 A OFF ON DIEG Yes 742 1B-B 5 A OFF ON DIEG Yes 742 1B-B 5 A ON ON PNL Yes 742 DGB 5 A ON ON PNL Yes 742 DGB 5 A OFF ON GEN Yes AUX/692 DGB WBN2 IPEEE DESIGN REPORT Page 165 AI-iACHM ENT 1 TENNESSEE VALLEY AUTHORITY Watts Bar Unit 2 Nuclear Plant IPEEE Seismic Margins Evaluation Report -Safe Shutdown Paths and Safe Shutdown Equipment List WBN-1-GEN

-082-0001B -B WBN-1-IACL-082-0101 WBN-1-IACL-082-0102 WBN-1-IACL-082-0103 WBN-1-IACL-082-0104 WBN-1-PNL-082-A -A WBN-1-PNL

-082-B -B WBN-1-SILN-082-0101 -A WBN-1-SILN-082-0102 -A WBN-1-SILN-082-0103 -B WBN-1-SILN-082-0104 -B WBN-1-SILN-082-0105 -A WBN-1-SILN-082-0106 -A WBN-1-SILN-082-0107 -B WBN-1-SILN-082-0108 -B WBN-2-ARB

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/F -B WBN-2-DIEG-082-A1 -A DIESEL GENERATOR 1B-B DG EN 1A1 INTAKE AIR CLEANER DG EN 1A2 INTAKE AIR CLEANER DG EN 1B1 INTAKE AIR CLEANER DG EN 1B2 INTAKE AIR CLEANER DIESEL GENERATOR CONTROL BOARD DIESEL GENERATOR CONTROL BOARD DG ENG 1A1 INTAKE SILENCER DG ENG 1A2 INTAKE SILENCER DG ENG 1B1 INTAKE SILENCER DG ENG 1B2 INTAKE SILENCER DG ENG 1A1 EXHST SILENCER DG ENG 1A2 EXHST SILENCER DG ENG 1B EXHST SILENCER DG ENG 1B EXHST SILENCER DG 2A-A PROTECT RELAY PNL 2 DG 2B-B PROTECT RELAY PNL 2 DIESEL GENERATOR 2A-A CNTL DISTRIBUTION PNL DIESEL GENERATOR 2B-B CNTL DISTRIBUTION PNL DIESEL GENERATOR ENGINE 2A1 1-47W839-1C 1-47W839-1 1-47W839-1 1-47W839-1A 1-47W839-1A NA NA 1-47W839-1 1-47W839-1 1-47W839-1A 1-47W839-1A 1-47W839-1 1-47W839-1 1-47W839-1A 1-47W839-1A NA Page Number ASS 5 A OFF ON GEN Yes AUX/692 5 A .. .. FLTR NA 760.5 5 A -- -FLTR NA 760.5 5 A -- -FLTR NA 760.5 S A .. .. FLTR NA 760.5 5 A ON ON PNL Yes AUX/729 5 A ON ON PNL Yes AUX/729 5 A .. .. SILN NA 760.5 5 A .. .. SILN NA 760.5 5 A -- -SILN NA 760.5 5 A -- -SILN NA 760.5 5 A -- -SILN NA 760.5 5 A -- -SILN NA 760.5 5 A .. .. SILN NA 760.5 5 A .. .. SILN NA 760.5 5 A ON ON PNL Yes 742 DGB DGS DG5 DG11 DG11 DGB DGB DG5 DG5 DG11 DG11 DG5 DG5 DG11 DG11 DGB 0 DGB DGB.2A-A NA NA 5 5 A O ON ON PNL Yes #N/A A ON ON PNL Yes 760 A ON ON PNL Yes 760 A OFF ON DIEG Yes AUX/713 NA 5 1-47W839-1B 5 WBN2 IPEEE DESIGN REPORT Page 166 ATTACHMENT 1

TENNESSEE VALLEY AUTHORITY Watts Bar Unit 2 Nuclear Plant IPEEE Seismic Margins Evaluation Report -Safe Shutdown Paths and Safe Shutdown Equipment List WBN-2-DIEG-082-A2 -A WBN-2-DIEG-082-B1 -B WBN-2-DIEG-082-B2 -B WBN-2-DPL

-082-A -A WBN-2-DPL

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-082-B -B WBN-2-SILN-082-0201 -A WBN-2-SILN-082-0202 -A WBN-2-SILN-082-0203 -B WBN-2-SILN-082-0204 -B WBN-2-SILN-082-0205 -A WBN-2-SILN-082-0206 -A WBN-2-SILN-082-0207 -B DIESEL GENERATOR ENGINE 2A2 DIESEL GENERATOR ENGINE 2B1 DIESEL GENERATOR ENGINE 2B2 DG 2A-A 125V DC DISTRIBUTION PANEL DG 2B-B 125V DC DISTRIBUTION PANEL DIESEL GENERATOR 2A-A DIESEL GENERATOR 2B-B DG EN 2A2 INTAKE AIR CLEANER DG EN 2A2 INTAKE AIR CLEANER DG EN 2B1 INTAKE AIR CLEANER DG EN 2B2 INTAKE AIR CLEANER DG 2A-A CONT. BRD DG 2B-B CONT BRD.DG ENG 2A1 INTK SILENCER DG ENG 2A2 INTK SILENCER DG ENG 211 INTK SILENCER DG ENG 2B2 INTK SILENCER DG ENG 2A1 EXHST SILENCER DG ENG 2A2 EXHST SILENCER DG ENG 2B1 EXHST SILENCER 1-47W839-1B 1-47W839-1C 1-47W839-1C NA NA 1-47W839-1B 1-47W839-1C 1-47W839-1B 1-47W839-1B 1-47W839-1C 1-47W839-1C NA NA 1-47W839-1B 1-47W839-1B 1-47W839-1C 1-47W839-1C 1-47W839-1B 1-47W839-1B 1-47W839-1C A A A A A A A A A A A A A A A A A A A A OFF OFF OFF ON ON ON ON ON ON ON Page Number A56 DIEG Yes AUX/713 DIEG Yes 0 DIEG Yes AUX/676 PNL Yes DGB/742 PNL Yes DGB/742 OFF ON GEN Yes 742 OFF ON GEN Yes 742-- FLTR NA 760.5-- FLTR NA 760.5 FLTR NA 760.5 FLTR NA 760.5 ON ON PNL Yes 742 ON ON PNL Yes 742-- SILN NA 760.5-- SILN NA 760.5-- SILN NA 760.5-- SILN NA 760.5-- SILN NA 760.5-- -SILN NA 760.5.. .. SILN NA 760.5 D D D D D 2A-A 2B-B DGB DGB 0 0 DG8 DG8 DG14 DG14 DGB DGB DG8 DG8 GD14 DG14 DG8 DG8 DG14 WBN2 IPEEE DESIGN REPORT Page 167 ATTACHMENT 1

TENNESSEE VALLEY AUTHORITY Watts Bar Unit 2 Nuclear Plant IPEEE Seismic Margins Evaluation Report -Safe Shutdown Paths and Safe Shutdown Equipment List WBN-2-SILN-082-0208 -B WBN-2-ROD

-085-Bl0 WBN-2-ROD

-085-B12 WBN-2-ROD

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-085-13 WBN-2-ROD

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-085-L3 DG ENG 282 EXHST SILENCER SHTDN BK B FULL LG ROD DRIVE MECH SHTDN BK A FULL LG ROD DRIVE MECH SHTDN BK A FULL LG ROD DRIVE MECH SHTDN BK B FULL LG ROD DRIVE MECH SHTDN BK C FULL LG ROD DRIVE MECH SHTDN BK C FULL LG ROD DRIVE MECH SHTDN BK D FULL LG ROD DRIVE MECH SHTDN BK B FULL LG ROD DRIVE MECH SHTDN BK B FULL LG ROD DRIVE MECH SHTDN BK A FULL LG ROD DRIVE MECH SHTDN BK A FULL LG ROD DRIVE MECH SHTDN BK D FULL LG ROD DRIVE MECH SHTDN BK C FULL LG ROD DRIVE MECH SHTDN BK B FULL LG ROD DRIVE MECH SHTDN BK B FULL LG ROD DRIVE MECH SHTDN BK B FULL LG ROD DRIVE MECH SHTDN BK B FULL LG ROD DRIVE MECH SHTDN BK C FULL LG ROD DRIVE MECH SHTDN BK D FULL LG ROD DRIVE MECH 1-47W839-1C N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N / A N/A N/A Page Number A57.. .. SILN NA 760.5 DG14 On OFF ROD No On OFF' ROD No On OFF ROD No On OFF ROD No On OFF ROD No On OFF ROD No On OFF ROD No On OFF ROD No On OFF ROD No On OFF ROD No On OFF ROD No On OFF ROD No On OFF ROD No On OFF ROD No On OFF ROD No On On On On OFF OFF OFF OFF ROD No ROD No ROD No ROD No WBN2 IPEEE DESIGN REPORT Page 168 ATTACHMENT 1

TENNESSEE VALLEY AUTHORITY Watts Bar Unit 2 Nuclear Plant IPEEE Seismic Margins Evaluation Report -Safe Shutdown Paths and Safe Shutdown Equipment List Page Number A58 WBN-2-ROD

-085-M14 SHTDN BK A FULL LG ROD DRIVE MECH N/A 5 B On OFF ROD No WBN-2-ROD

-085-M2 WBN-2-ROD

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-090-0117 -A WBN-2-RE -090-0271 -A WBN-2-RE -090-0272 -B SHTDN BK A FULL LG ROD DRIVE MECH SHTDN BK D FULL LG ROD DRIVE MECH SHTDN BK C FULL LG ROD DRIVE MECH SHTDN BK B FULL LG ROD DRIVE MECH SHTDN BK B FULL LG ROD DRIVE MECH SHTDN BK A FULL LG ROD DRIVE MECH SHTDN BK A FULL LG ROD DRIVE MECH CNTNMT BLDG LWR COMPT MON ISOL VALVE CNTNMT BLDG LWR COMPT MON ISOL VALVE CNTNMT BLDG LWR COMPT MON ISOL VALVE CNTNMT BLDG LWR COMPT MON ISOL VALVE CNTNMT BLDG LWR COMPT MON ISOL VALVE CNTNMT BLDG UPR COMPT MON ISOL VALVE CNTNMT BLDG UPR COMPT MON ISOL VALVE CNTNMT BLDG UPR COMPT MON ISOL VALVE CNTNMT BLDG UPR COMPT MON ISOL VALVE CNTNMT BLDG UPR COMPT MON ISOL VALVE UPPER INS CONTMT POST A CD AREA MONITOR UPPER INS CONTMT POST A CD AREA MONITOR N/A N/A N/A N/A N/A N/A N/A 1,2-47W610-90-3 1,2-47W610-90-3 1,2-47W610-90-3 1,2-47W610-90-3 1,2-47W610-90-3 1,2-47W610-90-3 1,2-47W610-90-3 1,2-47W610-90-3 1,2-47W610-90-3 1,2-47W610-90-3 2-45W610-4 2-45W610-4 5 5 5 5 5 5 5 5 5 5 5 5 5 5 On On On On on On On 0 0 0 0 0 0 0 OFF OFF OFF OFF OFF OFF OFF C C C C C C C ROD No ROD No ROD No ROD No ROD No ROD No ROD No AOV N AUX/676 AOV AOV AOV AOV AOV AOV N 0 N AUX/692 N AUX/692 N AUX/723 N AUX/713 N AUX/713 0 0 0 0 0 Al A01 A22 A23 A19 5 F 0 C AOV N AUX/692 5 F 0 C AOV N AUX/692 5 F 0 C AOV N AUX/692 5 A ON ON MON Y 0 5 A ON ON MON Y 0 WBN2 IPEEE DESIGN REPORT Page 169 ATTACHMENT 1

TENNESSEE VALLEY AUTHORITY Watts Bar Unit 2 Nuclear Plant IPEEE Seismic Margins Evaluation Report -Safe Shutdown Paths and Safe Shutdown Equipment List Page Number AS9 WBN-2-RE -090-0273 -A LOWER INS CONTNT POST A CD AREA MONITOR 2-45W610-4 5 A ON ON MON Y 0 0 WBN-2-RE -090-0274 -B WBN-2-NM -092-0131 -D WBN-2-NM -092-0132 -E WBN-2-NM -092-0138 -D WBN-2-PNL

-099-L116 WBN-2-PNL

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-099-R5 -E LOWER INS CONTNT POST A CD AREA MONITOR CH 2 NEUT MON SOURCE/INTER RNG AMP CH 2 NEUT MON SOURCE/INTER RNG AMP CHI NEUT MON APP R ISOLATOR REACTOR TRIP SW GEAR PNL (W-92250L3)

REACTOR PROT SET I PROCESS INSTRUMENTATION REACTOR PROT SET III PROCESS INSTRUMENTATION REACTOR PROT SET III PROCESS INSTRUMENTATION REACTOR PROT SET IV PROCESS INSTRUMENTATION REACTOR PROT SET IV PROCESS INSTRUMENTATION REACTOR PROT SET I PROCESS INSTRUMENTATION REACTOR PROT SET IV PROCESS INSTRUMENTATION REACTOR PROT SET I PROCESS INSTRUMENTATION REACTOR PROT SET I PROCESS INSTRUMENTATION SOLID STATE PROT SYSTEM TRAIN A INPUT PANEL SOLID STATE PROT SYSTEM TRAIN A LOGIC PANEL SS PROT SYSTEM INPUT OUTPUT PANEL TRAIN A SOLID STATE PROT SYSTEM TRAIN B INPUT PANEL REACTOR PROT SET II PROCESS INSTRUMENTATION 2-45W610-4 2-47W620-92-1 2-47W620-92-1 2-47W620-92-1 5 5 5 5 45W600-99-1 5 2-47W610-99-1 S 2-47W610-99-1 2-47W610-99-1 2-47W610-99-1 2-47W610-99-1 2-47W610-99-1 2-47W610-99-1 2-47W610-99-1 2-47W610-99-1 2-47W610-99-1 2-47W610-99-1 2-47W610-99-1 2-47W610-99-1 2-47W610-99-1 5 5 5 S 5 5 5 5 5 5 5 5 A A A A B A A A A A A A A A A A A A ON ON ON ON ON ON ON ON ON ON ON ON ON ON ON ON ON ON ON ON ON ON ON ON ON ON ON ON ON ON ON ON ON ,ON ON ON MON Y AMP Yes AMP Yes ISO Yes PNL Yes PNL Yes PNL Yes PNL Yes 0 729 737 729 PNL Yes 782 PNL Yes 708 708 708 708 708 0 A24 A7 A24 0 C4 0 0 0 0 C4 0 C202 C4'0 C4 0 0 PNL Yes 708 PNL Yes 708 PNL Yes 708 PNL Yes 708 PNL Yes 708 PNL Yes 708 PNL Yes PNL Yes 708 708 A ON ON PNL Yes 708 C4 WBN2 IPEEE DESIGN REPORT Page 170 ATTACHMENT 1

TENNESSEE VALLEY AUTHORITY Watts Bar Unit 2 Nuclear Plant IPEEE Seismic Margins Evaluation Report -Safe Shutdown Paths and Safe Shutdown Equipment List Page Number A60 WBN-2-PNL

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TENNESSEE VALLEY AUTHORITY Watts Bar Unit 2 Nuclear Plant IPEEE Seismic Margins Evaluation Report -Safe Shutdown Paths and Safe Shutdown Equipment List WBN-1-BD -212-B002 -B WBN-1-OXF

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TENNESSEE VALLEY AUTHORITY Watts Bar Unit 2 Nuclear Plant IPEEE Seismic Margins Evaluation Report -Safe Shutdown Paths and Safe Shutdown Equipment List WBN-1-MCC

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-215-B -B DG 1B-B 125V BATTERY WBN-1-CHGR-215-A -A DG 1A-A BATTERY CHARGER WBN-1-CHGR-215-B -B DG lB-B BATTERY CHARGER WBN-1-MCC-215-AO01 -A 480V DIESEL AUXILIARY BOARD IA1-A WBN-1-MCC-215-AO02 -A 480V DIESEL AUXILIARY BOARD 1A2-A WBN-1-MCC-21S-BO01 -B 480V DIESEL AUXILIARY BOARD 1B1-B 1-15E500-2 1-15E500-2 1-15E500-2 1-15E500-2 1-15E500-2, 1-15E500-2 1-15E500-2 1-15E500-2 1-15E500-2 1-15E500-2 1-15E500-2 1-15E500-2 1-15E500-2 45W727 45W727 45W727 45W727 1-15E500-2 1-15E500-2 1-15E500-2 5 5 5 5 5 5 5 5 A A A A A A A A ON ON ON ON ON ON ON ON Page Number A62 ON MCC Yes 772 ON MCC Yes 772 A2 A16 ON MCC Yes 772 ON MCC Yes 772 ON MCC Yes 772 ON MCC Yes 757 ON MCC Yes 757 ON MCC Yes 757 5 A ON ON MCC Yes 757 5 A ON ON MCC Yes 757 5 5 5 5 5 5 5 5 5 5 A A A A A A A A A A ON ON ON ON ON ON ON ON ON ON ON MCC Yes 757 ON MCC Yes 757 ON MCC Yes 757 ON BAT No 742 ON BAT No 742 ON CHG Yes 742 ON CHG Yes 742 ON MCC Yes 761 ON MCC Yes 761 ON MCC Yes 761 A16.A15 A15 A2 A2 AS A5 A21 A21 A24 A24 DGR1 DGR1 DGRI DGR1 DGB DGB DGB WBN2 IPEEE DESIGN REPORT Page 173 ATTACHMENT 1

TENNESSEE VALLEY AUTHORITY Watts Bar Unit 2 Nuclear Plant IPEEE Seismic Margins Evaluation Report -Safe Shutdown Paths and Safe Shutdown Equipment List WBN-1-MCC

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-235-0003 -F 480V DIESEL AUXILIARY BOARD 182-B DG 2A-A 125V BATTERY DG 28-B 125V BATTERY DG 2A-A BATTERY CHARGER DG 2B-B BATTERY CHARGER 48OV DIESEL AUXILIARY BOARD 2A1-A 480V DIESEL AUXILIARY BOARD 2A2-A 480V DIESEL AUXILIARY BOARD 281-B 480V DIESEL AUXILIARY BOARD 2B2-B REACTOR VENT BOARD 1A-A REACTOR VENT BOARD 1B-B REACTOR VENT BOARD 2A-A REACTOR VENT BOARD 28-B 120 AC VITAL INST POWER BOARD 1-I 120V AC VITAL INST POWER BOARD 1-Il 120V AC VITAL INST POWER BOARD 1-111 120V AC VITAL INST POWER BOARD 1-IV 120V AC VITAL INVERTER 1-I 120V AC VITAL INVERTER 1-Il 120V AC VITAL INVERTER 1-Ill 1-1SE5OO-2 45W727 45W727 45W727 45W727 1-15E500-2 1-15E500-2 1-15ES00-2 1-15E500-2 1-15E500-2 1-15ES00-2 I-ISE500-2 1-15E500-2 1-45N706-1 1-45N706-2 1-45N706-3 1-45N706-4 1-45N703-1 1-45N703-2 1-45N703-3 Page Number A63 5 A ON ON MCC Yes 761 5 A ON ON BAT No 742 5 ' A ON ON BAT No 742 DGB DGR2 DGR2 A ON A ON A ON A ON A ON A ON A ON A ON A ON A ON A ON A ON A ON ON ON ON ON ON ON ON ON ON ON.ON ON ON CHG Yes 742 CHG Yes 742 MCC Yes 762 MCC Yes 762 MCC Yes 762 MCC Yes 762 MCC Yes 772 MCC Yes 772 MCC Yes .772 MCC Yes 772 BRD Yes 757 BRD Yes 757 BRD Yes 757 DGR2 DGR2 DGB DGB DGB DGB Al A2 A16 A15 A4 A3 A23 A23 A2 A2 5 A ON- ON BRD Yes 757 5 A ON ON INV Yes 772 5 A ON ON INV Yes 772 5 A ON ON INV Yes 772 A15 WBN2 IPEEE DESIGN REPORT Page 174 ATTACHMENT 1

TENNESSEE VALLEY AUTHORITY Watts Bar Unit 2 Nuclear Plant IPEEE Seismic Margins Evaluation Report -Safe Shutdown Paths and Safe Shutdown Equipment List WBN-1-INV

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TENNESSEE VALLEY AUTHORITY Watts Bar Unit 2 Nuclear Plant IPEEE Seismic Margins Evaluation Report -Safe Shutdown Paths and Safe Shutdown Equipment List Page Number A65 WBN-0-CHGR-236-0002 -E 125V VITAL BTRY CHARGER II 45N703-2 5 A ON ON CHG Yes 772 A2 WBN-0-CHGR-236-0003 -F WBN-0-CHG R-236-0004 -G WBN-0-CHGR-236-0006 -S WBN-0-CHGR-236-0007 -S WBN-0-DPL

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-271-Al 125V VITAL BTRY CHARGER III 125V VITAL BTRY CHARGER IV 125 VITAL BTRY CHARGER 6-S 125V VITAL BTRY CHARGER 7-S 125V VITAL BTRY BD I DISTRIBUTION PANEL 125V VITAL BTRY BD 11 DISTRIBUTION PANEL 125V VITAL BTRY BD III DISTRIBUTION PANEL 125V VITAL BTRY BD IV DISTRIBUTION PANEL 125V VITAL BTRY BD V DISTRIBUTION PANEL A 125V VITAL BTRY BD V DISTRIBUTION PANEL B 120V INSTRUMENT POWER DISTRIBUTION PANEL 1A 120V INSTRUMENT POWER DISTRIBUTION PANEL 1B INSTRUMENT POWER TRANSFORMER 1A INSTRUMENT POWER TRANSFORMER 1B 120V INSTRUMENT POWER DISTRIBUTION PANEL 2A 120V INSTRUMENT POWER DISTRIBUTION PANEL 2B INSTRUMENT POWER TRANSFORMER 2A INSTRUMENT POWER TRANSFORMER 2B AUX BLDG 225 TON CRANE 45N703-3 4SN703-4 45N703-1 45N703-3 45N703-1 45N703-2 45N703-3 4SN703-4 45W703-9 45W703-9 45W708-1 45W708-1 45W700-1 4SW700-1 45W708-1 45W708-1 45W700-1 45W700-1 44N230-237 5 5 5 5 5 5 5 5 5 5 5 5 5 5 5 5 5 5 ON ON ON ON..ON ON ON ON ON ON ON ON ON ON ON ON ON ON CHG Yes CHG Yes CHG Yes CHG Yes PNL Yes PNL Yes PNL Yes PNL Yes 772 772 772 772 A15 A15 A2 A25 757 A4 757 A3 757 A14 757 A13 PNL Yes 757 PNL Yes 757 BRD Yes 757 BRD Yes 757 XFMR Yes XFMR Yes BRD Yes BRD Yes XFMR Yes XFMR Yes 757 757 757 757 757 757 A2 A2 A4 A3 A4 A3 A23 A23 A23 A22 A OFF OFF CRN No 789 A13 WBN2'IPEEE DESIGN REPORT Page 176 AI-IACH MENT 1 TENNESSEE VALLEY AUTHORITY Watts Bar Unit 2 Nuclear Plant IPEEE Seismic Margins Evaluation Report -Safe Shutdown Paths and Safe Shutdown Equipment List Page Number A66 WBN-2-CRN

