ML16012A365

2015-12 Final Operating Test

ML16012A365
Person / Time
Site:	Fort Calhoun
Issue date:	12/14/2015
From:	Vincent Gaddy Operations Branch IV
To:
References
Download: ML16012A365 (436)

Text

Appendix C JPM WORKSHEET Form ES-C-1 Facility: FCS JPM # NRC RA1 Task # 1361 K/A # 2.1.25 3.9 / 4.2

Title:

Perform a Time to Boil Determination Examinee (Print):

Testing Method:

Simulated Performance: Classroom: X Actual Performance: X Simulator:

Alternate Path: Plant:

Time Critical:

READ TO THE EXAMINEE I will explain the Initial Conditions, which steps to simulate or discuss, and provide an Initiating Cue.

When you complete the task successfully, the objective for this JPM will be satisfied.

Initial Conditions: Given the following conditions:

• The plant was shut down 5 days ago for a Reactor Coolant Pump seal repair.

• AOP-19, Loss of Shutdown Cooling, was just entered at 0800.

• Refer to the Attached ERFCS printout for page 195, Shutdown Status Board, for current plant conditions.

Initiating Cue: The Control Room Supervisor directs you to PERFORM the following:

• PERFORM a Time to Boil Determination per AOP-19, Loss of Shutdown Cooling, Attachment B, Time to Boil Determination Worksheet.

• RECORD required information on Attachment B, Time to Boil Determination Worksheet wherever appears.

• Instrument numbers will be looked up by another operator.

Task Standard: Utilizing AOP-19, located RCS at Mid Loop graph, recorded appropriate Time to Boil data, and determined Time to Boil at 18 +/- 1 minutes.

Required Materials: AOP-19, Loss of Shutdown Cooling, Rev. 18.

Validation Time: 7 minutes Completion Time: ________ minutes Comments:

Result: SAT UNSAT Examiner (Print / Sign): Date:

Page 1 of 5 NRC Admin JPM RA1 Rev. Final

Appendix C JPM WORKSHEET Form ES-C-1 CLASSROOM SETUP EXAMINER:

PROVIDE the examinee with a copy of:

• AOP-19, Loss of Shutdown Cooling.

• Attachment B, Time to Boil Determination Worksheet.

• Attachment C, Time to Boil Determination.

Page 2 of 5 NRC Admin JPM RA1 Rev. Final

Appendix C JPM STEPS Form ES-C-1

- Check Mark Denotes Critical Step START TIME:

Examiner Note: The following steps are from AOP-19, Attachment B.

Perform Step: 1 Time Shutdown Cooling was lost: _____

1 Standard: RECORDED time Shutdown Cooling was lost as 0800 on Attachment B.

Comment: SAT UNSAT Perform Step: 2 Last known RCS/SDCS temperature: _____ °F from instrument number:

2 _____

Standard: DETERMINED representative RCS temperature should be recorded for Core Exit Thermocouples, and RECORDED last known and HIGHEST RCS/SDCS temperature of 110°F from CETs on Attachment B.

Comment: SAT UNSAT Perform Step: 3 Record the following information and inform the Shift Manager on 10 3 minute intervals.

Standard: DETERMINED from ERFCS printout that the RCS is at MID LOOP, and referred to Mid Loop graph from Attachment B, and RECORDED 18 +/- 1 minutes on Attachment B.

Terminating Cue: This JPM is complete.

Comment: SAT UNSAT STOP TIME:

Page 3 of 5 NRC Admin JPM RA1 Rev. Final

Appendix C JPM CUE SHEET Form ES-C-1 INITIAL CONDITIONS: Given the following conditions:

• The plant was shut down 5 days ago for a Reactor Coolant Pump seal repair.

• AOP-19, Loss of Shutdown Cooling, was just entered at 0800.

• Refer to the Attached ERFCS printout for page 195, Shutdown Status Board, for current plant conditions.