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-278-M026B -B UNIT 2 POLAR CRANE (275 TON)TURBO-GENERATOR AUXILIARY-RELAY PANEL TRAIN A BOP INSTR RACK TRAIN B BOP INSTR RACK TRAIN B BOP INSTR RACK TRAIN A BOP INSTR RACK REACTOR VESSEL LEVEL IND. SYS FEEDWATER AUXILIARY RELAY PANEL SEPARATIONS AUX RELAY PANEL A SEPARATIONS AUX RELAY PANEL A SEPARATIONS AUX RELAY PNL A COMM SEPARATIONS AUX RELAY PANEL B SEP AUX RELAY PANEL B AUX CNTL RM PNL AUX CNTL RM PNL AUX CNTL RM PNL ALL MAIN CONTROL ROOM PANELS RADIATION MONT & RECORD DSL GEN 1A-A MAIN CONT RM DSLGEN 1B-B MAIN CONT RM 44N230-237 NA NA NA NA NA NA NA NA 5 5 S 5 5 5 5 5 S 5 5 5 5 S S NA NA NA NA NA NA NA NA NA NA NA A A A A A A A A A A A A A A A A A A A A OFF ON ON ON ON ON ON ON ON ON ON ON ON ON ON ON ON ON ON ON OFF ON ON ON ON ON ON ON ON ON ON ON ON ON ON ON ON ON ON ON CRN No XFMR Yes PNL Yes PNL Yes PNL Yes PNL Yes PNL Yes PNL Yes PNL Yes PNL Yes PNL Yes PNL Yes-PNL Yes PNL Yes PNL Yes PNL Yes PNL Yes PNL Yes PNL Yes PNL Yes 708 708 708 708 708 708 708 708 708 708 708 708 757 757 757 755 755 755 755 0 0 0 0 0 0 0 0 0 0 0 0 RXB 5 5 5 5 5 5 C12 C12 C12 C12 WBN2 IPEEE DESIGN REPORT Page 177 ATTFACHMENT 1

TENNESSEE VALLEY AUTHORITY Watts Bar Unit 2 Nuclear Plant IPEEE Seismic Margins Evaluation Report -Safe Shutdown Paths and Safe Shutdown Equipment List WBN-0-PNL

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& AUX POWER VENT, ICE CONTAINMENT, & RCTR BLDG 120VAC INST PWR RACK 120VAC INST PWR RACK PFD PWR, INST PWR A&B TRANSFER SWITCHES TVA JUNCTION BOX TVA JUNCTION BOX NA NA NA NA N/A N/A 1-45W1646-3 1-45W1646-4 1-45W1646-1 N/A N/A A A A A A A A A A F,G F,G Page Number A67 ON ON PNL Yes 755 ON ON PNL Yes 755 ON ON PNL Yes 755 ON ON PNL Yes 755 ON ON PNL Yes 755 ON ON PNL Yes 755 ON ON PNL Yes 0 ON ON PNL Yes AUX/692 ON ON PNL Yes 755 ON ON/OFF N/A Yes RXB/703 ON ON/OFF N/A Yes RXB/703 C12 C12 C12 C12 C12 C12 0 0 MCR 0 0 WBN2 IPEEE DESIGN REPORT Page 178 AI-TACHMENT 1

Revision 0 April 27, 2010 Attachment 2: Watts Bar Nuclear Plant, IPEEE Seismic Margins Evaluation, Unit 2 Relay Design Report Report No. WBNIPEEE-004 WBN2 IPEEE DESIGN REPORT Page 179 TENNESSEE VALLEY AUTHORITY Title: Wafts Bar Nuclear Plant, IPEEE Seismic Margins Evaluation, Unit 2 Relay Design Report Prepared By: Michael G. Eason Date: 04/26/2010 Report No.WBNI PEEE-004 Page 1 of 25 I This report was prepared by the WBN Seismic Margin Relay Evaluation Team consisting of the following Members: 1)2)3)Michael Eason, Electrical Engineering Louvain Edmondson, Instrumentation and Controls Engineering Jimmie Perkins, Instrumentation and Controls Engineering And Approved by: 1) Mohan Bali -Discipline Lead EGS of Electrical Engineering

2) Fred Dimitrew -Discipline Lead EGS of Instrumentation and Control Engineering This is a Design Report.Checking and Verification are to be done as described herein.At that time, Checking and Verification signatures will be provided.Prepared By: Prepared Prepared By: Date: Date: ,'g .M1 Date: ZrN 10 Date: , 4/(z 6 10 Date: ____ ___ ___Approved.py:

(- A,0)Approve"PBy.'--

WBN2 IPEEE DESIGN REPORT Page 180 ATTACHMENT 2

TENNESSEE VALLEY AUTHORITY Title: Watts Bar Nuclear Plant, Report No.IPEEE Seismic Margins Evaluation, Unit 2 Relay Design Report WBNIPEEE-004 Prepared By: Michael G. Eason Date: 04/26/2010 Page 2 of 25 TABLE OF CONTENTS Section Title Page

1.0 INTRODUCTION

AND PURPOSE 5 1.1 Introduction 5 1.2 Purpose 5

2.0 REFERENCES

6 3.0 DESIGN INPUT DATA 7-8 3.1 Safe Shutdown Equipment List (SSEL) 7 3.2 List of Low-Seismic-Ruggedness Relays 7 3.3 Control and Power Interface Review 7 3.4 Bad Actor Relay List 7-8 4.0 ASSUMPTIONS

& METHODOLOGY 8-10 4.1 Assumptions 8-10 4.2 Methodology 10 5.0 ANALYSIS 11-16 5.1 General 11 5.2 Mercury Switches 11 5.3 Sudden Pressure Switches 1.1-12 5.4 6.9kV Shutdown Boards 12 5.5 480v Shutdown Boards 12 5.6 120v AC Vital Instrument Power Boards 12 5.7 Diesel Generator Protection Relay Panel 12-13 5.8 LOCA Containment Hydrogen Monitoring Valves 13 5.9 System Differences 13-16 6.0 RESULTS 16 7.0 VERIFICATION 16 WBN2 IPEEE DESIGN REPORT Page 181 ATTACHMENT 2

TENNESSEE VALLEY AUTHORITY Title: Wafts Bar Nuclear Plant, Report No.IPEEE Seismic Margins Evaluation, Unit 2 Relay Design Report WBNIPEEE-004*

Prepared By.- Michael G. Eason Date: 04/26/2010 Page 3 of 25 Figure No. Title Page 5.1 Sudden Pressure Switches 17 5.2 6.9kV Shutdown Boards 18 5.3 Diesel Generator Protection Relay Panel 19 5.4 LOCA Containment H2 Monitoring Isolation Valves 20 WBN2 IPEEE DESIGN REPORT Page 182 ATTACHMENT 2

TENNESSEE VALLEY AUTHORITY LIST OF APPENDIX Appendix No.Title A B C List of Low-Ruggedness-Relays

-EPRI-NP7148-SL Control And Power Interface Document List of Potential Bad Actor Relays Page 21 22 23-25 WBN2 IPEEE DESIGN REPORT Page 183 ATTACHMENT 2

TENNESSEE VALLEY AUTHORITY Title: Watts Bar Nuclear Plant, Report No.IPEEE Seismic Margins Evaluation, Unit 2 Relay Design Report WBNIPEEE-004 Prepared By: Michael G. Eason Date: 04/26/2010 Page 5 of 25

1.0 INTRODUCTION

AND PURPOSE 1.1 Introduction This report describes the relay evaluation portion of the Watts Bar Nuclear Plant, Unit 2 (WBNU2) Individual Plant Examination for External Events (IPEEE). Per NUREG-1407, Table 3.1, WBN is a "0.3g Focused Scope" plant. As described in GENERIC LETTER 88-20, SUPPLEMENT 5, the required relay evaluation for a 0.3g Focused Scope plant is limited to a review for low seismic ruggedness relays (Bad Actor Relays). Bad Actor Relays are those that can be prone to "chatter" during a review level earthquake. "Chatter" is considered to be the inadvertent opening or closing of a relay, contact, or switch with a sustained output of two milliseconds or more.This WBNU2 Focused Scope approach has been refined from the approach used for the relay review performed for the WBNU1 IPEEE. That is, the WBNU1 relay evaluation included portions of the required relay review for a Full Scope plant, including a review for fail safe circuitry for all devices on the SSEL followed by the review for Bad Actor Relays. The WBNU1 relay review, completed in 1997, found that no low seismic ruggedness relays were used in applications which would qualify them as essential relays, and that no corrective actions were required for this element of the IPEEE.Although recognized as a refinement to the WBNU1 approach, the WBNU2 Focused Scope approach is considered to be fully adequate for the WBNU2 IPEEE. WBNU1 results were confirmatory, and the WBNU2 Focused Scope approach is fully consistent with the regulatory guidance.

Consistent with NUREG-1407, the WBNU2 design review for low ruggedness relays will be performed as the first step, followed by the fail safe review for identified low-ruggedness relays.1.2 Purpose The purpose of this relay chatter evaluation is to determine if any Bad Actor Relays exist in WBNU2 safe shutdown systems for which malfunction is unacceptable.

This evaluation will also determine if any Bad Actor Relays exist in upgraded or added systems since the WBNU1 IPEEE evaluation.

Proper design output requirements are in place to prevent the installation of "Bad Actor Relays" in seismically sensitive configurations which could degrade the design basis response of critical plant safety features.

Proper due diligence is required to confirm these requirements are in place and their intent is met.This design report is a snapshot in time. It will be verified once designed systems, devices and components are installed and tested.WBN2 IPEEE DESIGN REPORT Page 184 ATTACHMENT 2

TENNESSEE VALLEY AUTHORITY Title: Watts Bar Nuclear Plant, Report No.IPEEE Seismic Margins Evaluation, Unit 2 Relay Design Report WBNIPEEE-004 Prepared By: Michael G. Eason Date: 04/26/2010 Page 6 of 25

2.0 REFERENCES

2.1 WBNIPEEE-001, "Watts Bar Nuclear Plant IPEEE Seismic Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List" -Unit 1 Evaluation 2.2 WBNIPEEE-002, "Watts Bar Nuclear Plant IPEEE Seismic Margins Evaluation Relay Evaluation" -Unit 1 Relay Evaluation 2.3 WBNIPEEE-003, "Watts Bar Nuclear Plant IPEEE Seismic Margins Evaluation Safe Shutdown Paths and Safe Shutdown Equipment List" -Unit 2 Evaluation 2.4 EPRI NP-7147-SL, 'Seismic Ruggedness of Relays", August 1991.2.5 NUREG-1407, "Procedural and Submittal Guidance for the Individual Plant Examinations of External Events(IPEEE) for Server Accident Vulnerabilities", Final Report, US Nuclear Regulatory Commission, June 1991.2.6 EPRI NP-7148-SL, "Procedure for Evaluating Nuclear Power Plant Relay Seismic Functionality", December 1990.2.7 EPRI NP-6041 M, Revision 1, "A Methodology for Assessment of Nuclear Power Plant Seismic Margin", August 1991.2.8 EPRI NP-5223-SL, Revision 1, "Generic Seismic Ruggedness of Power Plant Equipment" August 1991.2.9 SS-E18.7.42, TVA Nuclear Power Standard Specifications for Process Instrument Switches.

Multiple listings of Approved Specifications for Procurement of Devices.2.10 Generic letter 88-20 Supplement 4 & 5, Individual Plant Examination of External Events" (IPEEE).2.11 WB-DC-30-27, "AC and DC Control Power Systems" (UNIT 1 / UNIT 2), Rev. 0029, October 2009.2.12 NEDP-8, "Technical Evaluation For Procurement of Materials And Services", Rev.0014, July 2009.2.13 WBN-VTD-W120-2568, "Westinghouse Full Length Rod Control System", Watts Bar Nuclear Plant.WBN2 IPEEE DESIGN REPORT Page 185 ATTACHMENT 2

TENNESSEE VALLEY AUTHORITY Title: Watts Bar Nuclear Plant, Report No.IPEEE Seismic Margins Evaluation, Unit 2 Relay Design Report WBNIPEEE-004 Prepared By: Michael G. Eason Date: 04/26/2010 Page 7 of 25 3.0 DESIGN INPUT DATA 3.1 Safe Shutdown Equipment List (SSEL)The SSEL, provided in Reference 2.3, identifies the systems and equipment needed to achieve WBNU2 safe shutdown.3.2 List of Low-Seismic-Ruggedness Relays (Appendix A)A list of Bad Actor Relays, as identified in Appendix E of Reference 2.7, is included in this report as Appendix A. The relay evaluation procedure for seismic demand determination and Generic Equipment Ruggedness Spectra (GERS) cannot be applied to these relays due to their low seismic ruggedness.

Such relays are considered acceptable if it can be demonstrated that the effects of the chatter can be reset by operator action or that the relay contact does not play a role in achieving safe shutdown.3.3 Control and Power Interface Review (Appendix B)The equipment on the WBNU2 SSEL requiring power or control was: reviewed to determine whether the control and power requirements were unlike that of WBNU1. The objective was to identify the relays, contacts and power sources used by the safe shutdown equipment that may be different than that of Unit 1. Flow, control, wiring and logic diagrams as well as single-line and schematic drawings were the source documents for this review. The fail safe logic established by the UI IPEEE would apply to WBNU2 unless any differences in devices, components or wiring were identified.

All differences will be reflected within Appendix B of the verification report.3.4 Bad Actor Relay List (Appendix C)The Relay List identifies potential Bad Actor Relays requiring further screening or analysis.This list was developed based on the SSEL, the Control and Power Interface Review, the Electric Power Research Institute (EPRI) NP-7147-SL Appendix E list of low ruggedness relays and the following TVA and vendor design input data:* Vendor Assembly and Bill of Material Drawings" Vendor Manuals" Manufacturer supplied Documentation" TVA Connection and Bill of Material Drawings* TVA Contract Files* TVA Design Change Notice (DCN) Document/Engineering Document Construction Release (EDCR) Packages WBN2 IPEEE DESIGN REPORT Page. 186 ATTACHMENT 2

TENNESSEE VALLEY AUTHORITY Title: Watts Bar Nuclear Plant, Report No.IPEEE Seismic Margins Evaluation, Unit 2 Rela Design Report WBNIPEEE-004 Prepared By: Michael G. Eason Date: 04/26/2010 -Page 8 of 25'TVA Design Criteria and Operation/Construction Database 4.0 ASSUMPTIONS AND METHODOLOGY 4.1.1 Per reference 2.5 "NUREG-1407", paragraph 3.2.4.2, "Relay Evaluation" and table 3.1,'Review Level Earthquake

-Plant Sites East of the Rocky Mountains", the scope of the Watts Bar Nuclear Plant relay chatter evaluation is to locate and evaluate low-seismic-ruggedness relays.4.1.2 Per reference 2.4, "EPRI NP-7147-SL" pages 3-3 and 3-4, the following are used in establishing the set of relays evaluated for relay chatter: " With the exception of the type General Electric IJD (NON-1 E) and the English Electric type YCG relays, the Bad Actor Relays are not structurally damaged as a result of an earthquake and will be functional after the period of strong shaking.* It is necessary to verify that relay malfunction during strong shaking does not result in an unacceptable seal-in, lockout, or system disabling action. In such cases, operator actions to reset or restore such circuits to their original condition are acceptable provided there is sufficient time, access, indication and procedures for such actions to be taken.* Specific relays and their associated circuits are considered non-essential for shutdown after an earthquake if: 1. The function provided by the system and associated relays is not needed during the period of strong shaking and relay malfunction will not make essential functions unavailable when needed after strong shaking, or operator action can be taken to restore the function.2. Relay malfunction does not prevent the desired function (e.g., reactor trip) from occurring.

3. Relay malfunction does not cause a spurious, unacceptable event.4. Relay malfunction affects only alarm circuits.* No corrective actions are required for non-essential relays.* Relay chatter is considered to be the inadvertent opening or closing of a relay contact with a sustained output of two milliseconds or more." Mechanically (versus electromechanical) actuated contacts, such as control switch contacts (e.g., hand switches, transfer switches, etc.) and limit switch contacts(e.g., those on motor operated valves) are judged to be not seismically vulnerable.

These manual and mechanically driven switches require the application of reasonable force in order to change state.WBN2 IPEEE DESIGN REPORT Page 187 ATTACHMENT 2

TENNESSEE VALLEY AUTHORITY Title: Watts Bar Nuclear Plant, Report No.IPEEE Seismic Margins Evaluation, Unit 2 Relay Design Report WBNIPEEE-004 Prepared By: Michael G. Eason Date: 04/26/2010 Page 9 of 25* Solid state relays (with no mechanically moving parts) are considered inherently rugged and no seismic capacity evaluation is required.4.1.3 Per Appendix A, transformer pressure surge sensing devices are considered low ruggedness relays (sudden pressure switches).

4.1.4 A relay may latch or seal-in due to the seismic event. The relay evaluation procedure, EPRI NP-7148-SL, allows for restoring or resetting systems providing there is sufficient time for the operator or an assistant to perform the necessary diagnostic evaluations and take the necessary resetting or restoring actions. This may be dependent on the system involved.However, one-half to one hour should be adequate time to reset most systems.4.1.5 EPRI NP-5223-SL (Reference 2.8) identifies the following position for relay chatter in switchgear enclosures:

The functionality of switchgear is governed by the control, protective, and associated auxiliary relays. In many tests the standard 2-ms chatter failure criterion was utilized.Due to this conservative failure criterion, many relays were found to be unacceptable when mounted on switchgear enclosures.

However, in other tests the switchgear was considered as a complete subsystem where chatter can be tolerated without loss of switchgear function.

The primary function of the switchgear is to either connect or disconnect a main power circuit to protect the remainder of the circuit against overcurrent, overvoltage, undervoltage, phase reversal, etc., or to switch circuits upon demand. As long as the switchgear can perform its primary function, the occurrence of minor chatter in a control, protective, or associated auxiliary relay is not a relevant concem. Thus, in contrast to control circuits where short duration (> 2ms) relay chatter is considered a failure to "operate", switchgear fails only by a non-commanded change of state. In order-to cause switchgear loss of function, relay chatter must have significant duration (usually greater than the period of one AC cycle or > 16ms). IEEE guidelines for'seismic testing of switchgear have recognized this chatter tolerance in switchgear.

4.1.6 The diesel generator is not required until after the seismic event which is consistent with the scenario described in Reference 2.7. Therefore, the controls (i.e., relays) necessary to start the diesel generator are not required during the strong motion of the seismic event.4.1.7 The 480V Diesel Auxiliary Boards are not used until after the diesel generator is supplying power to the plant. Section 4.1.6 states that the diesel generator is not required until WBN2 IPEEE DESIGN REPORT Page 188 ATTACHMENT 2

TENNESSEE VALLEY AUTHORITY Title: Watts Bar Nuclear Plant, Report No.IPEEE Seismic Margins Evaluation, Unit 2 Relay Design Report WBNIPEEE-004 Prepared By: Michael G. Eason Date: 04/26/2010 Page 10 of 25 after the seismic event. Only those relays that latch seal-in or otherwise become unavailable will be evaluated.

4.1.8 The SSEL, Relay Evaluation and IPEEE Report will be verified at a point in the future.This point will be after all operating procedures are available, start-up testing is finalized and construction is substantially complete.4.2 Methodology The following describes the approach which will be used in verifying the low ruggedness relays which could interfere with the safe shutdown of WBN:* Determine similarities between WBNU1 SSEL and WBNU2 SSEL and screen like-for-like equipment." Select the equipment requiring power from the Safe Shutdown Equipment List., Perform a "Control-Power Interface Review" (Appendix B) to identify the circuits and relays/contacts which affect the operation of the safe shutdown systems.* Identify the equipment in which potential "Bad Actor" relays reside. Using the design input data in Section 3.4, develop a list of potential "Bad Actor" relays types. It should be noted that the "Operating Mode" of the W Type SG and the GE Type HGA relays must be determined before classifying them as Bad Actor Relays.* Evaluate the application of the relays on the list and develop a list of relays requiring further study; relays are defined by Appendix E of Reference 2.6.* Screen the Bad Actor Relays whose chatter/malfunction will not prevent system/component functioning or cause an unacceptable condition." Screen the Bad Actor Relays where "operator action" to reset or restore system/function is acceptable." Designate relays which cannot be screened as "Essential Relays"." Identify required corrective action.WBN2 IPEEE DESIGN REPORT Page 189 ATTACHMENT 2

TENNESSEE VALLEY AUTHORITY Title: Watts Bar Nuclear Plant, Report No.IPEEE Seismic Margins Evaluation, Unit 2 Relay Design Report WBNIPEEE-004 Prepared By: Michael G. Eason Date: 04/26/2010 Page 11 of 25 5.0 ANALYSIS 5.1 General The scope of the relay evaluation as stated in NUREG-1407 is to "locate and evaluate low ruggedness relays" involved in the operation of Seismic Margin equipment.

The following activities are required to accomplish this: 1. Locate the set of relays, mercury switches, and transformer pressure surge sensing devices (sudden pressure switches) which affect the operation of the safe shutdown systems. (see Appendix A)2. Develop a list of low ruggedness relays, mercury switches, and transformer pressure surge sensing devices (sudden pressure switches) that are on the Bad Actor Relay list (see Appendix B).This is accomplished using the mechanics described in Section 4.2.3. Screen out non-essential Bad Actor Relays and switches.4. Identify essential Bad Actor Relays.5. Prove that the intentionally designed, fail safe circuitry allows for safe shutdown.This item is not required but performed.

6. Identify the corrective action required where essential Bad Actor Relays are used.This section describes the rationale used in arriving at the Potential Bad Actor Relay list. This section is derived from both the similarities between Unit 1 and Unit 2 and the differences between Unit I and Unit 2. By developing a list of all potential Bad Actors at WBN, the total effective fragility can be realized.5.2 Mercury Switches The use of mercury wetted contacts in pressure, temperature, level, and flow switches is prohibited and controlled by NEDP-8, Reference 2.12. Also the design requirements section of the Standard Specifications for each type of switch prohibits the use of mercury through the utilization of Form 10581-13 ."I&C Requisition Data Sheet", specifically Line 15.5.3 Sudden Pressure Switches The only sudden pressure switches which could affect the safe shutdown systems are associated with the two preferred power 161 to 6.9kV common station service transformers (CSST C&D). CSST C&D and the associated switches are not safety related and as such are not included on the SSEL. The switch's function is to initiate an automatic open-circuit fast transfer between the normal and alternate feeds to the 6.9kV shutdown boards. Pressure switch contact chatter associated with CSST "C" could cause spurious "fault pressure relaying" resulting in fast transfer from normal to alternate feed. This is the result of contacts WBN2 IPEEE DESIGN REPORT Page 190 ATTACHMENT 2

TENNESSEE VALLEY AUTHORITY Title: Watts Bar Nuclear Plant, Report No.IPEEE Seismic Margins Evaluation, Unit 2 Relay Design Report WBNIPEEE-004 Prepared By: Michael G. Eason Date: 04/26/2010 Page 12 of 25 C86C1X and C86C2X which signal closure for 6.9kV shutdown boards 1A-A and 2A-A (see Figure 5.1). Once auto fast transfer is initiated, further automatic fast transfer is prevented by the auto transfer lockout logic. Spurious fault pressure relaying associated with CSST "D" will cause a fast transfer for feeds 1 B-B and 2B-B. This inadvertent transfer is acceptable, and will not prevent the starting of the diesel generator initiated by the postulated loss of off-site power (LOOP). Therefore sudden pressure switch contact chatter is acceptable.

5.4 6.9kV Shutdown Boards The GE Model 12HGA14AF52 relays are interposing relays (AX) which provide contacts'to the closing circuit of the four diesel generator breakers.