INITIATING CUE: The Control Room Supervisor directs you to PERFORM the following:

• PERFORM a Time to Boil Determination per AOP-19, Loss of Shutdown Cooling, Attachment B, Time to Boil Determination Worksheet.

• RECORD required information on Attachment B, Time to Boil Determination Worksheet wherever appears.

• Instrument numbers will be looked up by another operator.

Page 4 of 5 NRC Admin JPM RA1 Rev. Final

Appendix C JPM CUE SHEET Form ES-C-1 Attachment A - ERFCS Page 195, Shutdown Status Board Page 5 of 5 NRC Admin JPM RA1 Rev. Final

Appendix C JPM WORKSHEET Form ES-C-1 Facility: FCS JPM # NRC RA2 Task # 1528 K/A # 2.1.43 4.1 / 4.3

Title:

Calculate an Estimated Critical Boron Concentration Examinee (Print):

Testing Method:

Simulated Performance: Classroom: X Actual Performance: X Simulator:

Alternate Path: Plant:

Time Critical:

READ TO THE EXAMINEE I will explain the Initial Conditions, which steps to simulate or discuss, and provide an Initiating Cue.

When you complete the task successfully, the objective for this JPM will be satisfied.

Initial Conditions: Given the following conditions:

• A Reactor Trip from 100% power occurred on 12/03/15 at 0400 following 6 months of full power operation.

• All Control Rods were fully withdrawn at the time of the trip.

• Boron concentration prior to the trip was 400 ppm.

• Average Core Burnup is 6 GWD/MTU.

• Current boron concentration is 500 ppm.

• Criticality is scheduled to occur with Regulating Group 4 at 78 inches.

• Reactor Engineering reports no correction is needed for boron depletion.

• Reactor Startup is scheduled for 12/10/15 at 0600.

Initiating Cue: The Control Room Supervisor directs you to PERFORM the following:

• CALCULATE estimated critical boron concentration per TDB-V-1-B, Estimated Critical Conditions Worksheet.

• COMPLETE data entry through TDB-V.1.B, Step E.3.d, Estimated Critical Boron Concentration.

Task Standard: Utilizing TDB-V.1.B and TDB-II, calculated Estimated Critical Boron Concentration.

Required Materials: TDB-V.1.B, Estimated Critical Conditions Worksheet, Rev. 26.

TDB-II, Technical Data Book Reactivity Curves, Rev. 35.

Validation Time: 40 minutes Completion Time: ________ minutes Comments:

Result: SAT UNSAT Examiner (Print / Sign): Date:

Page 1 of 5 NRC Admin JPM RA2 Rev. Final

Appendix C JPM WORKSHEET Form ES-C-1 CLASSROOM SETUP EXAMINER:

PROVIDE the examinee with a copy of:

• TDB-V.1.B, Estimated Critical Conditions Worksheet.

• TDB-II, Technical Data Book Reactivity Curves.

• Calculator

• Straight Edge Page 2 of 5 NRC Admin JPM RA2 Rev. Final

Appendix C JPM STEPS Form ES-C-1

- Check Mark Denotes Critical Step START TIME:

Examiner Note: The following steps are from TDB-V.1.B Perform Step: 1 Conditions at Time of Shutdown.

A Standard: ENTERED Conditions at Time of Shutdown in TDB-V-1-B Steps A.1 to A.5.

Examiner Note: Information found on Answer Key.

Comment: SAT UNSAT Perform Step: 2 Conditions at Time of Startup.

B Standard: ENTERED Conditions at Time of Startup in TDB-V.1.B Steps B.1 to B.4.

Examiner Note: Information found on Answer Key.

Comment: SAT UNSAT Perform Step: 3 ECC Applicability.

C Standard: DETERMINED early and late date/time limits for ECC Applicability and entered data in TDB-V.1.B Steps C.1 to C.4.

Examiner Note: Information found on Answer Key.

Comment: SAT UNSAT Perform Step: 4 Reactivity Changes Due To Shutdown.

D Standard: CALCULATED and ENTERED Reactivity Changes Due To Shutdown in TDB-V.1.B Steps D.1 to D.5.

Examiner Note: Information found on Answer Key.

Comment: SAT UNSAT Page 3 of 5 NRC Admin JPM RA2 Rev. Final

Appendix C JPM STEPS Form ES-C-1 Perform Step: 5 Estimated Critical Boron Concentration.

E Standard: CALCULATED and ENTERED Estimated Critical Boron Concentration in TDB-V.1.B Steps E.1 to E.3.

Step E.3.d: Calculated 884 +/- 25 ppm (Critical)

Examiner Note: Information found on Answer Key.

Terminating Cue: This JPM is complete.

Comment: SAT UNSAT STOP TIME:

Page 4 of 5 NRC Admin JPM RA2 Rev. Final

Appendix C JPM CUE SHEET Form ES-C-1 INITIAL CONDITIONS: Given the following conditions:

• A Reactor Trip from 100% power occurred on 12/03/15 at 0400 following 6 months of full power operation.

• All Control Rods were fully withdrawn at the time of the trip.

• Boron concentration prior to the trip was 400 ppm.

• Average Core Burnup is 6 GMWD/MTU.

• Current boron concentration is 500 ppm.

• Criticality is scheduled to occur with Regulating Group 4 at 78 inches.

• Reactor Engineering reports no correction is needed for boron depletion.

• Reactor Startup is scheduled for 12/10/15 at 0600.

INITIATING CUE: The Control Room Supervisor directs you to PERFORM the following:

• CALCULATE estimated critical boron concentration per TDB-V.1.B, Estimated Critical Conditions Worksheet.

• COMPLETE data entry through TDB-V.1.B, Step E.3.d, Estimated Critical Boron Concentration.

Page 5 of 5 NRC Admin JPM RA2 Rev. Final

Appendix C JPM WORKSHEET Form ES-C-1 Facility: FCS JPM # NRC RA3 Task # 0066 K/A # 2.2.35 3.6 / 4.5

Title:

Determine Technical Specification MODE of Operation Examinee (Print):

Testing Method:

Simulated Performance: Classroom: X Actual Performance: X Simulator:

Alternate Path: Plant:

Time Critical:

READ TO THE EXAMINEE I will explain the Initial Conditions, which steps to simulate or discuss, and provide an Initiating Cue.

When you complete the task successfully, the objective for this JPM will be satisfied.

Initial Conditions: Given the following conditions:

• Reactor Coolant System temperature is 200°F.

• Reactor Coolant System pressure is 150 psia.

• Reactor Coolant System boron concentration is 2180 ppm.

• Core Burnup is 1500 MWD/MTU.

• A Shutdown Cooling Pump is in operation.

Initiating Cue: The Control Room Supervisor directs you to PERFORM the following:

• DETERMINE Fort Calhoun Station Technical Specification Reactor Operating Condition.

• Refueling Boron Concentration _____ ppm.

• Operating Mode _____.

Task Standard: Utilizing Technical Specifications and Core Operating Limits Report, determined Fort Calhoun Station is in Operating Mode 5, Refueling Shutdown Condition.

Required Materials: Fort Calhoun Station Technical Specifications, Amendment #283.

TDB-VI, Core Operating Limits Report, Rev. 42.

Validation Time: 5 minutes Completion Time: ________ minutes Comments:

Result: SAT UNSAT Examiner (Print / Sign): Date:

Page 1 of 4 NRC Admin JPM RA3 Rev. Final

Appendix C JPM WORKSHEET Form ES-C-1 CLASSROOM SETUP EXAMINER:

PROVIDE the examinee with a copy of:

• Fort Calhoun Station Technical Specifications.

• TDB-VI, Core Operating Limits Report.

Page 2 of 4 NRC Admin JPM RA3 Rev. Final

Appendix C JPM STEPS Form ES-C-1

- Check Mark Denotes Critical Step START TIME:

Examiner Note: The following steps are from Technical Specifications and the Core Operating Limits Report.

Perform Step: 1 Refer to Technical Specifications for MODE definition.