The normal operating mode for the relays is de-energized and the normally open contact is used in the control circuit. Therefore, the relays do not qualify as Bad Actor Relays (see Figure 5.2).The GE Model 12HGA11J52 relays are auxiliary relays (51x, DCSI"x" & DCSI"x"E) which provide contacts to annunciation circuits and are screened by assumption 4.1.2.5.5 480V Shutdown Boards Three Westinghouse Type SG relays are in each Unit board and are under voltage relays (27S"xxx"X, DCS"x" & DCS"x"E) which provide contacts to annunciation circuits and are screened by assumption 4.1.2.Two Westinghouse Type SG relays are in each Unit board. The equipment controlled by these relays is not on the SSEL and therefore are considered screened.5.6 120V AC Vital Instrument Power Boards One Westinghouse Type SG relay is in each Unit board and serves as an under voltage relay (27/2-"x")

which provides a contact to an annunciation circuit and is screened by assumption 4.1.2.5.7 Diesel Generator Protection Relay Panel There are twelve GE Model 12HGA17M52 relays in each relay panel. The relays are designated as follows: 74, ES1AY1, RRX1A, ESX1A, ESXl 1A, ESX21A, SRX1A, SLX1A, R01, RC1, R02 and SPARE. Relay RRXIA's normal operating mode is de-energized and WBN2 IPEEE DESIGN REPORT Page 191 ATTACHMENT 2

TENNESSEE VALLEY AUTHORITY Title: Watts Bar Nuclear Plant, Report No.IPEEE Seismic Margins Evaluation, Unit 2 Relay Design Report WBNIPEEE-004 Prepared By: Michael G. Eason Date: 04/26/2010 Page 13 of 25 provides a normally closed contact to the "Safety Shutdown Reset" circuit and therefore qualifies as a Bad Actor Relay. The other relays are screened based on thleir operating mode.The function of relay RRXlA is to provide remote reset from main control room panel 0-M-26 (see Figure 5.3). Since there would be no reason for the circuit to be set, contact chatter would have no adverse effect and therefore the relay is screened.5.8 LOCA Containment Hydrogen Monitor Sampling Isolation Valves There is one GE Model 12HGAI 1J52 relay in the circuit of the four isolation valve pairs. The relays are designated as 42X1 & 42X2. During normal operation each relay is de-energized and two normally open contacts are used in the Auto-Open circuit of each valve pair (see Figure 5.4). GE type HGA relays used in this operating configuration do not qualify as Bad Actor Relays.5.9 WBNU1 "and WBNU2 System Differences In order to accept like for like equipment, it is also necessary to accept like for unlike equipment.

This section identifies these differences at a design level and will be an integral portion of the verification activities at WBN which will be included in the final IPEEE report.The below systems have been .identified by the design Relay Evaluation team to have dissimilar or new qualities than that of the WBNU1 relay evaluation.

5.9.1 System

043 -Sampling and Water Quality The only safety related function that System 43 provides is containment isolation of the containment penetrations to which the sample lines are attached.

Sampling System has fail-closed inboard and outboard containment isolation valves for each penetration to which a sample line is attached.

Sampling and Water Quality System has no outgoing signals other than valve position indication and illumination.

There is no component or electrical difference between Unit 1 and Unit 2 at WBN but the system, identified by Reference 2.2 (WBNU1 IPEE Relay Evaluation) as having a "Potential Bad Actor Relay," was not fully described in the Unit I Evaluation.

This system is now described for clarity and confirmation only.WBN2 IPEEE DESIGN REPORT Page 192 ATTACHMENT 2

TENNESSEE VALLEY AUTHORITY Title: Watts Bar Nuclear Plant, Report No.IPEEE Seismic Margins Evaluation, Unit 2 Relay Design Report WBNIPEEE-004 Prepared By: Michael G. Eason Date: 04/26/2010 Page 14 of 25 5.9.2 System 085 -Control Rod Drive Mechanisms This system consists of two types of rods: 1) shutdown rods and 2) control rods.Shutdown rods are required to provide sufficient negative reactivity thus ensuring the reactor remains subcritical.

These rods are fully withdrawn during normal operation.

Control rods are used to control the reactor core reactivity.

Shutdown and control rods are raised or lowered by a prescribed set of electromechanical actions performed by the control rod drive (CRD) mechanisms.

A review of the Vendor Technical Manual (WBN-VTD-W120-2568), Reference 2.13, for the Full Length Rod Control System was completed in order to verify whether any installed relays could negatively affect components identified by the SSEL. The review determined that Bad Actor Relays, as identified in Appendix E of Reference 2.7 "Low Ruggedness Relays", are not used in the Unit 1 Control Rod Drive System and will not be used in the Unit 2 system.After this design review, it was determined that the shutdown rods identified in the SSEL supply enough negative reactivity to maintain the reactor in a safe shutdown condition.

5.9.3 System

090 -Radiation Monitoring System There are ten flow control valves and four radiation monitors on the SSEL for this system. The four radiation monitors are for indication only and have no control function, therefore having no impact to the safe shutdown of WBNU2.The Unit 2 controls for these valves will be like the Unit I valve controls, therefore no bad actor relays will be used.5.9.4 System 092 -Neutron Monitoring System The WBNU2 Power Range and Auxiliary Equipment drawers are refurbished by Westinghouse and the relays used will be similar to Unit 1 but improved due to certain upgrades.

The function of this part of System 92 is identical to Unit 1 except a time delay circuit is added to Unit 2 Flux Deviation drawer, whereas Unit 1 has a time delay in the annunciator system for the QPTR alarms. Like the Unit 1 system, no bad actor relays will be used in the Unit 2 Neutron Monitoring System.WBN2 IPEEE DESIGN REPORT Page 193 ATTACHMENT 2

TENNESSEE VALLEY AUTHORITY Title: Watts Bar Nuclear Plant, Report No.IPEEE Seismic Margins Evaluation, Unit 2 Relay Design Report WBNIPEEE-004 Prepared By: Michael G. Eason Date: 04/26/2010 Page 15 of 25 5.9.5 System 099 -Reactor Protection System The purpose of the Reactor Protection System (RPS) is to provide automatic protection against unsafe and improper reactor operation during steady-state and transient power operations and to provide initiating signals to mitigate the consequences of faulted conditions.

The Reactor Protection System is composed of two subsystems, the reactor trip subsystem and the engineered safety features actuation subsystem.

The reactor trip subsystem automatically keeps the reactor operating within a safe region by shutting down the reactor whenever the limits of the region are approached.

The safe operating regionis defined by several considerations such as mechanical/hydraulic limitations on equipment, and heat transfer phenomena.

Therefore, the reactor trip subsystem keeps surveillance on process variables which are directly related to equipment mechanical limitations, such as pressure, pressurizer water level and also on variables which directly affect the heat transfer capability of the reactor (e.g., flow and reactor coolant temperatures).

Still other parameters utilized in the reactor trip system are calculated from various process variables.

In any event,-whenever a direct or calculated variable exceeds a set point;- the reactor will be shut down in order to protect against either damage to fuel cladding or loss of system integrity which could lead to release of radioactive fission products into the containment.

The engineered safety features actuation subsystem uses selected plant parameters, determines whether or not predetermined safety limits are being exceeded and, if they are, combines the signals into logic matrices sensitive to combinations indicative of primary or secondary system boundary ruptures.

Once the required logic combination is completed, the system sends actuation signals to the appropriate Engineered Safety Features.The reactor trip system, and engineered safety features actuation system shall be capable of providing the necessary protective actions during and after a Safe Shutdown Earthquake (SSE). Therefore, the reactor protection system shall be capable of tripping the reactor during and after a Safe Shutdown Earthquake.

The engineered safety features actuation system and the safety features systems shall be designed to initiate their protective functions during and after an SSE.Integral to the RPS is the Eagle 21 Process Protection System located in panel 2-R-28 and panels 2-R-1 through 2-R-13 and the Solid State Protection System (SSPS)located in panel 2-R-58 and panels 2-R-46 through 2-R-55. Westinghouse has been contracted for WBN Unit 2 to install the Eagle 21 equipment and refurbish the SSPS hardware in the aforementioned panels.WBN2 IPEEE DESIGN REPORT Page 194 ATTACHMENT 2

TENNESSEE VALLEY AUTHORITY Title: Watts Bar Nuclear Plant, Report No.IPEEE Seismic Margins Evaluation, Unit 2 Relay Design Report WBNIPEEE-004 Prepared By: Michael G. Eason Date: 04/26/2010 Page 16 of 25 Currently, designed system, components do not house or rely on Bad Actor Relays.Verification activities will include confirming the SSEL components of this system are not subject to unacceptable relay chatter.5.9.6 System 236 -125 VDC Vital Power An addition to the WBNU2 SSEL included the Fifth Vital Battery and the components to distribute and transfer power from them to either Train A or Train B via a transfer switch located on the panel; making it completely independent and able to serve either Train as a backup for maintenance or accident purposes.

The batteries, distribution panefs A&B and the 125 V Vital Battery Board V were confirmed to have no low seismic ruggedness relays and fail safe circuitry is adequate for the system. The charger for Battery V is not included in the SSEL since during the use of the batteries, charging them is performed through the use of the SPARE charger for the system it is serving. Both SPARE chargers were evaluated and qualified by the WBNU1 IPEEE.There are no Inverters specifically for the Vital Battery V loop. Existing inverters, also evaluated and qualified by the WBNUI IPEEE, are utilized from the battery that is unavailable.

6.0 RESULTS

Through initial design review, there are no low-seismic-ruggedness relays used in applications which would qualify them as an essential relay; therefore no corrective action is anticipated.

7.0 VERIFICATION

Verification activities include a full review, in respect to changes, of the design documents created for this design report. The design documents created for this review will first be verified and validated against the Unit 2 SSEL when the Unit 2 Master Equipment List (MEL), OPS Procedures and PSA Review can be completed and verified (Reference 2.3, Section 1.0).During this verification review, a Control and Power Interface will be created to document any differences between Unit 1 and Unit 2, including but not limited to SSEL, control and/or power and fail safe circuitry.

This Interface will be the source document for further investigation of Bad Actor Relays and will be available as Appendix B of the verification report. If no differences in Unit 1 and Unit 2 are identified, this verification report will conclude that WBN requires no corrective actions for the issue of Relay Chatter.WBN2 IPEEE DESIGN REPORT Page 195 ATTACHMENT 2

TENNESSEE VALLEY AUTHORITY Title: Watts Bar Nuclear Plant, IPEEE Seismic Margins Evaluation, Unit 2 Relay Design Report Prepared By: Michael G. Eason Date: 04/26/2010 Report No.WBNIPEEE-004 Page 17 of 25 FIGURE 5.1 TVA Drawing # 45W760-21 1-1 Rev 12 & 45W760-211-3 Rev 8 ThW~!A wiaFm oar CYDIAGAN fimw SlfllU W U1k-fftZ1tV~L IETULX 3aal1 LM l M SUM M460-LI*gO0v SHUTDOWN 50 1A-A (2A-A) NORMAL FEEDER BREAKER 17$6 (1816)WBN2 IPEEE DESIGN REPORT Page 196 ATTACHMENT 2

TENNESSEE VALLEY AUTHORITY Title: Watts Bar Nuclear Plant, Report No.IPEEE Seismic Margins Evaluation, Unit 2 Relay Design Report WBNIPEEE-004 Prepared By: Michael G. Eason Date: 04/26/2010 Page 18 of 25 FIGURE 5.2 TVA Drawing # 45W760-211-4 Rev 15 SA6E SASEC 14 se T_1-i 5- NCR 7 rtim VW.15C 1-(--74 ! -XS-S?-4B NOR AUX --51 -45'.I, SAMNISAWT C AUX;-, ________________________________

SAGCP/ k'x41[ 4 , I _____ "_ _/ S,-.5s-46, r i. I .Js-5T- EAI /I Ct pULL- r , A12 Al 5 -:. l--,A AUT --H---4-it'6A B, / I -4 AL. 0 CL /41J C .4.R. .l .. 2H TES 3 SAGC3 AS5 PS2 5 , 24 1 MR .22C ,- I A a 1 6- -1932 1. AX.3 52ST V3 3ASG-171. E. 171. 2. 1 AI- 1-401 12 91 N o AX 5 ! IG -P5 0 11 912 912 3 -.SA 5211. Wl A 5c 2 4 1711;7C 71S 7- 5 7 a m on3 <rm SS4X r 1 *13 5~L 10 P3 A N J- 1 4C 1 14 E4 UOS INGCK 01 I- 3 10211 1 100 .6~AUX *~U oi AUX 9 2 52HL.3C 7C TTEST'RIP'S-57-49 N 6 ilWXS-57-46

"'1 6.9kV SHUTDOWN BOARDS WBN2 IPEEE DESIGN REPORT Page 197 ATTACHMENT 2

TENNESSEE VALLEY AUTHORITY Title: Watts Bar Nuclear Plant, Report No.IPEEE Seismic Margins Evaluation, Unit 2 Relay Design Report WBNIPEEE-004 Prepared By: Michael G. Eason Date: 04/26/2010 Page 19 of 25 FIGURE 5.3 TVA Drawing # 45W760-82-2C Rev 8 DIESEL GENERATOR PROTECTION RELAY PANEL (UNIT 2)WBN2 IPEEE DESIGN REPORT Page 198 ATTACHMENT 2

TENNESSEE VALLEY AUTHORITY FIGURE 5.4 TVA Drawing # 45W600-43-2 Rev 25 4-FCV-43-201 (INBOARD)

AND FCV-43-433 (OUTBOARD)

TRAIN A L6WA h-. t~dklAidW4T IOLAT.OM VALV17 W1 RE FUSE LOCAL RELAY RELAY (ONSTANT IVA H. WPREFI PL PANEL LOCATIOH CURRENT______LIN m wIT 1 UmI r FCV-4N-201 A2 ', -4 FCV-43-433 A SNK I -A2!5 L-448 42X1 JB-6295 YC-4$-2z1 Fr;V-43-43q A SVYL I -A27 L-446 42X2 JB-6295 W-43-201 FCV-43-2073 5Wkl ElT-B34 -L-40) 42N1 J3-8294 X-.43-2D7 FCY-43-435 CV-43-2 SWN ,I -835 -L-449 42X2 J,-6294 XC-43-2D&LOCA CONTAINMENT HYDROGEN MONITOR ISOLATION VALVES WBN2 IPEEE DESIGN REPORT Page 199 ATTACHMENT 2

TENNESSEE VALLEY AUTHORITY APPENDIX A -LOW RUGGEDNESS RELAYS The relay evaluation procedure seismic demand determination and GERS cannot be applied to these relays because of their low seismic ruggedness or demonstrated sensitivity to high frequency vibration.

Case specific techniques or current qualification techniques must be utilized to demonstrate the adequacy of these relays.RELAY TYPE OPERATING MODE REFERENCES GE CFD ALL 1 (91-14/313, 82-26/348, 86-13/293), 2, 3, 4, 5, (IN 85-82), 6 GE CFVB ALL 2, 3, 6 GE CEH ALL 2,6 GE CPD ALL 2,6 GE IJD* (NON IE) ALL 2 GE PVD1I AND PVD21 ALL 1 (84-20/352), 3, 4, (GE)GE RAV1I ALL 4(GE)GE HGA (DE, NC) 1 (84-18/331, 86-15/269, 87-11/250), 4,5 (IN 88-14)GE HFA65 ALL 4 (BNL)W HLF ALL 2, 6 W HU (NON 1E) ALL 3,6 W ITH ALL 1 (81-44/346 AND 81-37/346)

W ARMLA ALL 5 (IN 82-55)W PMQ ALL 1 (85-16/247)

W SG (DE, NC) 4 (ANCO)W SV ALL 4 (BNL)W SC ALL 4 (BNL)W SSC ALL 4 (BNL)W CON-5 (NON 1 E)** ALL 1 (88-06/387)

ASEA ARMX-L ALL 1 (88-06/387)

ENGLISH ELECTRIC YCG* ALL 2 MERCURY SWITCHES ALL 1 (86-25/249), 2 SUDDEN PRESSURE ALL 2 SWITCHES ir

REFERENCES:

1)GERS 2)EARTHQUAKE EXPERIENCE DATA 3)SAFEGUARDS DATA 4)IEEE 501 TEST DATA 5)NOTICES, BULLETINS, ETC 6)INDUCTION CUP OR INDUCTION CYLINDER DESIGN* DE = DE-ENERGIZED

  • E = ENERGIZED* NC = NORMALLY CLOSED CONTACT* NO = NORMALLY OPEN* ALL = ALL MODES* DAMAGE HAS OCCURRED TO THIS RELAY IN AN EARTHQUAKE AND IT MUST BE ASSUMED THAT IT WILL BE INOPERABLE FOLLOWING AN SSE LEVEL EARTHQUAKE Tr TRANSFORMER PRESSURE SURGE SENSING DEVICES WITH SSC-T OR IITH UNIT WBN2 IPEEE DESIGN REPORT Page.200 ATTACHMENT 2

TENNESSEE VALLEY AUTHORITY Title: Watts Bar Nuclear Plant, Report No.IPEEE Seismic Margins Evaluation, Unit 2 Relay Design Report WBNIPEEE-004 Prepared By: Michael G. Eason ! Date: 04/26/2010 Page 22 of 25 APPENDIX B -Control and Power Interface Document PROVIDED AFTER COMPLETED VERIFICATION ACTIVITIES WBN2 IPEEE DESIGN REPORT Page 201 ATTACHMENT 2

TENNESSEE VALLEY AUTHORITY Title: Watts Bar Nuclear Plant, Report No.IPEEE Seismic Margins Evaluation, Unit 2 Relay Design Report WBNIPEEE-004 Prepared By: Michael G. Eason Date: 04/26/2010 Page 23 of 25 APPENDIX C -LIST OF POTENTIAL BAD ACTOR RELAYS Bad Actor See Component ID Component Description Relay Type #of Relays Section REFERENCE DOCUMENTS GE TYPE TVA DWG 45W727, DG 1A-A PROTECTION RELAY 12HGA17M52 45W760-82-1, -2, -3, -4, -5, -6, 1-ARB-082-A-A PANEL /W TYPE SC 12 / 1 5.7 D47495-3 GE TYPE TVA DWG 45W727, DG 1B-B PROTECTION RELAY 12HGA17M52 45W760-82-1, -2, -3, -4, -5, -6, 1-ARB-082-B-B.

PANEL /W TYPE SC 12/1 5.7 D47495-3 GE TYPE TVA DWG 45W727, DG 2A-A PROTECTION RELAY 12HGA17M52 45W760-82-1, -2, -3, -4, -5, -6, 2-ARB-082-A-A PANEL / W TYPE SC 12 / 1 5.7 D47495-3 GE TYPE TVA DWG 45W727, DG 2B-B PROTECTION RELAY 12HGA17M52 45W760-82-1, -2, -3, -4, -5, -6, 2-ARB-082-B-B PANEL / W TYPE SC 12 / 1 5.7 D47495-3 TVA DWG 45W724-1, 6.9 KV SHUTDOWN BOARD GE TYPE 45W760-211-4, 45BM247-1, 1-BD-211-A-A 1A-A 12HGA14AF52.

4 5.4 GE 84376,CONNECTION DWGS WVA DWG 45W724-1, 6.9 KV SHUTDOWN BOARD GE TYPE 45W760-211-4, 45BM247-2, 1-BD-211-B-B lB-B 12HGA14AF52 4 5.4 GE 84376,CONNECTION DWGS TVA DWG 45W724-1, 6.9 KV SHUTDOWN BOARD GE TYPE 45W760-211-4, 45BM247-1, 2-BD-211-A-A 2A-A 12HGA11J52 32 5.4 GE 84376,CONNECTION DWGS TVA DWG 45W724-1, 6.9 KV SHUTDOWN BOARD GE TYPE 45W760-211-4, 45BM247-2, 2-BD-211-B-B 28-B 12HGA11J52 32 5.4 GE 84376,CONNECTION DWGS WVA DWG 45W749-1, 480V SHUTDOWN BOARD W 75767A -DWGS 6947001, 1-BD-212-Al-A 1A1-A W TYPE SG 5 5.5 6947D02, 6947D10, 618F932 TVA DWG 45W749-2, 480V SHUTDOWN BOARD W 75767A -DWGS 6947DO1, 1-BD-212-A2-A 1A2-A W TYPE SG 5 5.5 6947D02, 6947D18, 6947D19 TVA DWG 45W749-3, 480V SHUTDOWN BOARD W 75767A -DWGS 6947D01, 1-BD-212-B1-B 181-B W TYPE SG 5 5.5 6947D02, 6947D27 TVA DWG 45W749-4, 480V SHUTDOWN BOARD W 75767A -DWGS 6947)01, 1-BD-212-B2-B 182-B W TYPE SG 5 5.5 6947D36, 6947D37 WBN2 IPEEE DESIGN REPORT Page 202 ATTACHMENT 2

TENNESSEE VALLEY AUTHORITY Title: Watts Bar Nuclear Plant, Report No.IPEEE Seismic Margins Evaluation, Unit 2 Relay Design Report WBNIPEEE-004 Prepared By: Michael G. Eason Date: 04/26/2010 Page 24 of 25 Bad Actor See Component ID Component Description Relay TyDe # of Relays Section REFERENCE DOCUMENTS TVA DWG 45W749-1A, 480V SHUTDOWN BOARD W 75767A -DWGS 6947D01, 2-BD-212-A1-A 2A1-A W TYPE SG 3 5.5 6947D58, 618F938 TVA DWG 45W749-2A, 480V SHUTDOWN BOARD W 75767A -DWGS 6947D01, 2-BD-212-A2-A 2A2-A W TYPE SG 3 5.5 6947D66, 6947D67 TVA DWG 45W749-3A, 480V SHUTDOWN BOARD W 75767A -DWGS 6947D01, 2-BD-212-B1-B 2B1-B W TYPE SG 3 5.5 6947D75, 618F941 480V SHUTDOWN BOARD TVA DWG 45W749-4A, 2-BD-212-B2-B 2B2-B W TYPE SG 3 5.5 W 75767A -DWGS 6947D01, 6947D84, 6947D85 TVA DWG 45W706-1, VTM W120-2064, 120VAC VITAL INSTRUMENT W 85216 DWG.1-BD-235-0001-D POWER BOARD 1-I W TYPE SG 2 5.6 CP-33419-MKE-BM-1, -2, -3, -4,-5 WVA DWG 45W706-1, VTM W120-2064, 120VAC VITAL INSTRUMENT W 85216 DWG 2-BD-235-0001-D POWER BOARD 2-1 W TYPE SG 2 5.6 CP-33419-MKE-BM-1, -2, -3, -4, -S TVA DWG 45W706-1, VTM W120-2064, 120VAC VITAL INSTRUMENT W 85216 DWG 1-BD-235-0002-E POWER BOARD 1-11 W TYPE SG 2 5.6 CP-33419-MKE-BM-1, -2,-3, -4, -S TVA DWG 45W706-1, VTM W120-2064, 120VAC VITAL INSTRUMENT W 85216 DWG 2-BD-235-0002-E POWER BOARD 2-11 W TYPE SG 2 5.6 CP-33419-MKE-BM-1, -2, -3, -4, -5 WVA DWG 45W706-1, VTM W120-2064, 120VAC VITAL INSTRUMENT W 85216 DWG 1-BD-235-0003-F POWER BOARD 1-111 W TYPE SG 2 5.6 CP-33419-MKE-BM-1, -2, -3, -4,-5 TVA DWG 45W706-1, VTM W120-2064, 120VAC VITAL INSTRUMENT W 85216 DWG 2-BD-235-0003-F POWER BOARD 2-111 W TYPE SG 2 5.6 CP-33419-MKE-BM-1, -2, -3, -4, -5 WVA DWG 45W706-1, VTM W120-2064, 120VAC VITAL INSTRUMENT W 85216 DWG 1-BD-235-0004-G POWER BOARD 1-IV W TYPE SG 2 5.6 CP-33419-MKE-BM-1, -2, -3, -4,-5 TVA DWG 45W706-1, VTM W120-2064, 120VAC VITAL INSTRUMENT W 85216 DWG 2-BD-235-0004-G POWER BOARD 2-IV W TYPE SG 2 5.6 CP-33419-MKE-BM-1, -2, -3, -4, -5 WBN2 IPEEE DESIGN REPORT Page 203 ATTACHMENT 2

TENNESSEE VALLEY AUTHORITY Title: Watts Bar Nuclear Plant, Report No.IPEEE Seismic Margins Evaluation, Unit 2 Relay Design Report WBNIPEEE-004 Prepared By: Michael G. Eason Date: 04/26/2010 Page 25 of 25 Bad Actor See Component ID Component Description Relay Type # of Relays Section REFERENCE DOCUMENTS JUNCTION BOX -LOCA CONTAINMENT H2 SAMPLING ISOLATION GE TYPE 45W600-43-2, 47W611-88-1, 1-JB-293-6294 VALVES 12HGAIIJ52 2 5.8 45W163-80 JUNCTION BOX -LOCA CONTAINMENT H2 SAMPLING ISOLATION GE TYPE 45W600-43-2, 47W611-88-1, 2-JB-293-6294 VALVES 12HGAlIJS2 2 5.8 45W163-80 JUNCTION BOX -LOCA CONTAINMENT H2 SAMPLING ISOLATION GE TYPE 45W600-43-2, 47W611-88-1, 1-JB-293-6295 VALVES 12HGA1_JS2 2 5.8 4SW163-80 JUNCTION BOX -LOCA CONTAINMENT H2 SAMPLING ISOLATION GE TYPE 45W600-43-2, 47W611-88-1, 2-JB-293-6295 VALVES 12HGA11IJ52 2 5.8 45W163-80 WBN2 IPEEE DESIGN REPORT Page 204 ATTACHMENT 2

Revision 0 April 27, 2010 Attachment 3: Watts Bar Nuclear Plant Unit 2 Individual Plant Examination for-External Events Summary of Fire Evaluation Process Report No. WBN-IPEEE-005 WBN2 IPEEE DESIGN REPORT Page 205 I WATTS BAR NUCLEAR PLANT UNIT 2 WBN-IPEEE-005 INDIVIDUAL PLANT EXAMINATION FOR EXTERNAL EVENTS

SUMMARY

OF FIRE EVALUATION PROCESS Prepared By J. 1eEIAegk j,/ Signature Print /Date'Reviewed By S s4Z~ 1 Signature Print Date Approved By 0414(11 Sinaur Print Dt WBN2 IPEEE DESIGN REPORT Page 206 ATTACHMENT 3

2 TABLE OF CONTENTS Executive Summary .......................................................................................................

4 1 .In tro d u c tio n ......................................................................................................................

5 1.1 Overview of the FIVE Methodology

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5 1.2 Implementation of the EPRI FIVE Methodology

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6 2. Qualitative Screening and Fire Compartment Interaction Analysis (FIVE Phase I) ... 11 2.1 General Description of the Watts Bar Nuclear Plant .................................

11 2.2 Fire Compartment Interaction Analysis .........................................................

12 3. Evaluation of Fire Ignition Frequencies (FIVE Phase II, Step 1) .............................

14 3.1 Fire Areas and Zones ....................................................................................

14 3.2 Plant Wide Components

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16 3.3 Results of Fire Ignition Frequency Evaluation

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19 4 Initial PRA Model Evaluation of Fires (FIVE Phase II, Step 2) ................................

23 5 Revised PRA Model Evaluation of Fires (FIVE Phase II, Step 3) ............................

27 6. Results (FIVE Phase III, Step 1) ...............................................................................

31 7. New and Remaining Issues (FIVE Phase III, Step 2) ..............................................

36 7.1 Evaluation of Containment Heat Removal and Isolation

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36 7.2 Resolution of Sandia Fire Risk Scoping Study Issues .................................

37 7.3 Requirements of NUREG-1407

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49 8. Validation of the Unit 2 "As Designed" FIVE ................................................

55 9 .R e fe re n c e s ....................................................................................................................

5 7 WBN2 IPEEE DESIGN REPORT Page 207 ATTACHMENT 3

3 Table 2.1 Table 3.1 Table 3.2 Table 3.3 Table 3.4 Table 4.1 Table 5.1 Table 5.2 Table 6.1 TABLES Areas Screened during FIVE Phase I ....................................................

13 Tabulation of Generic Plant Area Types ..................................................

15 Fire Ignition Sources and Frequencies by Applicable Plant Locations

....... 17 Tabulation of Plant-Wide Fire Ignition Sources .....................................

18 Fire Ignition Frequencies for Unscreened Areas ...................................

20 Areas Screened in Phase II, Step 2 ......................................................

25 Areas Within 30% of Screening Criteria .......................

29 Areas Screened in Revised Evaluation

.................................................

30 Summary of Area Screening Process ...................

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32 FIGURES Summary Report Organization

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9 Organization of Reports ...............................................................................

10 Sample Fire Ignition Calculation Worksheet

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18 Example Case Development Using Event Tree Structure

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28 Figure 1-1 Figure 1-2 Figure 3-1 Figure 5-1 WBN2 IPEEE DESIGN REPORT Page 208 ATTACHMENT 3

4 EXECUTIVE

SUMMARY

In 1997 Watts Bar Nuclear Plant submitted a response to Supplement 4 to Generic Letter 88-20 that determined the plant vulnerability to internal fire events for Unit 1 operation.

That evaluation was based on the Fire Induced Vulnerability Evaluation (FIVE) methodology developed by the Electric Power Research Institute.

This report summarizes the evaluation performed for the "as designed" configuration of Watts Bar Unit 2 and addresses the Unit 2 compartments (rooms) that were not addressed for the Unit 1 report and also includes any compartments that contain Unit 2 required components.

The Unit 2 evaluation is also based on the same FIVE methodology as the Unit 1 report. Since this evaluation is based on the "as designed" plant configuration, there will be a validation effort performed at the completion of Unit 2 that will be based on the "as built" plant configuration.

The FIVE methodology consists of a progressive screening evaluation in which plant fire areas are screened from consideration based on qualitative information (Phase I)or by quantitative analysis (Phase II). Approximately 10% (17 of 140) of the compartments were screened in Phase I (see Section 2). This phase consists of screening fire areas based on area fire boundary integrity and the absence of safe shutdown components and plant trip initiators.

The Unit 2 Reactor Building was screened based on qualitative factors.The quantitative analysis (Phase II) then consists of an initial quantitative evaluation (see Section 4), followed by a more detailed quantitative evaluation (see Section 5)for areas that were not screened, based on a fire-induced core damage frequency of less than 1 E-06.The initial quantitative evaluation consists of generating a room-specific fire ignition frequency, then assuming that the postulated fire damaged all components in the fire affected room. Plant components that could be damaged by a postulated fire included safety injection/RHR (SI/RHR) components and components identified in the dual unit Appendix R analysis (WBN-EEB-EDQ00099920090012).

A "conditional core damage frequency" was then generated for each area by incorporating the failed components into the Probabilistic Risk Assessment (PRA) model. Over 60%(71 of 116) of the rooms were screened during this phase.The detailed quantitative evaluation was then performed for the remaining rooms.For those Control Building, Turbine Building and Auxiliary Building areas that did not screen at an initial level of evaluation, more detailed review techniques were used (e.g., zone of influence reviews for potential fires, segmentation of fire scenarios using event trees).The results of the detailed evaluation process were that all remaining plant areas were screened from further consideration while maintaining a conservative level of assumed system failures within the analysis.

This "as designed" evaluation did not identify any fire-induced vulnerabilities associated with the addition and operation of Unit 2 at Watts Bar.WBN2 IPEEE DESIGN REPORT Page 209 ATTACHMENT 3

5

1.0 INTRODUCTION

This report describes the process used to evaluate the fire hazards for Unit 2 at Watts Bar Nuclear Plant and the results of that process. The evaluation was performed in response to the Individual Plant Examination for External Events (IPEEE) requested by Supplement 4 of Generic Letter 88-20 (NUREG-1407, Reference 19). The methodology used to perform this examination is based on the Fire Induced Vulnerability Evaluation (FIVE) methodology that was developed by the Electric Power Research Institute (EPRI), as described in Reference 2.The scope of the evaluation included the rooms that contain systems and equipment required to safely shutdown the unit in the event of a fire as identified in the Unit 2 Appendix R analysis.

The evaluation was based on the "as designed" configuration of Unit 2. A validation of this analysis will be conducted when plant construction is complete to confirm this evaluation.

1.1 Overview

of the FIVE Methodology The EPRI FIVE methodology was used as a basis for evaluation of fire hazards and for screening fires from further consideration, based on screening criteria of less than 1 E-06 core damage frequency due to fire related initiating events.The FIVE documentation describes the fire evaluation process in three phases. The steps involved in each of these phases are described below.Phase I Qualitative screening and fire compartment interaction analysis During this phase, plant areas can be removed from further consideration based on the absence of safe shutdown equipment and no identified need for plant trip. In addition, fire boundaries are reviewed to ensure that a fire could not develop and then spread to other areas that may contain safe shutdown equipment or components.

Phase II Quantitative evaluation of plant areas This phase accounts for the largest portion of effort for the fire hazard evaluation process and consists of the following steps.Phase II (Step 1) identified room specific and generic plant fire hazards and their associated fire ignition frequencies for those rooms that were not screened out during the Phase I analysis.

A total fire ignition frequency for each plant area is calculated as the sum of the individual ignition source frequencies in that area. EPRI identifies this frequency as F 1 andý if it has a value of less than 1E-06, then the room can be screened from further evaluation.

It has been noted that, if a room has any ignition sources, the F 1 will always be greater than 1 E-06. This was the case with Unit 2.WBN2 IPEEE DESIGN REPORT Page 210 ATTACHMENT 3

6 Phase II (Step 2) evaluated the plant model impact for a fire in each of the rooms from Step 1 of this phase. Using the information from the Unit 2 Appendix R analysis, which assumed that any required component in the room was damaged by the fire, the PRA model was run. The results of this run generated a "conditional core damage frequency" for each room.This conditional core damage frequency is referred to as P 2 in the FIVE methodology.

The product of the Ignition Sources (Fl) and the conditional core damage frequency (P 2) results in a fire related core damage frequency (F 2).F 2 = F 1 x P 2 If F 2 is less than 1 E-06, the room can be screened from further evaluation.

Phase II (Step 3) If F 2 for a room is greater than 1 E-06, then additional evaluations were conducted.

These additional evaluations included determining the zone of influence from fire ignition sources and dividing the fires into severity probabilities using an "event tree" approach.PHASE III Results and Issues The final phase of the fire evaluation process consists of the documentation of results and identification of any new or remaining issues, including those addressed by the Sandia Fire Risk Scoping Study (NUREG/CR-5088) and the evaluation of containment performance.

1.2 Implementation

of the EPRI FIVE Methodology The implementation of the EPRI FIVE methodology and the organization of this report is shown graphically in Figure 1-1. This implementation can be described as follows.Phase I The qualitative screening process is described in Section 2 of this report. A Fire Compartment Interaction Analysis (FCIA) was performed to determine if a given room contained any components required for fire safe shutdown (FSSD) or could require a plant trip. In addition, the room was evaluated for the potential for fire spread from an exposed room to an adjacent room.There were 17 rooms out of 140 rooms, listed in Table 6.1, screened from further evaluation through this process. This evaluation is documented in the Fire Compartment Interaction Analysis Report (Reference 8).Phase II The quantitative evaluation of the fire hazard frequency for each of the remaining rooms was then performed.

This was based on the guidance given in the EPRI FIVE documentation, which was implemented in a three step process.WBN2 IPEEE DESIGN REPORT.Page 211 ATTACHMENT 3

7 Phase II (Step 1) used the guidance in the EPRI FIVE documentation to generate fire ignition frequencies (F 1 values) for each room. These calculations are based on the plant-specific data listed in Sections 2 and 3.This process consisted of first allocating a plant area fire ignition frequency based on the assignment of each plant location to a generic type of area such as "Auxiliary Building, Switchgear Room", etc. Then the ignition sources were identified in each room and assigned a fire ignition frequency using the information provided by the FIVE documentation.

The sum of the individual fire ignition frequencies becomes the room's total fire ignition frequency (F 1). This evaluation is documented in the Ignition Source Data Report (Reference 10), and is summarized in Section 3 of this report. Only the Unit 2 Reactor Building and Annulus were screened, based on qualitative factors, from further evaluation at this level of evaluation.

Phase II (Step 2) assumed that all components in the room were impacted (i.e. failed) by a postulated fire. The credited safe shutdown components located outside the room with the fire were then used as input to the PRA model and their unavailability factor was determined and this factor is referred to as P 2.The product of the room fire frequency, F 1 and the conditional core damage frequency P 2 yields a value that is the probability of a fire in the room causing core damage. This fire damage frequency number is referred to as F 2 and, if it is less than 1 E-06, then that room can be screened from further evaluation.

This process is documented in the Determination of Fire Scenario Safe Shutdown Path Unavailability Report (Reference 5), which is summarized in Section 4 of this report. There were 71 rooms screened from further consideration at this level of analysis.Phase II (Step 3) is a more detailed look at those rooms remaining for further evaluation after Step 2 with a F 2 value greater than 1 E-06. This evaluation consisted of a review of significant ignition sources in each room and the plant components (e.g., equipment and cables) that could be within the zone of influence of the ignition sources. The EPRI FIVE methodology was used to determine the zone of influence for each fire source.Supporting documentation for this phase of the evaluation regarding walkdowns and zone of influence calculations is provided by Fire Damage Zone of Influence Report (Reference 12).The remaining rooms were also evaluated using a probabilistic model of fire behavior (i.e., an "event tree" methodology).

This was accomplished by segmenting the fire ignition frequency into individual cases for evaluation, based on fire severity, and area fire suppression.

This phase of the fire evaluation is summarized in Section 5 of this report.The Main Control Room was evaluated by using the guidance provided in Appendix M of the Fire Risk Analysis Implementation Guide (Reference 3)..This evaluation consisted of a review of the control functions that could be affected by potential fires in various locations within the Main Control Room WBN2 IPEEE DESIGN REPORT SPage 212 ATTACHMENT 3

8 and is documented in the IPEEE (Fire) Quantitative Screening (Phase II -Detailed Analysis)

Report (Reference 13).Phase III This phase documents the summary of the results of the fire hazards evaluation and also discusses the resolution of outstanding fire-related issues, including response to the issues arising from the Sandia Laboratories Fire Risk Scoping Study (NUREG/CR-5088, Reference 6).Phase III (Step 1) consists of documenting the results of the fire hazard evaluation performed in Phase I and Phase I1. These results are presented in the individual sections of this report and are summarized in Section 6.Phase III (Step 2) provides the resolution of any outstanding fire-related issues, including containment isolation and heat removal, response to the issues arising from the Sandia Laboratories Fire Risk Scoping Study (NUREG/CR-5088) and resolution of the requirements of NUREG-1407 (Reference 19). The resolution of these issues is discussed in Section 7 of this report.The organization of the various reports generated through the fire evaluation process is shown in Figure 1-2.WBN2 IPEEE DESIGN REPORT Page 213 ATTACHMENT 3

9 FIGURE 1-1 FIRE INDUCED VULNERABILITY EVALUATION (FIVE)

SUMMARY

REPORT ORGANIZATION PHASE I: QUALITATIVE ANALYSIS FIRE SAFE SHUTDOWN EQUIPMENT FIRE COMPARTMENT INTERACTION ANALY (SECTION 2)PHASE II: QUANTITATIVE ANALYSIS I STEP 1: FIRE IGNITION FREQUENCY (F,)(SECTION 3)SCREENED ROOMS SIS LISTED IN TABLE 2.1 CONTAINMENT AREAS SCREENED N ROOMS LISTED IN /TABLE 3.4 SCREENED ROOMS LISTED IN TABLES 4.1 & 5.2 STEP 2: INITIAL QUANTITATIVE EVALUATIO]

F 2 = FIx P 2 < 1E-06 (SECTION 4)F 2 = F 1 x P 2 > 1/E-06 STEP 3: REVISED QUANTITATIVE EVALUAT F 2 = FIx P 2 < 1E-06 (SECTION 5)PHASE III: RESULTS AND REMAINING ISSUES ISTEP 1: DOCUMENTATION OF RESULTS (SECTION 6)4,L STEP 2: RESOLUTION OF REMAINING ISSUES (SECTION 7)FIGURE 1-2 WBN2 IPEEE DESIGN REPORT Page 214 ATTACHMENT 3

10 FIRE INDUCED VULNERABILITY EVALUATION (FIVE)ORGANIZATION OF REPORTS PHASE I: QUALITATIVE ANALYSIS I, FIRE COMPARTMENT INTERACTION ANALYSIS PHASE II: QUANTITATIVE ANALYSIS FIRE COMPARTMENT INTERACTION ANALYSIS REPORT (WBN-IPE-001 U2)STEP 1: FIRE IGNITION FREQUENCY IGNITION SOURCE DATA REPORT (WBN-IPE-002 U2)STEP 2: INITIAL QUANTITATIVE EVALUATION DETERMINATION OF FIRE SAFE SHUTDOWN PATH UNAVAILABLITY REPORT (WBN-IPE-003 U2)I STEP 3: REVISED QUANTITATIVE EVALUATION ZONE OF INFLUENCE REPORT (WBN-IPE-004 U2)QUANTITATIVE SCREENING DETAILED ANALYSIS REPORT (WBN-IPE-005 U2)PHASE HI: RESULTS AND REMAINING ISSUES STEP 1: DOCUMENTATION OF RESULTS STEP 2: RESOLUTION OF REMAINING ISSUES-F

SUMMARY

OF FIRE EVALUATION PROCESS (THIS DOCUMENT)WBN2 IPEEE DESIGN REPORT Page 215 ATTACHMENT 3

11 2.0 QUALITATIVE SCREENING AND FIRE COMPARTMENT INTERACTION ANALYSIS (FIVE PHASE I)The initial level of evaluation described in the FIVE documentation evaluates the various areas of the plant based on qualitative factors such as the presence of equipment required to perform a safe shutdown or equipment that could initiate a plant trip. This process is described in the Fire Compartment Interaction Analysis Report (Reference 8), which is summarized in Section 2.2.Section 2.1 provides a general description of the Watts Bar Nuclear Plant.2.1 General Description of the Watts Bar Nuclear Plant Watts Bar Nuclear Plant is a two unit Westinghouse Pressurized Water Reactor plant.Only Unit 1 is complete and operating, and Unit 2 is currently under construction.

This report addresses the Unit 2 rooms that were not included in the Unit 1 IPEEE report and re-evaluates those Unit 1 and common rooms that also contain Unit 2 components (equipment and cables). The major buildings included in this evaluation are Auxiliary Building Control Building Emergency Diesel Generator Building (4 separate diesels -2 per unit)Intake Pumping Station Unit 2 Reactor Building Turbine Building All of the buildings contain components that support the operation of both of the reactors, except the Reactor Buildings (i.e., Unit 1 Reactor Building contains components required only for Unit 1 reactor and Unit 2 Reactor Building contains components required only for Unit 2 reactor).The Auxiliary Building contains most of the support equipment required for safe shutdown (e.g., Charging Pumps, Safety Injection Pumps, RHR Pumps, Component Cooling Water Pumps, Auxiliary Feedwater Pumps, 6.9KV and 480-V shutdown boards and motor control centers, Auxiliary Control Room, etc.) and is separated from adjacent buildings by 3-hour rated fire resistive barriers (12-inch to 36-inch thick reinforced concrete).

Most of these components are located in individual rooms with at least 12-inch thick concrete walls, floors, ceilings with a minimum fire resistive rating of 2-hours. For example, the Train A and B Charging Pumps are located in separate rooms that have 2-hour fire resistive ratings. Train A 6.9KV Shutdown Board Room is separated from Train B 6.9KV Shutdown Board Room by 2-hour fire resistive barriers.The individual rooms and their fire resistive ratings are documented on the Fire Compartmentation drawings (47W240 series). In addition to the fire resistive barriers, the individual rooms (except for high radiation tank rooms, heat exchanger rooms, etc.) are also provided with smoke and/or heat detectors and most are provided with automatic suppression.

The pump rooms without automatic suppression have very low combustible loads (i.e., less that 5,000 Btu/ft 2 , which equates to a fire severity of less than 4 minutes).WBN2 IPEEE DESIGN REPORT Page 216 ATTACHMENT 3

12 The Control Building contains the Main Control Room, the Spreading Room and rooms that contain balance of plant components such as the 250-V batteries and battery boards, Auxiliary Instrument Rooms, Plant Computer Room and Relay Room. The Control Building is separated from other buildings with 3-hour fire resistive barriers (typically 36-inch thick reinforced concrete).

The rooms within the Control Building are generally separated from each other with 2-hour fire resistive barriers (typically 8-inch to 12-inch thick reinforced concrete).

The rooms are provided with smoke detectors (and thermal detectors for the Auxiliary Instrument Rooms and Plant Computer Room)and most are protected by an automatic suppression system. Smoke detectors are also provided inside selected Main Control Room cabinets.The Emergency Diesel Generator Building contains the four (two train A and two train B) emergency diesel generators and their associated electric boards. Each diesel generator unit is separated from the other by 3-hour rated fire resistive barriers (typically 12-inch thick reinforced concrete).

The diesel generator rooms and their associated electric board rooms are provided with a fire detection system and automatic CO 2 suppression system.The Intake Pumping Station (IPS) contains the Essential Raw Cooling Water Pumps and the electric driven fire pumps and the associated strainers for these pumps. The Raw Cooling Water pumps are located on an outside deck that is part of the Intake Pumping Station. The electric board room is provided with a fire detection system and automatic suppression system. The strainer rooms are provided with a detection system. The ERCW pumps and electric driven fire pumps are located on the upper floor of the Intake Pumping Station which is open to the atmosphere.

The Train A portions of the IPS are typically separated from the Train B portions by (12-inch to 36-inch thick reinforced concrete) barriers that have a fire resistive rating of 3-hours.2.2 Fire Compartment Interaction Analysis (FCIA)The qualitative evaluation consists primarily of reviewing the various plant fire compartments for two criteria: 1. The compartment does not contain Appendix R safe shutdown components or plant trip initiators.

2. There is no potential for the fire to spread to adjacent compartments (i.e. no concentration of combustibles near the boundary, lack of continuous combustible pathway between compartments, and no unprotected openings).

The FIVE methodology allows a compartment to be screened if the above conditions are true. The rooms that were screened from further consideration based on these considerations are listed in Table 2.1.WBN2 IPEEE DESIGN REPORT Page 217 ATTACHMENT 3

13 TABLE 2.1 AREAS SCREENED DURING FIVE PHASE I (INFORMATION TAKEN FROM REPORT WBN-IPE-001 U2)L FIRE ROOM 1 ROOM AREA NO. J DESCRIPTION COMMENTS 1 692.0-A5 Gas Decay Tank Room 692.0-A27 Concentrate Filter Room 692.0-A29 Boric Acid Evaporator Package B Room 8 713.0-A4 Radio-Chemistry Lab 71 713.0-A21 Unit 2 Reactor Building Access 8 713.0-A23 CVCS Valve Gallery 14 737.0-A8 Unit 2 Letdown Heat Exchanger 72 737.0-Al0 Air Lock 14 737.0-Al1 Air Lock 73 737.0-A14 Air Lock 74 737.0-A16 Unit 2 Gross Failed Fuel Detector Room 10 786.0-Al Fan Room 47 786.0-A2 Roof Access Air Lock 786.0-A3 Mechanical Equipment Room 786.0-A4 Mechanical Equipment Room 60 728.IPS-RSW Intake Pumping Station -Outdoor Deck Deck containing RSW Pumps 53 742.0-D8 Fuel Oil Transfer Valve Room The quantitative evaluation of the rooms that were not screened from further evaluation in Phase I was then performed in Phase II using a three step process.Step 1 consisted of the development of fire ignition frequencies for each of the remaining rooms. This process is described in WBN-IPE-002 U2 and summarized in Section 3, below.Step 2 consisted of an initial quantitative evaluation of each room, assuming that all non-Appendix R and unprotected Appendix R (i.e. only one train protected from fire damage) equipment for each room was failed by postulated fire in that room.This process is described in WBN-IPE-003 U2 and summarized in Section 4, below.Step 3 consisted of a revised quantitative evaluation of each of the remaining rooms, evaluating each room on an individual basis for the relaxation of overly conservative assumptions from Step 2. This process is described in WBN-IPE-005 U2 and summarized in Section 5, below.WBN2 IPEEE DESIGN REPORT Page 218 ATT-ACHMENT 3

14 Note: Of the individual entries listed in Table 6.1, the following 5 areas were not considered to be separate rooms for purposes of room count for calculation of fire ignition frequency:

Stair D-1 814.75-ACS Areas Screened in Phase I DG Building Stairs Roof Access Included with 760.5-D1 Included with 786.0-Al Areas Included in Quantitative Analysis (Phase II)692.0-A4 Chemical Drain Tank Room Included with 692.0-Al 713.0-A22 Valve Gallery Included with 713.0-Al 763.5-A2 U2 Additional Equip Bldg Included with 729.0-A15 742.0-D3 Toilet Included with 742.0-D9 755.0-C20 DPSO Shop Included with 755.0-Cl 3 3.0 EVALUATION OF FIRE IGNITION FREQUENCIES (FIVE PHASE II, STEP 1)The calculation of fire ignition frequencies for the rooms that remained for quantitative screening is described in the Ignition Source Data Report (WBN-IPE-002 U2). This evaluation resulted in fire ignition frequencies above 1 E-06 for all rooms evaluated, with the exception of the Unit 2 Reactor Building, which was screened based on specific guidance given in the FIVE documentation.

Therefore, only the Unit 2 Reactor Building (includes Annulus) was screened from further evaluation in this step. For reference, the rooms that remain for quantitative evaluation in Steps 2 and 3 of Phase II are listed in Table 3.4, below, along with the associated fire ignition frequencies.

3.1 Fire Areas and Fire Zones The first part of the fire ignition frequency evaluation methodology described in the FIVE documentation requires that the various plant areas be assigned to a generic type. Those generic types described in FIVE are listed below in Table 3.1 along with the number of each type at Watts Bar.WBN2 IPEEE DESIGN REPORT Page 219 AI-TACHMENT 3

15 TABLE 3.1 TABULATION OF GENERIC PLANT AREA TYPES PLANT LOCATION TYPE NUMBER OF LOCATIONS AUXILIARY BUILDING I DIESEL GENERATOR ROOM 6 SWITCHGEAR ROOM 17 BATTERY ROOM 9 CONTROL ROOM 1 CABLE SPREADING ROOM 1 INTAKE STRUCTURE 1 TURBINE BUILDING 1 RADWASTE AREA I TRANSFORMER YARD I The Watts Bar Appendix R analysis considered 94 separate fire areas. The Unit 1 report considered all of these fire areas except the rooms containing Unit 2 components not required for Unit 1 operation.

Each of these fire areas is separated from adjacent fire areas by rated fire barriers.

The Auxiliary Building is comprised of 75 fire areas (numbers 1 through 47, 1-1, 1-2, 2-1, 2-2, 3-1, 3-2, 9-1, 15-1, 15-2, 65 through 77 and 71-1). The Control Building is one fire area (number 48) and is subdivided into 37 fire compartments.

The Diesel Generator Building is comprised of 8 fire areas (numbered 49 through 56). The Additional Diesel Generator Building is fire area 57. The Intake Pumping Station is comprised of three fire areas (numbered 58 through 60). The Unit 1 Reactor Building is fire area 61 and the Unit 2 Reactor Building is fire area 77. The Condensate Demineralizer Waste Evaporator Building is fire area 62; the Turbine Building is fire area 63; and the Yard is fire area 64. The Unit 2 IPEEE FIVE analysis addresses the rooms that were not considered during the Unit 1 analysis.

The Unit 2 analysis also re-evaluates the Unit 1 report rooms that contain components required for fire safe shutdown of Unit 2.For the remainder of this report the terms fire area, fire zone and fire compartment will be used interchangeably to refer to the evaluation of an individual plant area/room/compartment.

Table 6.1 lists each fire area and provides a description of the room or rooms that comprise the fire area.Each of these plant rooms were then assigned to a generic type of area corresponding to Table 3.1. The determination of fire ignition frequency for each of these rooms was based on the type of plant location and is documented in the Ignition Source Data Report (WBN-IPE-002 U2). The results of this evaluation are summarized in Table 3.4 of this report. A sample calculation worksheet is shown in Figure 3.1 and it details the calculation of fire ignition frequency for room 713.0-Al, which is used as an example in Section 5 of this report.WBN2 IPEEE DESIGN REPORT Page 220 ATT-ACHMENT 3

16 3.2 Plant Wide Components The next step in determining the overall fire ignition frequency for a room is to assign each ignition source in the room a generic ignition frequency which is provided in the FIVE data reference tables (see Table 3.2 for a reproduction of that data from Reference 2). The FIVE methodology also provides a means of determining a weighting factor for each ignition source based upon the total number of like components in the plant or similar area and the number of ignition sources in the room.The number of ignition sources was derived primarily from the Combustible Loading Summary (drawing 47W893-035), the Appendix R analysis, walkdowns and the Unit 1 IPEEE report. Table 3.3 provides a summary of the number of the various types of ignition sources in the plant.WBN2 IPEEE DESIGN REPORT Page 221 ATTACHMENT 3

17 TABLE 3.2 FIRE IGNITION SOURCES AND FREQUENCIES BY APPLICABLE PLANT LOCATIONS IGNITION SOURCE FIRE PLANT FIRE IGNITION SOURCE WEIGHTING FREQUENCY LOCATION FACTOR 1,2 Auxiliary Building Electrical Cabinets B 1.9E-02 Pumps B 1.9E-02 Diesel Generator Room Diesel Generators A 2.6E-02 Electrical Cabinets A 2.4E-03 Switchgear Room Electrical Cabinets A 1.5E-02 Battery Room Batteries A 3.2E-03 Control Room Electrical Cabinets A 9.5E-03 Cable Spreading Room Electrical Cabinets A 3.2E-03 Intake Structure Electrical Cabinets A 2.4E-03 Fire Pumps A 4.OE-03 Others A 3.2E-03 Turbine Building T/G Exciter B 4.OE-03 T/G Oil B 1.3E-02 T/G Hydrogen B 5.5E-03 Electrical Cabinets B 1.3E-02 Other Pumps B 6.3E-03 Main Feedwater Pumps A 4.OE-03 Boiler B 1.6E-03 Radwaste Area Miscellaneous Components A 8.7E-03 Transformer Yard Yard Transformers (propagating to Turbine Building)

A 4.OE-03 Yard Transformer (LOSP) A 1.6E-03 Yard Transformer (Others) F 1.5E-02 Plant-Wide Fire Protection Panels F 2.4E-03 Components RPS MG Sets F 5.5E-05 Non-qualified cable run E 6.3E-03 Junction Box/splice in non-qualified cable E 1.6E-03 Junction Box in qualified cable E 1.6E-03 Transformers F 7.9E-03 Battery Chargers F 4.OE-03 Hydrogen Tanks G 3.2E-03 Misc. Hydrogen Fires C 3.2E-03 Gas Turbines G 3.1E-02 5 Air Compressors F 4.7E-03 Ventilation Subsystems F 9.5E-03 Elevator Motors F 6.3E-03 Dryers F 8.7E-03 Transients D 1.3E-03 3,4 Cable fires caused by welding C 5.1E-03 4 Transient fires caused by 4 welding and cutting C 3.1E-02 1. Frequencies are per reactor year unless otherwise noted. Reactor year is the sum each reactor multiplied by the calendar years between operating license issuance date and December 31, 1988.2. Fire frequencies are per fraction of ignition sources per year.3. Fire frequency represents one event. The thirteen transient events which occurred during power operation are considered by the weighting factor.4. Fire frequency represents years at power operation.

WBN2 IPEEE DESIGN REPORT Page 222 ATTACHMENT 3

18 5. Fire frequency represents an estimated 130 gas-turbine-operating years.TABLE 3.3 TARIII ATIAM ný D1 AKIT-WinCl CIDC I KMITIfMI QAIID9-TYPE OF COMPONENT NUMBER Air Compressors 16 Battery Chargers 34 Fire Protection Panels 45 Junction Boxes 1 2634 Hydrogen Storage Tanks 2 Motor Generator Sets 15 Non-qualified Cable Runs (Btu total) 5.99E+10 Transformers 114 Ventilation Sub-systems 334 1 Total number of junction boxes in the safety related areas of the plant.Figure 3-1 Sample Fire Ignition Calculation Worksheet Fire Area 8 Fire Compartment 713.0-Al Description Corridor Generic Location Number Number Ignition Room Fire Weighting Devices in Devices in Source Fire Frequency Factor Room Bldg/Plant Factor Frequency (f) (W) (a) (b) (WO) (F 1)Compartment Ignition Source Elec. Cabinets 1.90E-02 2 85 867 9.80E-02 3.73E-03 Pumps 1.90E-02 2 19 137 1 1.39E-01 5.27E-03 Plant Wide Ignition Source Battery Chargers 4.OOE-03 2 1 34 2.94E-02 2.34E-04 Cable Trays 6.30E-03 2 3.47E+09 5.99E+10 5.79E-02 7.30E-04 FP Panels 2.40E-03 2 2 45 4.44E-02 2.13E-04 HVAC 9.50E-03 2 10 334 2.99E-02 5.69E-04 Junction Boxes 1.60E-03 2 128 2634 4.86E-02 1.56E-04 Transformers 7.90E-03 2 4 114 3.51E-02 5.54E-04 Transients 1.30E-03 2 10 264 3.79E-02 9.85E-05 Welding Cable 5.1OE-03 2 1 264 3.79E-03 3.86E-05 Welding/Cutting 3.1OE-02 2 1 264 3.79E-03 2.35E-04 Total 1.16E-02 Notes: 1. f- from EPRI Table 1.2 (Table 3.2 in this report)2. WL -Number of units (reactors)/Number of buildings (Auxiliary Building)3. W, -alb 4. F, = fXWLXWl WBN2 IPEEE DESIGN REPORT Page 223 ATTFACHMENT 3

19 3.3 Results of Fire Ignition Frequency Evaluation The Reactor Building contains the reactor vessel, steam generators, pressurizer and reactor coolant pumps. This Reactor Building structure is separated from other buildings by 3-hour rated fire resistive barriers.

The Reactor Building was screened from the FIVE fire frequency determination because: 1. A hot gas layer is unlikely to form in most areas of the Reactor Building which can damage cables.2. Reactor Coolant Pump fires are unlikely to occur due to compliance with 10CFR50 Appendix R, Section 111.0, oil collection system.3. Previous fire PRA's did not show that Reactor Building fires are risk significant.

Therefore, a separate evaluation of fire ignition frequency was not generated for the Reactor Building.

Based on the above guidelines, a severe fire inside the Reactor Building is judged to be highly unlikely.The fire ignition frequencies for the remaining unscreened areas are shown in Table 3.4. This table includes 116 individual fire ignition frequencies used to evaluate the 121 individual areas (rooms 692.0-A4, 713.0-A22, 755.0-C20, 742.0-D3 and 763.5-A2 are combined with rooms 692.0-Al, 713.0-Al, 755.0-C13, 742.0-D9 and 729.0-A15, respectively, for purposes of fire ignition frequency development) that remained for further evaluation following the completion of Phase I.WBN2 IPEEE DESIGN REPORT Page 224 ATTACHMENT 3

20 TABLE 3.4 FIRE IGNITION FREQUENCIES FOR UNSCREENED AREAS (INFORMATION TAKEN FROM REPORT WBN-IPE=-002

[121 FIRE ROOM ROOM 1 FIRE IGNITION AREA NO. DESCRIPTIONJ FREQUENCY 1 676.0-Al Corridor 2.61 E-03 676.0-A14 Containment Spray Pump 2A-A Room 6.75E-04 676.0-A15 Containment Spray Pump 2B-B 7.17E-04 692.0-Al & A4 Corridor & Chemical Drain Tank Room 9.62E-03 692.0-A30 Boric Acid Evaporator Package Room A 1.41 E-03 1-1 692.0-A20 Safety Injection Pump 2B-B Room 1.04E-03 1-2 692.0-A19 Safety Injection Pump 2A-A Room 9.53E-04 2-2 676.0-A13 RHR Pump 2B-B Room 7.61E-04 3-2 676.0-A12 RHR Pump 2A-A Room 7.61 E-04 8 713.0-Al & A22 Corridor &Valve Gallery 1.16E-02 713.0-A14 Sample Room 2 5.6E-04 713.0-Al 5 Heat Exchanger 2A Room 7.46E-05 713.0-A16 Heat Exchanger 2B Room 7.58E-05 713.0-A17 Seal Water Heat Exchanger 2A 7.46E-05 10 729.0-A6 Nitrogen Storage Area 1.14E-04 729.0-A9 Unit 2 Post Accident Sampling Room 3.09E-03 757.0-A13 Refueling Room 1.46E-02 772.0-A9 HEPA Filter Plenum Room 3.35E-04 14 737.0-Al Corridor 1.14E-02 15-1 737.0-A3 Heat and Vent Equipment Room 1 3.88E-04 15-2 737.0-A12 Heat and Vent Equipment Room 2 5.14E-04 17 757.0-A2 6.9KV and 480-V Shutdown Board Room A 5.31 E-04 757.0-A9 Personnel and Equipment Access 8.59E-04 18 757.0-A3 125-V Vital Battery Board II Room 7.09E-04 19 757.0-A4 125-V Vital Battery Board I Room 7.95E-04 20 757.0-Al Auxiliary Control Room 3.71E-04 21 757.0-A25 Auxiliary Control Instrument-Room 1A 1.89E-04 22 757.0-A26 Auxiliary Control Instrument Room 1B 1.90E-04 23 757.0-A27 Auxiliary Control Instrument Room 2A 1.98E-04 24 757.0-A28 Auxiliary Control Instrument Room 2B 1.91E-04 25 782.0-Al Control Rod Drive Equipment Room I 2.95E-03 782.0-A2 Pressurizer Heater Transformer Room 1 9.91E-04 27 757.0-A5 480-V Shutdown Board Room 1 B 4.73E-04 28 757.0-A21 480-V Shutdown Board Room 2A 5.52E-04 29 757.0-A22 125-V Vital Battery Board Room IV 6.32E-04 30 757.0-A23 125-V Vital Battery Board Room III 6.32E-04 31 757.0-A17 Personnel and Equipment Access 5.14E-04 757.0-A24 6.9KV & 480-V Shutdown Board Room B 1.10E-03 32 772.0-Al 480-V Board Room 1A 6.10E-04 33 772.0-A2 480-V Board Room 1 B 1.91 E-03 34 772.0-A3 125-V Vital Battery II 4.27E-04 35 772.0-A4 125-V Vital Battery I 4.27E-04 36 772.0-A5 480-V Transformer Room 1 B 1.06E-03 WBN2 IPEEE DESIGN REPORT Page 225 ATTACHMENT 3

21 FIRE IGNITION FREQUENCIES FOR UNSCREENED AREAS (INFORMATION TAKEN FROM REPORT WBN-IPE-002 U2)FIRE ROOM ROOM FIRE IGNITION AREA NO. DESCRIPTION FREQUENCY 37 772.0-A6 480-V Transformer Room 1A 1.15E-03 38 772.0-A7 Mechanical Equipment Room 1.00E-03 39 772.0-A8 5 th Vital Battery & Battery Board Room 8.61E-04 40 772.0-A10 Mechanical Equipment Room 9.79E-04 41 772.0-Al1 480-V Transformer Room 2B 1.14E-03 42 772.0-Al 2 480-V Transformer Room 2A 1.04E-03 43 772.0-Al 3 125-V Vital Battery Room IV 4.27E-04 44 772.0-A14 125-V Vital Battery Room III 4.27E-04 45 772.0-Al 5 480-V Board Room 2B 2.36E-03 46 772.0-A16 480-V Board Room 2A 1.64E-03 47 786.0-AR Auxiliary Building Roof 4.75E-04 48 692.0-C5 250-V Battery Board Room 2 2.45E-03 692.0-C6 250-V Battery Room 2 4.43E-04 692.0-C7 24-V & 48-V Battery Room 4.15-04 692.0-C8 24-V & 48-V Battery Bd and Charger Room 3.07E-03 692.0-C9 Communications Room 2.49E-03 692.0-C10 Mechanical Equipment Room 1.42E-03 692.0-Cl 1 Corridor 6.96E-04 692.0-C12 Secondary Alarm Station 1.37E-03 708.0-C2 Corridor 5.84E-04 708.0-C3 Plant Computer Room 4.21 E-03 708.0-C4 Auxiliary Instrument Room 2 1.13E-02_729.0-Cl Cable Spreading Room 1.86E-03 755.0-C12 Main Control Room 1.94E-02 755.0-C13

& Relay Room & DPSO Shop 4.67E-03 C20 49 742.0-D4 Diesel Generator lA-A Room 5.93E-02 760.5-D3 Air Exhaust Room IA 8.86E-04 760.5-D4 480-V Board Room 1A 1.36E-03 760.5-D5 Air Intake Room 1A 7.92E-05 50 742.0-D5 Diesel Generator 2A-A Room 5.93E-02 760.5-D6 Air Exhaust Room 2A 7.06E-04 760.5-D7 480-V Board Room 2A 1.36E-03 760.5-D8 Air Intake Room 2A 2.93E-04 51 742.0-D6 Diesel Generator 1B-B Room 5.93E-02 760.5-D9 Air Exhaust Room I B 7.06E-04-760.5-D10 480-V Board Room 1B 1.36E-03 760.5-D11 Air Intake Room 1 B 7.92E-05 52 742.0-D7 Diesel Generator 2B-B Room 5.90E-02 760.5-D12 Air Exhaust Room 2B 7.06E-04 760.5-D13 480-V Board Room 2B 1.29E-03 760.5-D14 Air Intake Room 2B 7.92E-05 53 742.0-D3 & D9 Toilet & Corridor 1.24E-03 55 742.0 Train A Cable Chase 1.04E-05 56 742.0 Train B Cable Chase 1.04E-05 WBN2 IPEEE DESIGN REPORT Page 226 ATTACHMENT 3

22 FIRE IGNITION FREQUENCIES FOR UNSCREENED AREAS (INFORMATION TAKEN FROM REPORT WBN-IPE-002 U2)FIRE ROOM ROOM FIRE IGNITION AREA NO. DESCRIPTION FREQUENCY 58 722.1PS-STR A Train A ERCW Strainer Room 1.11 E-03 741.IPS-ERCW Train A ERCW Pump Room 1.49E-03 Pump Room A 741.IPS SWP A Train A Screen Wash Pump & Fire Pump -3.83E-03 Room 59 722.IPS-STR B Train B ERCW Strainer Room 1.04E-03 741.IPS-ERCW Train B ERCW Pump Room 1.49E-03 Pump Room B 741.1PS SWP B Train B Screen Wash Pump & Fire Pump 3.61 E-03 Room 60 711 .IPS-Elec 480-V Board Room 1.86E-03 BD Room 63 Turbine Turbine Building 7.19E-02 Building 65 676.0-A17 Pipe Gallery (Unit 2) 3.35E-04 692.0-A24 Unit 2 Pipe Chase & Valve Gallery 3.33E-04 713.0-A29 Unit 2 Pipe Chase & Valve Gallery 9.17E-04 66 692.0-A21 Charging Pump Room 2C 7.56E-04 67 692.0-A22 Centrifugal Charging Pump Room 2B-B 9.90E-04 68 692.0-A23 Centrifugal Charging Pump Room 2A-A 9.90E-04 69 692.0-A26 Turbine Driven AFW Pump Room 2A-S 1.53E-03 70 692.0-A25 Unit 2 Pipe Gallery 4.54E-03 71 713.0-A19 Unit 2 Pipe Gallery 1.86E-03 71-1 713.0-A20 Unit 2 Volume Control Tank Room 1.72E-03 72 729.0-Al 1 Unit 2 South Main Steam Valve Room 3.67E-04 73 729.0-Al0 Unit 2 North Main Steam Valve Room 5.02E-04 729.0-A13 Unit 2 Steam Valve Instrument Room 7.58E-05 729.0-A15

& Unit 2 Additional Equipment Building (all 2.01E-03 763.5-A2 elevations) 729.5-A17 Unit 2 Shield Building Vent Radiation Monitor 3.13E-04 Rm 74 737.0-A9 Unit 2 Ventilation

& Purge Air Room 8.19E-04 75 757.0-A14 Unit 2 Reactor Building Access Room 1.25E-04 757.0-A16 Emergency Gas Treatment Filter Room 3.93E-03 782.0-A3 Unit 2 Control Rod Drive Equipment Room 2.71 E-03 782.0-A4 Unit 2 Pressurizer Heater Transformer Room 8.94E-04 76 757.0-Al 5 Unit 2 Reactor Building Equipment Hatch 7.75E-05 WBN2 IPEEE DESIGN REPORT Page 227 ATTACHMENT 3

23 4. INITIAL PRA MODEL EVALUATION OF FIRES (FIVE Phase II, Step 2)Following the development of room fire ignition frequencies, as described in report WBN-IPE-002 U2 (Reference

10) and Section 3 of this report, the updated Watts Bar Nuclear Plant Probabilistic Risk Assessment (PRA) model (Reference
21) was modified in order to develop the conditional core damage frequency values required to generate a fire-related core damage frequency for each remaining room. This was performed in a two step process using the following general equation: F 2 = F 1 x P 2 Where F 2 is the fire-related core damage frequency for the room under evaluation.

F 1 is the room fire ignition frequency for the room under evaluation, (see WBN-IPE-002 U2).P 2 is the core damage frequency for the room under evaluation, assuming a fire ignition frequency of 1.0. This gives what is known as the "conditional" core damage frequency (CCDF). That is, given a fire with frequency of F 1 , this P 2 value can be used directly in the equation above to calculate the core damage frequency associated with such a fire.If the fire-related core damage frequency is less than 1E-06, the room can be screened from further evaluation.

As noted above, this process was performed in two main steps. The first of these is documented in report WBN-IPE-003 U2, entitled"Determination of Fire Scenario Safe Shutdown Path Unavailability

-,Watts Bar Nuclear Plant Unit 2" (Reference 5). This report is summarized in this section.The detailed portion of the quantitative evaluation is summarized in Section 5.During this step of the quantitative evaluation, any equipment not on the Appendix R Safe Shutdown Equipment List (SSEL), except for safety injection and RHR components (SI/RHR) that support bleed and feed cooling, was arbitrarily assumed to fail. The SI/RHR related equipment was credited because the equipment and cabling locations were identified, which allowed fire impacts to this equipment to be correctly modeled. The resulting "conditional" core damage frequency, based on the failure of the aforementioned equipment, was then multiplied by the fire ignition frequency for the area of interest shown in Table 3.4. If the resulting fire-related core damage frequency was below 1 E-06, the area could be screened from further consideration.

For the Control Building areas, a conditional core damage frequency of 1.0 was arbitrarily assumed. This assured that all remaining Control Building areas would be retained for detailed analysis.Areas outside the Control Building were evaluated using a modified version of the updated WBN plant model.WBN2 IPEEE DESIGN REPORT Page 228 ATTACHMENT 3

24 The data used as input for the model included (Reference 5).1. Unit 0 (common) and Unit 2 SSEL equipment and associated fire zones 2. Unit 0 and Unit 2 cable routings with associated fire zones, fire wrap commitments and end devices 3. Correlation between fire zones, room numbers and analysis volumes 4. SSEL equipment not to be assumed fully functional for fire analysis (i.e., without motive power)5. Table of cables which provide power to the 6.9KV shutdown boards 2-BD-21 1-A-A and 2-BD-211-B-B and their associated fire zones. The cables are labeled as to whether their failure in a fire would cause loss of offsite power to the associated shutdown board, cause only failure of power from the backup emergency diesel generators, or would cause failure of both power sources to the boards.6. Safety injection and recirculation related equipment with associated cable locations.

Items 1 and 2 above were used in this evaluation to identify the SSEL equipment and to associate them to fire zones along with their supporting cables and the fire zones the cables traverse.

The data in item 3 above correlates the analysis volumes to fire zones. These analysis volumes were developed during the Unit 1 Appendix R analysis (and continued in the Unit 2 analysis) to ensure that one redundant train of equipment required for safe shutdown would be in accordance with the separation criteria of Appendix R,Section III.G.2 and therefore remain free of fire damage. For the purposes of the screening evaluation, fire propagation from the fire zones in one analysis to another does not have to be assumed. The data in item 4 above was provided to identify those SSEL components that while included in the list, not all functions of the equipment can be credited.

For example, the control cables are included to allow tripping of the equipment, but cables for motive power are not traced and therefore credit for operation of the equipment cannot be taken. Nevertheless, failures involving transfer of valves to the undesired position are considered as a result of the fire (i.e., motive power is conservatively available for such cases). Item 5 above tabulates the routings for those cables which provide both offsite and onsite power to the 6.9KV shutdown boards. Item 6 identifies additional equipment not on the SSEL which were credited in this study. This added equipment is used to perform bleed and feed cooling in the event that AFW fails. Some of the equipment added were considered on the SSEL list but were considered not "fully functional".

This initial quantitative evaluation, including the incorporation of fire-related impacts into the Watts Bar PRA model structure, is documented in WBN-IPE-003 U2. This resulted in the screening of over half (71 of 116) of the rooms that remained for further quantitative evaluation.

For reference, the areas that were screened from further consideration are listed in Table 4.1, below. The evaluation of rooms that remained for further analysis is described in Section 5.WBN2 IPEEE DESIGN REPORT Page 229 ATTACHMENT 3

25 TABLE 4.1 AREAS SCREENED IN PHASE II, STEP 2 (INFORMATION FROM WBN-IPE-003 U2)FIRE ROOM ROOM ARE NO. DESCRIPTION F 1 P 2 F 2 A 1 676.0-Al Corridor 2.61E-03 8.88E-05 2.32E-07 3-2 676.0-A12 RHR Pump 2B-B Room 7.61E-04 7.62E-05 5.80E-08 2-2 676.0-A13 RHR Pump 2A-A Room 7.61 E-04 7.62E-05 5.80E-08 1 676.0-A14 Containment Spray Pump Room 2A-A 6.75E-04 1.88E-05 1.27E-08 1 676.0-Al5 Containment Spray Pump Room 2B-B 7.17E-04 1.48E-05 1.06E-08 65 676.0-A17 Unit 2 Pipe Gallery 3.35E-04 1.59E-03 5.30E-07 1 692.0-Al & A4 Corridor & Chemical Drain Tank Room 9.62E-03 8.88E-05 8.54E-07 1-2 692.0-A19 Safety Injection Pump Room 2A-A 9.53E-04 7.62E-05 7.27E-08 1-1 692.0-A20 Safety Injection Pump Room 2B-B 1.04E-03 1.66E-05 1.72E-08 66 692.0-A21 Charging Pump Room 2C 7.56E-04 1.48E-05 1.12E-08 67 692.0-A22 Charging Pump Room 2B-B 9.90E-04 1.57E-05 1.56E-08 68 692.0-A23 Charging Pump Room 2A-A 9.90E-04 1.65E-05 1.63E-08 65 692.0-A24 Unit 2 Pipe Chase & Valve Gallery 3.33E-04 1.59E-03 5.29E-07 69 692.0-A26 Turbine Driven AFW Pump 2A-S 1.53E-03 2.29E-05 3.51 E-08 1 692.0-A30 Boric Acid Evaporator Package Room A 1.41 E-03 8.21E-05 1.16E-07 8 713.0-A14 Sample Room 2B 5.60E-04 7.62E-05 4.27E-08 8 713.0-Al 5 Heat Exchanger 2A 7.46E-05 1.67E-05 1.25E-09 8 713.0-A16 Heat Exchanger 2B 7.58E-05 7.62E-05 5.78E-09 71-1 713.0-A20 Unit 2 Volume Control Tank Room 1.72E-03 6.40E-05 1.10E-07 10 729.0-A6 Nitrogen Storage Area 1.14E-04 4.97E-3 5.65E-07 10 729.0-A9 Fuel Transfer Valve Room 3.09E-03 1.48E-05 4.58E-08 73 729.0-A10 Unit 2 North Main Steam Valve Room 5.02E-04 1.48E-05 7.44E-09 72 729.0-Al I Unit 2 South Main Steam Valve Room 3.67E-04 3.23E-05 1 .19E-08 73 729.0-A13 Unit 2 Steam Valve Instrument Room A 7.58E-05 1.48E-05 1.12E-09 73 729.0-A15-Unit 2 Additional Equipment Building (all 2.01E-03 1.48E-05 2.98E-08 763.5-A2 elevations) 73 729.5-A17 Unit 2 Shield Building Vent Rad Monitor Rm 3.13E-04 1.48E-05 4.64E-09 15-1 737.0-A3 Heat and Vent Equipment Room 1 3.88E-04 1.63E-04 6.34E-08 74 737.0-A9 Unit 2 Ventilation and Purge Air Room 8.19E-04 2.54E-04 2.08E-07 15-2 737.0-A12 Heat and Vent Equipment Room 2 5.14E-04 4.65E-04 2.39E-07 18 757.0-A3 125V Vital Battery Board Room II 7.09E-04 9.62E-05.

6.82E-08 75 757.0-A14 Unit 2 Reactor Building Access Room 1.25E-04 8.41E-05 1.05E-08 76 757.0-A15 Unit 2 Reactor Building Equipment Hatch 7.75E-05 8.41E-05 6.51E-09 21 757.0-A25 Train A Instrument Room 1A 1.89E-04 8:83E-05 1.67E-08 22 757.0-A26 Train B Instrument Room 1B 1.90E-04 1.60E-05 3.05E-09 33 772.0-A2 480V Board Room 1B 1.91E-03 4.80E-05 9.16E-08 34 772.0-A3 125V Vital Battery Room II 4.27E-04 1.50E-05 6.41E-09 35 772.0-A4 125V Vital Battery Room I 4.27E-04 2.27E-05 9.71 E-09 36 772.0-A5 480V Transformer Room 1 B 1.06E-03 2.40E-05 2.55E-08 37 772.0-A6 480V Transformer Room 1A 1.15E-03 1.48E-05 1.71E-08 38 772.0-A7 Mechanical Equipment Room 1 1.001E-03 1.48E-05 1.49E-08 WBN2 IPEEE DESIGN REPORT Page 230 ATTACHMENT 3

26 AREAS SCREENED IN PHASE II, STEP 2 (INFORMATION FROM WBN-IPE-003 U2)FIRE ROOM ROOM ARE NO. DESCRIPTION F, P 2 F 2 A 39 772.0-A8 5 th Vital Battery & Board Room 8.61E-04 7.21E-05 6.21E-08 40 772.0-Al0 Mechanical Equipment Room 2 9.97E-04 1.53E-04 1.50E-07 43 772.0-A13 125V Vital Battery Room IV 4.27E-04 7.21E-05 3.08E-08 44 772.0-A14 125V Vital Battery Room III 4.27E-04 5.64E-05 2.41E-08 25 782.0-Al Unit 1 Control Rod Drive Equipment Room 2.95E-03 1.48E-05 4.37E-08 25 782.0-A2 Unit 1 Pressurizer Heater Transformer Rm 9.91 E-04 1.48E-05 1.47E-08 53 742.0-D3 & D9 Toilet & Corridor 1.24E-03 1.48E-05 1.84E-08 49 742.0-D4 Diesel Generator Unit 1A-A 5.93E-02 1.48E-05 8.78E-07 50 742.0-D5 Diesel Generator Unit 2A-A 5.93E-02 1.48E-05 8.78E-07 51 742.0-D6 Diesel Generator Unit 1 B-B 5.93E-02 1.48E-05 8.78E-07 52 742.0-D7 Diesel Generator Unit 2B-B 5.90E-02 1.48E-05 8.74E-07 55 742.0-CHASE A Train A Cable Tray Chase 1.04E-05 1.48E-05 1.54E-10 56 742.0-CHASE B Train B Cable Tray Chase 1.04E-05 1.48E-05 1.54E-10 49 760.5-D3 DG 1A-A Air Exhaust Room 8.86E-04 1.48E-05 1.31E-08 760.5-D4 DG 1A-A 480-V Board Room 1.36E-03 1.48E-05 2.01E-08 760.5-D5 DG lA-A Air Intake Room 7.92E-05 1.48E-05 1.17E-09 50 760.5-D6 DG 2A-A Air Exhaust Room 7.06E-04 1.48E-05 1.05E-08 760.5-D7 DG 2A-A 480-V Board Room 1.36E-03 1.48E-05 2.01 E-08 760.5-D8 DG 2A-A Air IntakeRoom 2.93E-04 1.48E-05 4.33E-09 51 760.5-D9 DG 1 B-B Air Exhaust Room 7.06E-04 1.48E-05 1.05E-08 760.5-D10 DG 1 B-B 480-V Board Room 1.36E-03 1.48E-05 2.01 E-08 760.5-Dl1 DG 1B-B Air Intake Room 7.92E-05 1.48E-05 1.17E-09 52 760.5-D12 DG 2B-B Air Exhaust Room 7.06E-04 1.48E-05 1.05E-08 760.5-D13 DG 2B-B 480-V Board Room 1.29E-03 1.48E-05 1.91E-08 760.5-D14 DG 2B-B Air Intake Room 7.92E-05 1.48E-05 1.17E-09 58 722.IPS-STR Train A ERCW Strainer Room 1.11E-03 1.28E-04 1.42E-07 RM A 59 722.IPS-STR Train B ERCW Strainer Room 1.04E-03 1.28E-04 1.33E-07 RM B 58 741.IPS-ERCW Train A ERCW Pump Room 1.49E-03 1.28E-04 1.91E-07 PMP RM A 59 741.IPS-ERCW Train B ERCW Pump Room 1.49E-03 1.66E-04 2.48E-07 PMP RM B 58 741.IPS-SWP Train A Screen Wash Pumps & HPFP Pump 3.83E-03 1.28E-04 4.90E-07 RM Room 59 741.IPS-HPFP Train B Screen Wash Pumps & HPFP Pump 3.61E-03 1.66E-04 6.OOE-07 RM B Room __I WBN2 IPEEE DESIGN REPORT Page 231 ATTIACHMENT 3

27 5. REVISED PRA MODEL EVALUATION OF FIRES (FIVE Phase II, Step 3)For the areas that remained for further evaluation following the initial quantitative evaluation, a revised evaluation was then performed.

For those areas that remained for further evaluation, the use of EPRI fire events database (Reference

1) allowed the development of fire severity factors to be incorporated into an "event tree" methodology of segmenting the fire ignition frequency into various cases (i.e. successful automatic suppression, etc.). The EPRI database has been reviewed and was incorporated into J.R. Houghton's work on fire severity for the Reliability and Risk Assessment Branch of the U.S. NRC (Reference 4). An evaluation of the severity factors developed by EPRI for specific types of ignition sources, which are described in Appendix D of the Fire PRA Implementation Guide (Reference 3), was also performed to confirm the validity of these values. This level of evaluation also involved the review of ignition sources located in various areas. This review is described in report WBN-IPE-004 U2. The 45 rooms that required further evaluation during this step represent all remaining areas to be evaluated.

The results for these areas are summarized in Table 5.2.The areas remaining for further evaluation in Step 3 of Phase 2 were evaluated using a modified version of the WBN plant model. This model was develope~d for the initial quantitative evaluation described in Section 4. The development and modification of this plant risk model are discussed in References 5 and 13. All area fires were therefore modeled as being no less severe than a plant trip due to total loss of main feedwater.

The fire-related impacts for each area also included failure of all equipment located in, or routed through, the affected area.The 45 remaining plant rooms were evaluated based on the presence of significant fire ignition sources in the area and information from the fire events database to determine the likelihoodof a fire spreading to become a significant event.In order to simplify the development of individual cases for this evaluation, an "event tree" structure was used, with severity values and non-suppression probabilities, to lay out the various cases. For example, this type of evaluation was used to evaluate Auxiliary Building Corridor Area 713.0-Al (713.0-Al

& A22). Four cases were evaluated for this room: Case 1 A minor fire starts in the affected area and self-extinguishes, or is extinguished with portable extinguishers.

This results in a plant trip with a total loss of main feedwater and no other fire impacts.Case 2 A potentially significant fire starts and is suppressed by actuation of the installed sprinkler system. All non-Appendix R equipment, with the exception of SI/RHR components to support bleed and feed cooling, are assumed to fail prior to fire suppression.

This results in a plant trip with a total loss of main feedwater.

WBN2 IPEEE DESIGN REPORT Page 232 ATTACHMENT 3

28 Case 3 A potentially significant fire starts in the affected area and is eventually suppressed with hose streams. This fire is then conservatively assumed to damage all non-Appendix R equipment, including SI/RHR equipment that supports bleed and feed cooling.Plant trip is again assumed to occur due to a fire-induced total loss of main feedwater.

Case 4 A potentially significant fire starts in the affected area and is not suppressed.

This fire is then assumed to fail (1) all plant equipment that is not specifically credited in the Appendix R evaluation, except SI/RHR equipment and (2) equipment located in or routed through the affected area. As shown in Table 7 of WBN-IPE-003 U2 for the analysis volume (AV26) found to be bounding for this room, this includes both motor driven AFW pumps, component cooling water pump 2A-A, both RHR pumps, and centrifugal charging pump 2B-B.The development of the frequency for each of these cases can be shown graphically as: Figure 5-1 Example Case Development Using Event Tree Structure I.F8 eve SM resso B9Ad5E0 _ .4-nce F 1.E ICaseP 1 0.82 Minor 9.50E-03 1.54E-06 1.46E-08 (Case 1)1. 16E-02 (Minor)0.18 0.95 Damage Minor 1.98E-03 1.48E-05 2.94E-08 (Case 2)Damage (Severe)(Yes)0.05 0.9 Multiple 9.39E-05 7.62E-05 7.16E-09 (Case 3)(No) (Yes) Damage 0.1 Extensive 1.04E-05 8.40E-02 8.77E-07 (Case 4)(No) Damage Total Fire-Induced CDF = 9.28E-07 NOTE: 1. Brigade is synonymous with Fire Department.

2. Fire ignition frequency based on fire severity and mitigating factors.The total fire-related core damage frequency for all of these cases is less than 1E-06;therefore, this area can be screened from further evaluation.

The discussion for each of these areas includes justification for the evaluation remaining conservative.

For all areas evaluated with this technique, only 7 were within a factor of two of the screening criteria of 1 E-06 (i.e. a fire-related core damage frequency above 5E-07).Only four areas were within 30% of the screening criteria: WBN2 IPEEE DESIGN REPORT Page 233 ATTACHMENT 3

29 TABLE 5.1 AREAS WITHIN 30% of SCREENING CRITERIA Fire Related Core Room Name Room Number Da ae Re quenC y Damage Frequency Main Control Room 757-C12 9.65E-7 Corridor 713-Al & 713-A22 9.28E-7 125-V Vital Battery Board 757-A22 8.35E-7 Room IV Refueling Room 757-A13 7.46E-7 The areas listed in Table 5.2 were screened from further evaluation based on this level of review.Note: Room 737.0-Al is evaluated as three areas 737.0-AlA, 737.0-A1B and 737.0-A1C in Table 5.2, but is shown with only one entry in Table 6.1 (the highest CDF which was in 737.0-Al B). The potential for fire spread across the two 21 foot buffer areas (737.0-AlAN and 737.0-Al BN) between 737.0-AlA and A1B and the 27 foot buffer (737.0-AlCN) to 737.0-A1C to involve all of room 737.0-Al was previously evaluated by NRC (Reference 19).The potential for cross zone fire spread for this area is discussed under NUREG-1407 Issue Number 4 in Section 7.3.WBN2 IPEEE DESIGN REPORT Page 234 ATTIACHMENT 3

30 TABLE 5.2 AREAS SCREENED IN THE REVISED EVALUATION (INFORMATION FROM WBN-IPE-005 U2)FIRE ROOM ROOM I AREA NO. DESCRIPTIONJ F 2 8 713.0-Al & A22 Corridor &Valve Gallery 9.28E-07 713.0-A17 Seal Water Heat Exchanger 2A 4.57E-09 10 757.0-A13 Refueling Room 7.46E-07 772.0-A9 HEPA Filter Plenum Room 3.641E-08 14 737.0-Al Corridor 5.05E-07 17 757.0-A2 6.9KV and 480-V Shutdown Board Room A 4.27E-08 757.0-A9 Personnel and Equipment Access 7.94E-08 19 757.0-A4 125-V Vital Battery Board I Room 7.88E-08 20 757.0-Al Auxiliary Control Room 1.95E-08 23 757.0-A27 Auxiliary Control Instrument Room 2A 1.91 E-08 24 757.0-A28 Auxiliary Control Instrument Room 2B 2.86E-08 27 757.0-A5 480-V Shutdown Board Room 1 B 2.35E-08 28 757.0-A21 480-V Shutdown Board Room 2A 4.03E-08 29 757.0-A22 125-V Vital Battery Board Room IV 8.35E-07 30 757.0-A23 125-V Vital Battery Board Room III 5.96E-08 31 757.0-A17 Personnel and Equipment Access 7.54E-08 757.0-A24 6.9KV & 480-V Shutdown Board Room B 2.27E-07 32 772.0-Al 480-V Board Room 1A 2.84E-08 41 772.0-Al1 480-V Transformer Room 2B 1.15E-07 42 772.0-Al 2 480-V Transformer Room 2A 1.25E-07 45 772.0-Al 5 480-V Board Room 2B 8.09E-08 46 772.0-Al 6 480-V Board Room 2A 2.88E-08 47 786.0-AR Auxiliary Building Roof 3.10E-07 48 692.0-C5 250-V Battery Board Room 2 1.87E-07 692.0-C6 250-V Battery Room 2 3.38E-08 692.0-C7 24-V & 48-V Battery Room 3.166E-08 692.0-C8 24-V & 48-V Battery Bd and Charger Room 2.34E-07 692.0-C9 Communications Room 1.90E-07 692.0-C10 Mechanical Equipment Room 1.001E-07 692.0-Cl 1 Corridor 5.30E-08 692.0-C12 Secondary Alarm Station 1.051E-07 708.0-C2 Corridor 4.45E-08 708.0-C3 Plant Computer Room. 2.56E-07 708.0-C4 Auxiliary Instrument Room 2 6.83E-07 729.0-Cl Cable Spreading Room 1.42E-07 755.0-C12 Main Control Room 9.65E-07 755.0-Cl 3& C20 Relay Room & DPSO Shop 2.20E-07 60 711.1PS-Elec BD 480-V Board Room 3.28E-08 Room 63 Turbine Building Turbine Building 5.92E-07 65 713.0-A29 Unit 2 Pipe Chase & Valve Gallery 4.09E-08 WBN2 IPEEE DESIGN REPORT Page 235 AI-IACHMENT 3

31 AREAS SCREENED IN THE REVISED EVALUATION (INFORMATION FROM WBN-IPE-005 U2)FIRE ROOM ROOM AREA NO. DESCRIPTION F 2 70 692.0-A25 Unit 2 Pipe Gallery 7.69E-08 71 713.0-A19 Unit 2 Pipe Gallery 3.23E-08 75 757.0-A16 Emergency Gas Treatment Filter Room 7.57E-08 782.0-A3 Unit 2 Control Rod Drive Equipment Room 5.42E-08 782.0-A4 Unit 2 Pressurizer Heater Transformer Room 1.87E-08 6. RESULTS (FIVE Phase III, Step 1)Phase III of the FIVE process consists of documentation of results from the screening process and resolution of any new or remaining issues, primarily those identified in the Sandia Fire Risk Scoping Study (NUREG/CR-5088, Reference 6). The results of this evaluation are discussed in this section. New and remaining issues are then discussed in Section 7.This evaluation, which was performed in accordance with the requirements of Supplement 4 to Generic Letter 88-20 (NUREG-1407, Reference

19) and the guidance provided by the.EPRI FIVE documentation (Reference 2), has confirmed that there are no fire-induced vulnerabilities of concern associated with the continued operation of the Watts Bar Nuclear Plant.The results of the screening process performed in this evaluation can be summarized as follows: Of the 140 plant areas listed in Table 6.1, 17 were screened from further consideration in Phase I, the fire compartment interaction analysis.

These screened areas are indicated with an "X" under the "Phase I" column in Table 6.1 and are listed individually in Table 2.1.Of the (140-17 =) 123 areas listed in Table 6.1 that remained for quantitative evaluation, only those associated with the Unit 2 Reactor Building were screened from further consideration due to fire ignition frequency below 1 E-06. This is indicated with "11.1" in the "Screened at Phase 11.2" column in Table 6.1.Of the (121-5) = 116 areas (five rooms were combined with five other rooms) listed in Table 6.1 that remained for quantitative evaluation, 71 were screened from further consideration based on the initial quantitative evaluation.

These areas are indicated with an "X" in the "Phase 11.2" column in Table 6.1 and are individually listed in Table 4.1.Of the (116 -71 =) 45 areas that remained for detailed quantitative evaluation, all were screened from further consideration in Phase II, Step 3. These areas are WBN2 IPEEE DESIGN REPORT Page 236 ATTACHMENT 3

32 indicated with an "X" in the "Phase 11.3" column in Table 6.1 and are individually listed in Table 5.2.TABLE 6.1

SUMMARY

OF AREA SCREENING PROCESS FIRE ROOM ROOM SCREENED AT PHASE AREA NO. DESCRIPTION I 11.2 11.3 1 676.0-Al Corridor X 676.0-A14 Containment Spray Pump Room 2A-A X 676.0-A15 Containment Spray Pump Room 2B-B X 692.0-Al & A4 Corridor & Chemical Drain Tank Room X 692.0-A5 Gas Decay Tank Room X 692.0-A27 Concentrate Filter Room X 692.0-A29 Boric Acid Evaporator Package B Room X 692.0-A30 Boric Acid Evaporator Package A Room X 1-1 692.0-A20 Safety Injection Pump Room 2B-B X 1-2 692.0-A19 Safety Injection Pump Room 2A-A X 2-2 676.0-Al 3 RHR Pump Room 2B-B X 3-2 676.0-A12 RHR Pump Room 2A-A X 8- 713.0-Al & A22 Corridor & Valve Gallery X 713.0-A4 Radio Chemistry Lab X 713.0-A14 Sample Room 2B X 713.0-A15 Heat Exchanger Room 2A X 713.0-A16 Heat Exchanger Room 2B X 713.0-A17 Seal Water Heat Exchanger Room 2A X 713.0-A23 CVCS Valve Gallery X 10 729.0-A6 Nitrogen Storage Area X 729.0-A9 Unit 2 Fuel Transfer Valve Room X 757.0-A13 Refueling Floor, X 772.0-A9 HEPA Filter Plenum Room X 786.0-Al Fan Room X 14 737.0-Al Corridor X 737.0-A8 Unit 2 Letdown Heat Exchanger Room X 737.0-Al1 Air Lock X 15-1 737.0-A3 HVAC Equipment Room 1 X 15-2 737.0-A12 HVAC Equipment Room 2 X 17 757.0-A2 6.9KV & 480V Shutdown Board Room A X 757.0-A9 Personnel

& Equipment Access Room X 18 757.0-A3 125V Vital Battery Board Room II X 19 757.0-A4 125V Vital Battery Board Room I X 20 757.0-Al Auxiliary Control Room X 21 757.0-A25 Unit 1 Train A Instrument Room X 22 757.0-A26 Unit 1 Train B Instrument Room X 23 757.0-A27 Unit 2 Train A Instrument Room X 24 757.0-A28 Unit 2 Train B Instriment Room X 25 782.0-Al Unit 1 Control Rod Drive Equipment Room X 25 782.0-A2 Unit 1 Pressurizer Heater Transformer Room X 27 757.0-A5 480V Shutdown Board Room 1B X WBN2 IPEEE DESIGN REPORT Page 237 ATTACHMENT 3

33

SUMMARY

OF AREA SCREENING PROCESS FIRE ROOM ROOM SCREENED AT PHASE AREA NO. DESCRIPTION I 11.2 11.3 28 757.0-A21 480V Shutdown Board Room 2A X 29 757.0-A22 125V Vital Battery Board Room IV X 30 757.0-A23 125V Vital Battery Board Room III X 31 757.0-A17 Personnel

& Equipment Access Room X 757.0-A24 6.9KV & 480V Shutdown Board Room B X 32 772.0-Al 480V Board Room 1A X 33 772.0-A2 480V Board Room 1B X 34 772.0-A3 125V Vital Battery Room II X 35 772.0-A4 125V Vital Battery Room I X 36 772.0-A5 480V Transformer Room 1 B X 37 772.0-A6 480V Transformer Room IA X 38 772.0-A7 Mechanical Equipment Room 1 X 39 772.0-A8 Fifth 125V Vital Battery & Board Room X 40 772.0-Al0 Mechanical Equipment Room 2 X 41 772.0-All 480V Transformer Room 2B X 42 772.0-A12 480V Transformer Room 2A X 43, 772.0-Al 3 125V Vital Battery Room IV X 44 772.0-A14 125V Vital Battery Room III X 45 772.0-A15 480V Board Room 2B X 46 772.0-A16 480V Board Room 2A X 47 786.0-AR Auxiliary Building Roof X 786.0-A2 Roof Access Air Lock X 786.0-A3 Mechanical Equipment Room X 786.0-A4 Mechanical Equipment Room X 48 692.0-C5 250V Battery Board Room 2 X 692.0-C6 250V Battery Room 2 X 692.0-C7 24V & 48V Battery Room X 692.0-C8 24V & 48V Battery Board & Charger Room X 692.0-C9 Communications Room X 692.0-C10 Unit 2 Mechanical Equipment Room X 692.0-Cl1 Corridor X 692.0-C12 Secondary Alarm Station X 708.0-C2 Corridor X 708.0-C3 Computer Room X 708.0-C4 Unit 2 Auxiliary Instrument Room X 729.0-Cl Cable Spreading Room X 755.0-C12 Main Control Room X 755.0-C13

& C20 Relay Room & DPSO Shop X 49 742.0-D4 Diesel Generator Unit 1A-A Room X 760.5-D3 Diesel Generator 1A-A Air Exhaust Room X 760.5-D4 Diesel Generator 1A-A Board Room X 760.5-D5 Diesel Generator 1A-A Air Intake Room X 50 742.0-D5 Diesel Generator Unit 2A-A Room X 760.5-D6 Diesel Generator 2A-A Air Exhaust Room X 760.5-D7 Diesel Generator 2A-A Board Room X WBN2 IPEEE DESIGN REPORT Page 238 ATTIACHMENT 3

34

SUMMARY

OF AREA SCREENING PROCESS FIRE ROOM ROOM SCREENED AT PHASE AREA NO. DESCRIPTION I 11.2 11.3 50 760.5-D8 Diesel Generator 2A-A Air Intake Room X 51 742.0-D6 Diesel Generator Unit 1 B-B Room X 760.5-D9 Diesel Generator I B-B Air Exhaust Room X 760.5-D10 Diesel Generator 1B-B Board Room X 760.5-DI 1 Diesel Generator 1 B-B Air Intake Room X 52 742.0-D7 Diesel Generator Unit 2B-B Room X 760.5-D12 Diesel Generator 2B-B Air Exhaust Room X 760.5-D13 Diesel Generator 2B-B Board Room X 760.5-D14 Diesel Generator 2B-B Air Intake Room X 53 742.0-D3 & D9 DG Building Toilet & Corridor X 742.0-D8 Fuel Oil Transfer Valve Room X-55 742.0-Chase A Train A Cable Chase X 56 742.0-Chase B Train B Cable Chase X 58 722.1PS-STR A Train A ERCW Strainer Room X 741 .IPS-ERCW Train A ERCW Pump Room X PUMP RM A 741.IPS- SWP A Train A Screen Wash & Fire Pump Room 59 722.IPS-STR B Train B ERCW Strainer Room X 741.IPS-ERCW Train B ERCW Pump Room X PUMP RM B 741.IPS- SWP B Train B Screen Wash & Fire Pump Room X 60 711.IPS-ELEC Intake Pumping Station Electrical Board X BD ROOM Room 728.IPS-RSW Intake Pumping Station -RSW Pump Deck X DECK 63 TURBINE BLDG Turbine Building X 65 676.0-A17 Unit 2 Pipe Gallery X 692.0-A24 Unit 2 Pipe Chase & Valve Gallery X 713.0-A29 Unit 2 Pipe Chase & Valve Gallery X 66 692.0-A21 Charging Pump 2C Room X 67 692.0-A22 Centrifugal Charging Pump 2B-B Room X 68 692.0-A23 Centrifugal Charging Pump 2A-A Room X 69 692.0-A26.

Turbine Driven AFW Pump 2A-S Room X 70 692.0-A25 Unit 2 Pipe Gallery X 71 713.0-A19 Unit 2 Pipe Gallery X 713.0-A21 Unit 2 Reactor Building Access Room X 71-1 713.0-A20 Unit 2 Volume Control Tank Room X 72 729.0-Al 1 Unit 2 South Main Steam Valve Room X 737.0-A10 Air Lock X 73 729.0-Al0 Unit 2 North Main Steam Valve Room X 729.0-A13 Unit 2 Steam Valve Instrument Room X 729.0-A15

& Unit 2 Additional Equipment Building (all X 763.5-A2 elevations) 729-5-A17 Unit 2 Shield Building Vent Rad Monitor Rm. X 737.0-A14 Air Lock X I_ I WBN2 IPEEE DESIGN REPORT Page 239 ATTACHMENT 3

35

SUMMARY

OF AREA SCREENING PROCESS FIRE ROOM ROOM SCREENED AT PHASE AREA NO. DESCRIPTION I 11.2 11.3 74 737.0-A9 Unit 2 Ventilation

& Purge Air Room X 737.0-A16 Unit 2 Gross Fail Fuel Detector Room X 75 757.0-A14 Unit.2 Reactor Building Access Room X 757.0-A16 Emergency Gas Treatment Filter Room X 782.0-A3 Unit 2 Control Rod Drive Equipment Room X 782.0-A4 Unit 2 Pressurizer Heater Transformer Room X 76 757.0-A15 Unit 2 Reactor Building Equipment Hatch X 77 U2 ANN Unit 2 Annulus 11.1 U2 Primary Unit 2 Primary Containment 11.1 WBN2 IPEEE DESIGN REPORT Page 240 ATTACHMENT 3

36 7. NEW AND REMAINING ISSUES (FIVE PHASE III, Step 2)The last part of the fire evaluation process addresses the response to and resolution of the Sandia Fire Risk Scoping Study (NUREG/CR-5088, Reference

6) issues and the evaluation of containment isolation and heat removal. Also, the individual requirements for performance and documentation of a fire IPEEE, as specified in NUREG-1407 (Reference 19), are addressed.

These issues are discussed in the following report sections: Section 7.1 Containment heat removal and isolation Section 7.2 Sandia Fire Risk Scoping Study (NUREG/CR-5088) issues Section 7.3 Individual requirements of NUREG-1407.

References to the Fire Protection Report and plant documentation (e.g., procedures) in this section will be reviewed and updated during the validation of the Unit 2 "as designed" FIVE analysis.7.1 Evaluation of Containment Heat Removal and Isolation Section 4.1.5 (Perform Containment Analysis) of NUREG-1407 states that the licensee should: "Perform containment analysis if containment failure modes differ significantly from those found in the IPE internal events evaluation.

The EPRI FIVE report (Reference

2) provides a justification for limiting containment performance evaluations related to heat removal and isolation.

It states "...fires leading to the potential loss of safe shutdown functions above the threshold value of 1 E-06/year and having plant damage states and minimum operable equipment not included in the IPE, should be flagged for containment analysis evaluation." In the Unit 2 fire analysis, all of the areas screened-out and had fire-induced core damage frequencies less that 1 E-06/year; therefore, further containment evaluation was unnecessary.

The containment discussion below pertaining to the Fire Protection Report and the Appendix R analysis is for information.

As stated previously, the EPRI FIVE report provides justification for not evaluating containment isolation and heat removal if the fire-related core damage frequency for all areas is below 1 E-06/year.

A separate discussion of the requirements for containment isolation in response to fires is provided in Section 4.12 of Part III of the Fire Protection Report. Review of these considerations, relative to containment failure confirm that no containment failure modes are introduced by the fire evaluation that differ significantly from those seen in the IPE.In addition to containment isolation, the Fire Protection Report (FPR) addresses several issues related to reactor coolant system integrity following plant fires. The first of these issues, reactor coolant pump (RCP) seal integrity, is addressed in Part Ill, Section 4.1 of the FPR.The Appendix R analysis also considered the pressurizer PORV and PORV block valves as Fire Safe Shutdown (FSSD) components, such that they were explicitly tracked through the fire evaluation process. Due to the requirement for a "hot short" condition to open the PORV, WBN2 IPEEE DESIGN REPORT Page 241 ATTACHMENT 3

37 coincident with a failure of the associated block valve to close on demand from the operator, the routing of these components as redundant fire safe shutdown components provides adequate assurance against a fire-induced LOCA through this path. For the single plant area where this mode of failure was noted (Main Control Room Panel 2-M-5, as described in Section 6.5 of report WBN-IPE-005 U2), no other major plant components were identified as being affected by a fire. For a significant, unsuppressed fire in this area, Control Room evacuation would be required, allowing isolation of these components from outside the Control Room using the Alternate Shutdown Capability described in Section IV of the Fire Protection Report and Abnormal Operating Instruction, AOI-30.2.Finally, the Fire Protection Report addresses containment heat removal by including the associated heat removal path (using Emergency Raw Cooling Water -ERCW -as an ultimate heat sink) with the long term decay heat removal path using the RHR system. This heat removal path would also be used for long term containment heat removal following plant shutdown with bleed and feed cooling. This mode of cooling is only credited for the evaluation of specific cases in reports WBN-IPE-003 U2 and WBN-IPE-005 U2. Otherwise, the Fire Protection Report requires short term heat removal through the steam generators using Auxiliary Feedwater (AFW).7.2 Resolution of Sandia Fire Risk Scoping Study Issues The EPRI FIVE documentation discusses the following six issues to be addressed.

1. Seismic/fire interactions.
2. Fire barrier qualification.
3. Manual fire fighting effectiveness.
4. Total environment equipment survival.5. Control systems interaction.
6. Improved analytical codes.These issues, which were originally taken from the Fire Risk Scoping Study (NUREG/CR-5088) performed by Sandia Laboratories (the Sandia Fire Risk Scoping Study Issues) are discussed below. The specific responses for each of these concerns for the Watts Bar analysis are listed in italics directly below the description of the Sandia issue.WBN2 IPEEE DESIGN REPORT Page-242 ATTACHMENT 3

38 7.2.1 Seismic/Fire Interactions The issue of seismic/fire interactions centers on the following 3 areas of interest:* Seismically induced fires. In particular, this concern centers on fires caused by flammable gas or liquid storage containers or systems that could rupture during a seismic event." Seismic actuation of fire suppression systems. In particular, this concern centers on the failure of electrical or other components due to water sprays." Seismic degradation of fire suppression systems. In particular, this concern reviews the plant design for fragility of fire suppression systems to a seismic event.Each of these areas of interest is described in detail below.7.2.1.1 Seismically Induced Fires As part of the seismic assessment walkdown, verify hydrogen or other flammable gas or liquid storage vessels in areas with seismic safe shutdown or safety related equipment are not subject to leakage under seismic conditions.

Examples would be improperly anchored hydrogen or oxygen bottles, hydrogen tanks used for primary coolant chemistry control, etc.Response Hydrogen or flammable gas/liquid storage vessels are not kept on a permanent basis in the Auxiliary Building, Diesel Generator Buildings, Control Building or the Intake Pump Station. Hydrogen storage tanks are located outside in the yard area. Hydrogen supply lines are seismically designed and provided with excess flow shutoffs.Site Standard Practice SSP-13.02 and Fire Protection Instruction FPI-0100 provide the requirements for control of this type of combustible, including the requirement that compressed gas cylinders be tied to permanent structural features, using methods described in the standard practice.In addition, the seismic walkdown required for the seismic portion of the Unit I IPEEE identified potential seismic class II components affecting seismic class I components in safety related areas (Reference 16). These were shown to be acceptable by the IPEEE Outlier Interaction Evaluation (Reference 17). For IPEEE Unit 2, the seismic walkdowns will be performed as a validation activity.WBN2 IPEEE DESIGN REPORT Page 243 ATTACHMENT 3

39 7.2.1.2 Seismic Actuation of Fire Suppression Systems As part of the seismic assessment, verify that the design of the water suppression system considers the effects, if appropriate, of inadvertent suppression system actuation and discharge on that equipment credited as part of the seismic safe shutdown path in a margins assessment that was not previously reviewed relative to the internal flooding analysis or concerns such as those discussed in NRC I&E Notice 83-41.Response This issue was also addressed by Information Notice 94-12, Effects of Fire Suppression System Actuation on Safety Related Systems.The Watts Bar response to these issues was as follows: 1. Mercury Relays. No mercury relays are present in the fire protection control systems.2. Seismic Dust/Smoke Detectors.

Smoke and/or heat detectors are used at the Watts Bar Nuclear Plant to actuate fire suppression systems in various areas of the plant. The C0 2 systems are actuated by heat detectors or by a combination of smoke and heat detectors.

Therefore, dust particles created during a seismic event alone will not activate the C0 2 systems.Watts Bar Nuclear Plant has no open head spray systems located inside safety related buildings.

As part of the Appendix R analysis, fire suppression damage evaluations have been made, as documented in calculation EPM-RAC-032392, Evaluation of Suppression System Discharge (Reference 15). It has been concluded that spurious discharge of water from fire suppression systems will have no adverse impact on the safe shutdown capability of the plant.3. Water Deluge Systems. As noted above, Watts Bar Nuclear Plant has no open head spray systems located inside safety related buildings.

4. Fire Suppressant Availability during a Seismic Event. Halon systems are not used to protect areas that contain safety related equipment.

The C0 2 systems are seismically qualified, with the exception of the refrigeration system, which is not required except for prolonged periods. The water suppression system uses four electric motor driven pumps and one diesel driven fire pump. The electric pumps and associated 6.9kV shutdown boards are located in seismic class I structures.

WBN2 IPEEE DESIGN REPORT Page 244 ATTACHMENT 3

40 5. Switchgear Fires. There are cases where electrical cables and raceways are located close to the top of electrical cabinets and could become directly involved in a fire. These cases are identified in report WBN-IPE-004 U2 and evaluated in the detailed analysis described in report WBN-IPE-005 U2, which is summarized in Section 5 of this report.6. Electro-Mechanical Components in Cable Spreading Rooms. No high voltage electric cabinets are present in these areas at the Watts Bar Nuclear Plant. HVAC equipment and control panels in these areas are installed such that tipping or sliding is prevented.

7.2.1.3 Seismic Degradation of Fire Suppression Systems As part of the seismic assessment walkdown, verify that plant fire suppression systems have been structurally installed in accordance with good industrial practice and reviewed for seismic considerations, such that suppression system piping and components will not fail and damage safe shutdown path components, nor is it likely that leaking or cascading of the suppressant will result.Response The fire protection system piping is designed to maintain pressure boundary integrity where spray damage to safety related components would affect the safe shutdown capability of the plant.The fire protection system piping is designed at a minimum for position retention (seismic Il/I design criteria).

Additionally, the seismic walkdown required for the seismic portion of the Unit 1 IPEEE identified potential seismic class II components affecting seismic class I components in safety related areas (Reference 16).These were shown to be acceptable by the IPEEE Outlier Interaction Evaluation (Reference 17). For the Unit 2 IPEEE, detailed seismic verification walkdowns of the fire protection piping will be performed as part of the Integrated Interaction Program (liP) and this will be used as the basis for screening out this concern during seismic IPEEE validation activities.

7.2.2 Fire Barrier Qualifications The concern for fire barrier qualification centers on the following 4 areas of interest: " Fire barrier surveillance program.* Inspection and maintenance of fire doors.* Installation, inspection, surveillance and maintenance of penetration seal assemblies.

  • Inspection, testing and maintenance of fire dampers.Each of these areas of interest is described in detail below.WBN2 IPEEE DESIGN REPORT Page 245 ATTACHMENT 3

41 7.2.2.1 Fire Barriers Fire barriers and components such as fire dampers, fire penetration seals and fire doors for fire barriers are included in the plant surveillance program.Response Fire barriers are included in the Watts Bar Nuclear Plant surveillance program. Fire Operating Instruction 0-FOR-304-1, Fire Barrier/Mechanical, Conduit, Cable Tray, and Fire Damper (External)

Penetration Visual Inspection

-Auxiliary, Control and Diesel Generator Building, is performed to verify the functional status of required fire rated barriers, including mechanical pipe fire rated penetration seals and external electrical conduit fire rated seals by performing a visual inspection.

0-FOR-304-2, Electrical Raceway Fire Barrier Systems Visual Inspection

-Auxiliary Building addresses these requirements for electrical raceways in the Auxiliary Building.The Reactor Buildings are addressed by 1-FOR-304-1, Visual Inspection of Fire-Rated Assemblies Located in the Reactor Building (2-FOR-304-1 is being written for Unit 2 Reactor Building).

7.2.2.2 Fire Doors A fire door inspection and maintenance program should be implemented at the plant.Response The inspection of fire doors is addressed by Fire Operating Instruction 0-FOR-410-1, 31 Day Fire Door Inspection.

Operational testing of fire doors is addressed by 0-FOR-410-2, 12 Month Fire Door Operational Test.7.2.2.3 Penetration Seal Assemblies

a. A penetration seal inspection and surveillance program should be implemented at the plant.Response The surveillance and inspection of penetration seals is addressed in Fire Operating Instruction 0-FOR-304-1, Fire Barrier IMechanical, Conduit, Cable Tray, and Fire Damper (External)

Penetration Visual Inspection

-Auxiliary, Control and Diesel Generator Building.b. Fire barrier penetration seals have been installed and maintained to address concerns such as those identified in NRC Information Notice 88-04.Response Fire barrier penetration seals at the Watts Bar Nuclear Plant have been installed and are maintained in compliance with the relevant Appendix R requirements, as described in Part II, Section 12.10.6 WBN2 IPEEE DESIGN REPORT Page 246 ATTIACHMENT 3

42 (Penetration Seals) of the Watts Bar Nuclear Plant Fire Protection Report.7.2.2.4 Fire Dampers a. An inspection and maintenance program for fire dampers should be implemented at the plant.Response The inspection and testing of fire dampers is addressed by Fire Operating Instructions 0-FOR-304-1, Fire Barrier/Mechanical, Conduit, Cable Tray, and Fire Damper (External)

Penetration Visual Inspection

-Auxiliary, Control and Diesel Generator Building and 0-FOR-304-3, Fire Damper (Internal)

Visual Inspection

-Auxiliary, Control and Diesel Generator Buildings.

b. Damper installations address concerns such as those identified in NRC Information Notice 89-52, "Potential Fire Damper Operational Problems," dated June 8, 1989 and NRC Information Notice 83-69, "Improperly Installed Fire Dampers at Nuclear Power Plants," dated October 21, 1983.Response Fire dampers at the Watts Bar Nuclear Plant are installed to meet the applicable Appendix R compartmentation requirements.

These dampers are described in Part II, Section 12.10.5 (Fire Dampers) of the Watts Bar Nuclear Plant Fire Protection Report. Inspection and testing for this equipment is discussed under item (a) above.7.2.3 Manual Fire Fi-qhtinq Effectiveness The concern for manual fire fighting effectiveness centers on the following 6 areas of interest: " Fire reporting, including the use and availability of portable fire extinguishers and plant procedures for reporting fires, including plant communication." Fire brigade makeup and equipment.

  • Fire brigade training in the classroom." Fire brigade practice in hands-on structural fire training and in the use of equipment.
  • Fire brigade drills." Fire brigade training records.Each of these areas of interest is described in detail below.7.2.3.1 Reportinq Fires WBN2 IPEEE DESIGN REPORT Page 247 ATTACHMENT 3

43 a. Appropriate plant personnel are knowledgeable in the use of portable fire extinguishers.

Response Plant personnel and fire department (WBN has a dedicated fire department instead of a fire brigade) members receive regular training in the use of portable fire extinguishers.

b. Portable extinguishers are located throughout the plant.Response Part II, Section 12.4.1 (Portable Extinguishers) of the Fire Protection Report specifies that "Portable extinguishers of a size and type compatible with specific hazards are located throughout the plant." c. A plant procedure is in use for reporting fires in the plant.Response Fire Protection Instruction FPI-01 10 (Emergency Response) directs the person reporting the fire to call extension 3911, which will connect the caller to the Main Control Room.d. A plant communication system that includes contact to the control room is operable at the plant.Response The Control Room can be contacted by telephone, from one of the intemal plant communication stations or by plant operations/fire brigade radio.7.2.3.2 Fire Brigade Makeup and Equipment 7.2.3.,2.1 A fire brigade that is made up of at least 5 trained people on each shift should be maintained at the plant.Response This requirement (1 brigade leader and at least 4 other members) is specified in Section 9.1 (Fire Brigade Staffing) of Part II of the Fire Protection Report.7.2.3.2.2" The fire brigade leader and at least two other brigade members on each should be knowledgeable in plant systems and operations.

brigade shift Response All fire department personnel pass the two week Nuclear Systems Training Course, in addition to completing a Plant Systems Familiarization Qualification Card. This requirement is also specified for personnel transferring from other nuclear sites in Fire Protection Instruction FPI-0120 (Emergency Response Training).

7.2.3.2.3 Each brigade member should receive an annual review of physical evaluate his ability to perform fire fighting activities.

condition to WBN2 IPEEE DESIGN REPORT Page 248 ATTACHMENT 3

44 Response This requirement is specified in Section 9.3 (Training and Qualifications) of Part II of the Watts Bar Nuclear Plant Fire Protection Report.7.2.3.2.4 A minimum amount of equipment should be provided for the onsite fire brigade: a. Personal protective equipment should be provided such as SCBA, turnout coats, boots, gloves, and-hard hats.b. Emergency communications equipment should be provided for fire brigade use.c. Portable lights should be provided for fire brigade use.d. Portable ventilation equipment should be provided for fire brigade use.e. Portable extinguishers should be provided for fire brigade use.Response These requirements are included in Section 9.4 (Firefighting Equipment), Part II of the Fire Protection Report.7.2.3.3 Fire Brigade Traininq Brigade members should receive an initial classroom instruction program consisting of the following:

a. A review of the plant fire fighting plan and identification of each individual's responsibilities.
b. Identification of typical fire hazards and associated types of fires that may occur in the plant.c. Identification of the location of firefighting equipment and familiarization with the layout of the plant, including access and egress routes.d. Training on the proper use of available firefighting equipment and the correct method of fighting each type of fire. The types of fires covered should include fires in energized electrical equipment, fires in cables and cable trays and fires involving flammable and combustible liquids and gases.e. Training on the proper use of communication, lighting, ventilation and emergency breathing equipment.
f. Training on techniques for fighting fires inside buildings and confined spaces.g. A review of fire fighting strategies and procedures.

Response Fire Department training requirements, including those that address the items listed above are specified in Section 9.3 (Training and WBN2 IPEEE DESIGN REPORT Page 249 ATTACHMENT 3

45 Qualification) of Part II of the Watts Bar Nuclear Plant Fire Protection Report.7.2.3.4 Fire Brigade Practice Fire brigade members should receive hands-on structural fire fighting training at least once a year to provide experience in actual fire extinguishment and the use of emergency breathing apparatus.

Response Fire Department drill requirements, including annual requirements, such as actual fire extinguishment and the use of emergency breathing apparatus, are specified in Section 9.3 (Training and Qualifications) of Part II of the Watts Bar Nuclear Plant Fire Protection Report.7.2.3.5 Fire Brigade Drills a. Fire brigade drills are performed in the plant so that each fire brigade shift can practice as a team.Response Fire department drill requirements for fire department team members are specified in Fire Protection Instruction FPI-0120 (Emergency Response Training).

b. Drills should be performed at regular intervals for each shift fire brigade.Response Section 9.3 (Training and Qualifications) of Part II of the Fire Protection Report specifies a minimum of one drill per shift every 92 days.c. At least one unannounced fire drill for each shift fire brigade should be performed per year.Response This requirement is specified in Section 9.3 (Training and Qualifications) of Part II of the Watts Bar Nuclear Plant Fire Protection Report).d. At least one drill per year should be performed on a "backshift" for each shift fire brigade.Response This requirement is specified in Section 9.3 (Training and Qualifications) of Part II of the Watts Bar Nuclear Plant Fire Protection Report).e. Drills should be preplanned to establish training objectives and critiqued to determine how well the training objectives have been met.Response These requirements, including scheduling of additional training for identified deficiencies, are specified in Section 9.3 (Training and WBN2 IPEEE DESIGN REPORT Page 250 ATTACHMENT 3

46 Qualifications) of Part II of the Watts Bar Nuclear Plant Fire Protection Report. The forms used for fire drill critique are provided in Fire Protection Instruction FPI-0120 (Emergency Response Training).

f. At least triennially, an unannounced drill should be performed for and critiqued by qualified individuals, independent of the licensee's staff.Response This requirement is specified in Section 9.3 (Training and Qualifications) of Part II of the Watts Bar Nuclear Plant Fire Protection Report.g. Pre-fire plans should be developed for safety related areas of the plant (as a minimum).Response This requirement for pre-fire plans is specified in Section 9.5 (Fire Emergency Procedures and Prefire Plans) of Part II of the Watts Bar Nuclear Plant Fire Protection Report.h. The pre-fire plans should be updated and used as part of the brigade training.Response The preparation and revision of prefire plans is addressed in Fire Protection Instruction FPI-0130 (Control of Prefire Plans). Prefire plan use and adequacy for drills is addressed by the drill critique form in FPI-0120 (Emergency Response Training).
i. Fire brigade equipment is maintained in good condition and ready for use brigade.by the fire Response Section 9.4 (Firefighting Equipment) of Part II of the Fire Protection Report addresses the staging of equipment to facilitate availability and address surveillance test concerns relative to life safety and ALARA. Also, equipment operability is verified prior to storage after each drill.7.2.3.6 Fire Brigade Traininq Records Records are provided for each fire brigade member, demonstrating the minimum level of training and refresher training has been provided.Response Fire department training records are maintained in order to address and document compliance with the requirements of Section 9.3 (Training and Qualification) of Part II of the Watts Bar Nuclear Plant Fire Protection Report. These requirements address initial, as well as refresher, training.7.2.4 Total Environment Equipment Survival WBN2 IPEEE DESIGN REPORT Page 251 ATTIACHMENT 3

47 The general issue of total environmental equipment survival centers on the following 3 areas of interest: " Adverse effects of combustion products on plant equipment." Spurious or inadvertent fire suppression system actuation." Impact on effectiveness of operator actions.Each of these areas of interest is discussed in detail below.7.2.4.1 Potential Adverse Effects on Plant Equipment by Combustion Products a. The FIVE methodology does not currently provide for an evaluation of non-thermal environmental effects of smoke on equipment.

See Section 4.2.2 of EPRI TR-1 00370, Fire-Induced Vulnerability Evaluation (FIVE).Response During the screening evaluation, all equipment in the affected area was assumed to be damaged by the fire. More specific plant model impacts were modeled during the detailed analysis.

This treatment is judged to conservatively bound the impact of non-thermal environmental effects on plant equipment.

Also, these non-thermal effects, such as corrosion or degradation due to soot or other smoke.products occur over a much longer period than that required to establish cold shutdown conditions.

These impacts on plant equipment, such as control circuitry and switchgear, would be addressed during the ensuing plant outage period, as part of corrective maintenance following the fire.b. Plant staff should be aware of and sensitive to the potential impact of smoke and products of combustion on human performance in safe shutdown operations in application of FIVE.Response Plant operations personnel receive regular training in the effective use of SCBA equipment.

Also, operator actions were considered to fail for fires in a given area within the plant model by failing the associated plant equipment.

7.2.4.2 Spurious or Inadvertent Fire Suppression Activation Verify that the design of fire suppression systems considers the effects, if appropriate, of inadvertent suppression system actuation and discharge on equipment credited for safe shutdown for concerns such as those discussed in NRC I&E Information Notice 83-41.Response This issue was also addressed by Information Notice 94-12, Effects of Fire Suppression System Actuation on Safety Related Systems.The Watts Bar Nuclear Plant response to these issues is discussed under Section 7.2.1.2, above.7.2.4.3 Operator Action Effectiveness WBN2 IPEEE DESIGN REPORT Page 252 ATTACHMENT 3

48 a. There are safe shutdown procedures that identify the steps for planned shutdown when necessary, in the event of a fire.Response Safe shutdown instructions have been developed to address the fires that could develop in each area of the plant. For severe fires, operator actions are directed by the appropriate section of Abnormal Operating Instruction A01-30.2, Fire Safe Shutdown.

These procedures provide detailed instructions to direct the control room operator's response to the potential loss of equipment and support cables located in each area of the plant.b. Operators should receive training on the safe shutdown procedures.

Response Discussions with plant operators have confirmed that they regularly receive training in the use of the fire safe shutdown instructions.

c. If, in performance of these procedures, operators are expected to pass through or perform manual actions in areas that may contain fire or smoke suitable SCBA equipment and other protective equipment are available for operators to perform their function.Response SCBA equipment is located in key locations throughout the plant, in addition to the equipment that is located in the fire brigade lockers.Plant operators receive regularly scheduled training in the effective use of this equipment.

7.2.5 Control

Systems Interactions This issue centers on the concern that safe shutdown circuits are physically independent of, or can be isolated from, the control room for a fire in the control room fire area.Response The remote shutdown system at Watts Bar consists of the Auxiliary Control Room and shutdown boards that are located in the Auxiliary Building.

The remote shutdown system circuits are physically independent of, or can be electrically isolated from, the Main Control Room. Therefore, safe shutdown can be accomplished from outside the Control Building in the event of a severe fire in the Control Building that would cause Main Control Room abandonment.

This capability is described in Part IV of the Fire Protection Report. The implementation of this capability is directed by Appendix C.69 of Abnormal Operating Instruction 30.2, Fire Safe Shutdown.7.2.6 Improved Analytical Codes WBN2 IPEEE DESIGN REPORT Page 253 ATTACHMENT 3

49 The issue of analytical codes centers on the fire modeling techniques that have been incorporated into the FIVE methodology.

These modeling techniques, which are derived from the basic correlations used in the COMPBRN Ille fire modeling program, have been reviewed for use in the modeling of fire progression.

Response The correlations shown in the FIVE documentation were used to generate the zones of influence shown in Report WBN-IPE-004 U2.These correlations are based on fire modeling techniques from those reviewed in the Sandia study.7.3 Requirements of NUREG-1407 The analysis described in this report was performed in order to meet the requirements of NUREG-1407.

In particular, NUREG-1407 specifies the documentation for the following areas of interest (Appendix C, Section C.3): informational submittal of 1. A description of the methodology and key assumptions used in performing the fire IPEEE and a discussion of the status of Appendix R modifications.

Response The fire IPEEE methodology consists of a progressive screening analysis, based on the EPRI FIVE methodology, as described in EPRI report TR-100370.

Watts Bar Nuclear Plant is currently designed for compliance with applicable Appendix R related requirements.

2. A summary of walkdown findings and a concise description of the walkdown team and the procedures used. This should include a description of the efforts to ensure that cable routing used in the analysis represents as-built information and the treatment of any existing dependence between remote shutdown and control room circuitry.

Response Plant walkdowns were performed to confirm the locations of potential fire ignition sources and to identify any safety related electrical raceways and components which could be affected by a postulated fire generated by these ignition sources.The walkdown findings are described in report WBN-IPE-004 U2.This report documents the zone of influence (ZOI) calculations for plant ignition sources and identifies the equipment and cable routing within the ZOL Cable routing information was confirmed during this process by physical area walkdown and review of plant documentation.

The walkdown teams were from the Operations staff and they received guidance and support from Fire Protection Engineering and System Engineering.

First and second party data collection was used to ensure that the walkdown data adequately reflects the WBN2 IPEEE DESIGN REPORT Page 254 A'I-IACHMENT 3

50 fire ignition sources. This two party process, in conjunction with the team staffing and support, provided confidence in the walkdown results.The remote shutdown capability was only credited for severe fires in the Control Building, which were conservatively assumed to require Control Room evacuation (see report WBN-IPE-005 U2). This system was specifically designed to provide an independent control capability for identified plant systems and functions, including any required control circuitry.

The remote shutdown capability system is described in Part IV (Alternate Shutdown Capability) of the Watts Bar Nuclear Plant Fire Protection Report.3. A discussion of the criteria used to identify critical fire areas and a list of critical areas, including (a) single areas in which equipment failures represent a serious erosion of safety margin, and (b) same as (a), but for double or multiple areas that share common barriers, penetration seals, HVAC ducting, etc.Response Based on the EPRI FIVE guidance, critical fire areas are considered to be those areas that contain either any Fire Safe Shutdown (FSSD)components or a Plant Trip Initiator (PTI) or could have a fire spread from an adjacent area. The qualitative screening analysis is documented in report WBN-IPE-O01 U2, which is summarized in Section 2. The specific areas that were not screened from consideration during this evaluation (i.e. "critical fire areas') are listed in Table 3.4. Fire ignition frequencies were generated for these areas and are documented in report WBN-IPE-002 U2.Each of the remaining areas was then evaluated on a quantitative basis, assuming that any postulated fire would totally engulf the area and result in a plant trip. If the resulting fire ignition frequency or resulting fire-related core damage frequency was less than 1E-06, further quantitative analysis was judged to be unnecessary and the area was screened from further consideration.

This process is described in reports WBN-IPE-002 U2 and WBN-IPE-003 U2, which are summarized in Sections 3 and 4, respectively, of this report.Detailed area analysis was then performed for the remaining Auxiliary Building, Control Building, Intake Pumping Station and Turbine Building.

This analysis is described in report WBN-IPE-005 U2. The results of this evaluation are summarized in Section 5.Fire hazards that could extend to include multiple fire areas were screened from further consideration, based on the fire barrier screening guidelines given in the EPRI FIVE documentation.

The potential for a multiple area fire developing in area 737.0-AIA, -ALAN, -A I B, -A I BN, -A IC and -AICN, and propagating to involve the adjacent areas is discussed under item (4) below.WBN2 IPEEE DESIGN REPORT.Page 255 ATTACHMENT 3

51 4. A discussion of the criteria used to determine fire size and duration and the treatment of cross-zone fire spread and associated major assumptions.

Response Fire size was conservatively assumed to be engulfing for postulated fires analyzed in the screening analysis described in report WBN-IPE-003 U2, which is summarized in Section 4. Fires were assumed to consume the fire ignition source for components located in the zone of influence in report WBN-IPE-004 U2.The Fire Events Database (NSA C/1 78L) was used as a basis for fire size for Turbine Building, Control Building and Auxiliary Building fires analyzed in Sections 1, 6 and 7 of report WBN-IPE-005 U2. These areas are summarized in Table 5.2. Fire duration was as required to consume the source.Cross-zone spread of fires was evaluated using the EPRI FIVE criteria, as described in report WBN-IPE-001 U2, which is summarized in Section.2. The potential for a postulated fire involving all of room 737. O-A 1 has been separately reviewed by NRC staff (Reference 20). In essence, this review concluded that the protection of trained cables for both divisions across the two 21 foot buffer areas (737.0-AIAN and 737.0-AIBN) provided between areas 737.0-A IA and -A IB and the 27 foot buffer area (737.0-AICN) separating 737.0-AIC from the above are adequate to prevent the spread of a fire developing in any of these areas in such a way as to impact any combination of the -A IA, -AIB, and -AIC sections of the room. Discussion of the evaluation of these areas, including the impact of automatic sprinkler suppression, which is installed throughout the room, is provided in report WBN-IPE-005 U2.5. A discussion of the fire initiating event database, including the plant specific database used. Provide documentation in each case where the plant specific data is less conservative than the data used in the approved fire vulnerability methodologies.

Describe methods for handling data, including major assumptions, the role of expert judgment, and the identification and evaluation of sources of data uncertainty.

Response The EPRI Fire Events Database (documented in NSA C/I 78L) was used to generate fire ignition frequencies, as described in the EPRI FIVE documentation.

Review of plant experience shows plant specific data to be no less conservative than the data given in the FIVE documentation.

Due to the use of a progressive screening analysis, data uncertainty was not explicitly modeled. For each of the areas that remained for more detailed analysis in report WBN-IPE-005 U2, a qualitative discussion of conservative assumptions is given in the associated report section. It should be noted that, with the exception of the use of the Alternate Shutdown Capability for selected severe fires in the Control Building, WBN2 IPEEE DESIGN REPORT Page 256 ATTACHMENT 3

52 recovery of equipment from fire-induced damage is conservatively not credited in this analysis.6. A discussion of the treatment of fire growth and spread, the spread of hot gases and smoke, and the analysis of detection and suppression and their associated assumptions, including the treatment of suppression induced damage to equipment.

Response Fire growth between areas is addressed by using the EPRI FIVE criteria.Detection and suppression are not evaluated as mitigating any fires in the initial quantitative screening evaluation documented in report WBN-IPE-003 U2. Detection and suppression were credited for selected cases, on a case-by-case basis, as described in the analysis of the Turbine Building, Control Building and Auxiliary Building areas of report WBN-IPE-005 U2.Suppression-induced damage is addressed under the associated Sandia issue in Section 7.2.1.2 7. A discussion of fire damage modeling, including the definition of fire-induced failures related to fire barriers and control systems and fire induced damage to cabinets.

A discussion of how human intervention is treated and how fire induced and non-fire induced failures are combined.

Identify recovery actions and types of fire mitigating actions for which credit is taken in these sequences.

Response Fire barrier effectiveness was evaluated using the EPRI FIVE criteria, as described in report WBN-IPE-001 U2 and documented in EPRI report TR-100370 (Reference 2). These criteria are summarized in Section 2 of this report. For this analysis, control systems were assumed to fail in such a way as to fail the function of the affected system. It should be noted that this analysis conservatively assumes that "hot short" failures occur whenever necessary to fail the system function.Electrical cabinet damage was conservatively assumed to occur for postulated fires in the area, with the exception of those Turbine Building, Control Building and Auxiliary Building areas evaluated in the revised report WBN-IPE-005 U2. In these areas, which are listed in Table 5.2, component damage was typically assumed to occur, based on the severity of the individual case under consideration.

Manual actions (including required time to accomplish the action and staffing limitations) were used in the Appendix R analysis and are documented in plant operating procedures (e.g. Abnormal Operating Procedure (AOI) 30.2). Plant Operations staff train regularly on this procedure.

These "human interventions" were credited in the FIVE analyses as justification that ability to safely shutdown was guaranteed.

Fire Brigade response for fire suppression activities were integrated into the revised report (WBN-IPE-005 U2) for selected areas of the Turbine, WBN2 IPEEE DESIGN REPORT Page 257 ATTACHMENT 3

53 Control and Auxiliary Buildings.

Non-fire induced failures are combined with fire-related impacts through use of the Level I PRA plant model.8. Discuss the treatment of fire detection and suppression, including fire fighting procedures, fire brigade training and adequacy of existing fire brigade equipment and treatment of access routes versus existing barriers.Response Fire suppression was only considered in the detailed analysis described in report WBN-IPE-005 U2 and only on a case-by-case basis for certain Turbine Building, Control Building and Auxiliary Building areas. Fire brigade training, equipment availability and procedures are described under the associated Sandia issue in Section 7.2.9. All functional and systemic event trees associated with fire-initiated sequences.

Response The plant model and associated event trees are as described in the Level I PRA report. Fire-initiated scenarios were incorporated by failing individual basic events within the Level 1 plant model. The individual event trees that were used to segment fire ignition frequency into individual cases, where this technique was used, are shown in report WBN-IPE-005 U2.10. A description of dominant functional and systemic sequences leading to core damage, along with their frequencies and percentage contribution to overall core damage frequency due to fire. Sequence selection criteria are as provided in Generic Letter 88-20 and NUREG-1335.

The description of the sequences should include a discussion of specific assumptions and human recovery actions.Response The results of the fire risk analysis are summarized and discussed in Section 6. Due to the use of a progressive screening approach, as described in the EPRI FIVE documentation, individual scenarios are not listed for areas that were screened from further consideration, based on fire-related core damage frequency of less than 1E-06.11. The estimated core damage frequency, the timing of the associated core damage, a list of analytical assumptions, including their bases, and the sources of uncertainty.

Response The results of this analysis are summarized in Section 6. The analytical assumptions used to evaluate each plant area are provided with the discussion in the associated text. Due to the use of a screening analysis, plant damage states would only be evaluated for unscreened areas.Also, a separate analysis of data uncertainty was not performed due to use of a screening analysis.12. Any fire induced containment failures identified as being different from those identified in the internal events analysis.WBN2 IPEEE DESIGN REPORT Page 258 ATTACHMENT 3

54 Response Containment failure due to fire-induced damage was addressed in Section 7.1. This review concluded that no significant containment failures were introduced by the analysis of internal fires.13. Documentation with regard to the decay heat removal function and Fire Risk Scoping Study issues addressed by the submittal, the basis and assumptions used to address these issues, and a discussion of the findings and conclusions.

Evaluation results and potential improvements should be specifically highlighted.

Specifically, NUREG-1407 (Section 4) specifies that the submittal should address the following Fire Risk Scoping Study issues:-Seismic/fire interactions.

-Effect of fire suppressant systems on safety equipment.-Control system interactions.

Response The issues raised in the Fire Risk Scoping Study (NUREG/CR-5088) are addressed in Section 7.2.14. When an existing PRA is used to address the fire IPEEE, the licensee should describe sensitivity studies related to the use of the initial hazard, supplemental plant walkdown results and subsequent evaluations.

The licensee should examine the above list to fill in those items missed in the existing fire PRA.Response Only the plant model was used from the Level 1 PRA. In particular, this model was used specifically to capture the non-fire induced failures that could occur and to model plant response, following the incorporation of fire-induced failures.WBN2 IPEEE DESIGN REPORT Page 259 ATTACHMENT 3

55 8. Validation of the Unit 2 "As Designed" Fire Induced Vulnerability Evaluation The Fire Induced Vulnerability Evaluation (FIVE) documented in this summary represents an analysis of the "as designed" condition of Watts Bar Nuclear Unit 2.Since Generic Letter (GL) 88-20 specifies that the IPEEE evaluation should be based upon the "as built" configuration of the plant, the FIVE analyses of the "as designed" configuration will be validated when construction is complete to meet the "as built" GL 88-20 criterion.

The FIVE validation effort will be comprised of the following activities:

Validation Activities for Phase I The Unit 2 population of rooms with Appendix R Safe Shutdown (SSD) Equipment will be reviewed to ensure that no safe shutdown components or plant trip initiators have been added to the scope. If any of these are discovered, they will be evaluated via the FIVE process. A representative population of rooms will be reviewed to ensure that each room's configuration, barrier ratings, room use, etc.has not changed. Based on the results of this review, rooms will be reanalyzed as necessary and changes incorporated into the analysis.Validation Activities for Phase II, Step 1 A representative population of Unit 2 rooms will be reviewed to verify that there have been no significant changes in the room ignition -frequencies which would result in a less conservative analysis result. New walkdowns will be performed and incorporated into the analysis as necessary.

Validation Activities for Phase II, Step 2 The "as built" equipment and location data for the Unit 2 Appendix R SSD equipment and safety injection/recirculation equipment will be reviewed and incorporated into the Plant Probabilistic Risk Assessment (PRA) as necessary to update the analysis.

Manual actions credited in the analysis will be confirmed.

Also, the latest Plant PRA will be compared to the "as designed" version of the model and updated if needed.Validation Activities for Phase II, Step 3 Report WBN-IPEO-005 U2 will be updated as necessary.

This includes reviewing and updating both the assumptions and event trees as required.Validation Activities for Phase III All applicable reports, including the summary, associated with the Unit 2 FIVE evaluation will be updated as necessary.

Other Validation Activities A peer review of the Unit 2 analysis will be performed prior to submittal of the Validated "As Built" Analysis Report. This review will be similar to the review performed for the Unit 1 evaluation.

WBN2 IPEEE DESIGN REPORT Page 260 ATTACHMENT 3

56 9. REFERENCES

1. Electric Power Research Institute, "Fire Events Database for U.S. Nuclear Power Plants, "NSAC/178L, Revision 1, January 1993.2. Electric Power Research Institute, "Fire-Induced Vulnerability Evaluation (FIVE)," TR-100370, April 1992.3. Electric Power Research Institute, "Fire PRA Implementation Guide," TR-105928, Final Report, December, 1995.4. Houghton, J.R., "Special Study -Fire Events -Feedback of U.S. Operating Experience," AEOD/S97-03, prepared for the U.S. Nuclear Regulatory Commission, June, 1997.5. ABSG Consulting, "Determination of Fire Scenario Safe Shutdown Path Unavailability Watts Bar Nuclear Plant Unit 2," Report WBN-IPE-003 U2.6. Sandia National Laboratories, "Fire Risk Scoping Study: Investigation of Nuclear Power Plant Fire Risk, Including Previously Unaddressed Issues," prepared for the U.S. NRC, NUREG/CR-5088, January, 1989.7 Tennessee Valley Authority, "Combustible Loading Data Summary," Drawing Series 45W893-035 (Replaced calculation EPMDOM012990).
8. EPM, "Fire Compartment Interaction Analysis," Report WBN-IPE-001 U2.9. Tennessee Valley Authority, "Ignition Source Data," Unit 1 Report WBN-IPE-002.
10. EPM, "Ignition Source Data," Report WBN-IPE-002 U2.11. Tennessee Valley Authority, "Watts Bar Nuclear Plant Fire Protection Report," Revision 37, May 23, 2008 12. EPM, "Fire Damage Zone of Influence," Report WBN-IPE-004 U2.13. ABS Consulting, "IPEEE (Fire) -Quantitative Screening

-Phase 2 (Detailed Analysis)," Report WBN-IPE-005 U2.14. Tennessee Valley Authority, "Fire Safe Shutdown," Watts Bar Nuclear Plant Abnormal Operating Instruction AOI-30.2, Revision 31, January 6, 2010.15. Tennessee Valley Authority, "Evaluation of Suppression System Discharge," Calculation EPM-RAC-032392, Revision 1.16. Tennessee Valley Authority, "Seismic Capability Walkdown for IPEEE," Calculation WCG-1-1840, Revision 1, July, 1997.WBN2 IPEEE DESIGN REPORT Page 261 ATTACHMENT 3

57 17. Tennessee Valley Authority, "IPEEE Outlier Interaction Evaluation," Calculation WCG-1-1842, Revision 1, February 2, 1998.18. EPM, "Unit 1 and 2 Appendix R Safe Shutdown.

Analysis", Calculation EDQ00099920090012, Revision 0.19. U.S. NRC, "Individual Plant Examination for External Events (IPEEE)," Supplement 4 to Generic Letter 88-20, NUREG-1407, Final Report, June, 1991.20. U.S. NRC, "Safety Evaluation Report Related to the Operation of Watts Bar Nuclear Plant, Units 1 and 2," NUREG-0847, Supplement 18, October 1995 21. Tennessee Valley Authority, "Watts Bar Unit 2 PRA Model", Revision 0, January 22, 2010 WBN2 IPEEE DESIGN REPORT Page 262 ATTACHMENT 3

Enclosure 2 List of Commitments

1. Prior to fuel load, a final report will be submitted following certain validation activities as described in the IPEEE Design Report (Enclosure 1).
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