Standard: REFERRED to Technical Specification Definitions, Page 2 and DETERMINED that plant is either in MODE 4 or 5 depending on boron concentration.

Comment: SAT UNSAT Perform Step: 2 Refer to Core Operating Limits Report to determine REFUELING BORON CONCENTRATION.

Standard: REFERRED to Core Operating Limits Report and DETERMINED REFUELING BORON CONCENTRATION at 1500 MWD/MTU is 2160 ppm.

Comment: SAT UNSAT Perform Step: 3 Determine Plant Operational Mode based on Reactor Coolant System Boron Concentration.

Standard: REFERRED to Technical Specification Definitions, Page 2 and DETERMINED that Plant is in Operating Mode 5, Refueling Shutdown Condition based on Reactor Coolant System Boron Concentration greater than REFUELING BORON CONCENTRATION.

Terminating Cue: This JPM is complete.

Comment: SAT UNSAT STOP TIME:

Page 3 of 4 NRC Admin JPM RA3 Rev. Final

Appendix C JPM CUE SHEET Form ES-C-1 INITIAL CONDITIONS: Given the following conditions:

• Reactor Coolant System temperature is 200°F.

• Reactor Coolant System pressure is 150 psia.

• Reactor Coolant System boron concentration is 2180 ppm.

• Core Burnup is 1500 MWD/MTU.

• A Shutdown Cooling Pump is in operation.

INITIATING CUE: The Control Room Supervisor directs you to PERFORM the following:

• DETERMINE Fort Calhoun Station Technical Specification Reactor Operating Condition.

• Refueling Boron Concentration _____ ppm.

• Operating Mode _____.

Page 4 of 4 NRC Admin JPM RA3 Rev. Final

Appendix C JPM WORKSHEET Form ES-C-1 Facility: FCS JPM # NRC RA4 Task # 1269 K/A # 2.3.11 3.8 / 4.3

Title:

Respond to Voids in the Reactor Coolant System Examinee (Print):

Testing Method:

Simulated Performance: Classroom: X Actual Performance: X Simulator:

Alternate Path: Plant:

Time Critical:

READ TO THE EXAMINEE I will explain the Initial Conditions, which steps to simulate or discuss, and provide an Initiating Cue.

When you complete the task successfully, the objective for this JPM will be satisfied.

Initial Conditions: Given the following conditions:

• A Small Break Loss of Coolant Accident (LOCA) has occurred.

• EOP-03, Loss of Coolant Accident, has been implemented.

• Reactor Coolant System (RCS) conditions:

• RCS Pressure = 450 psia.

• RCS TCOLD = 402°F.

• Pressurizer (PZR) conditions:

• PZR Level [actual] = 60% and stable.

• PZR Temperature = 456°F and stable.

• Reactor Vessel Level Monitoring System (RVLMS) is 83% and stable.

• Containment conditions:

• Containment Safety Function is satisfied.

• All Containment Ventilation Fans are operating.

• Containment Pressure = 1.2 psig.

• Containment Temperature = 118°F.

• Containment Hydrogen concentration = 1.2%.

• RC-5, Pressurizer Quench Tank (PZR QT), conditions:

• PZR QT Level = 70%.

• PZR QT Pressure = 5 psig.

• HPSI Stop and Throttle has been performed for a LOCA.

• RCS and Pressurizer sample results are normal.

• The Technical Support Center (TSC) is activated.

• Use of EOP/AOP Attachment IC-14, RCS Void Elimination, has been unsuccessful in eliminating the RCS voids.

Initiating Cue: The Control Room Supervisor directs you to PERFORM the following:

• IMPLEMENT OI-RC-12, Post Accident Venting of Noncondensable Gases from the Reactor Coolant System.

• VENT from (CIRCLE one): Reactor Vessel Head PZR

• VENT to (CIRCLE one): Containment PZR Quench Tank Page 1 of 9 NRC Admin JPM RA4 Rev. Final

Appendix C JPM WORKSHEET Form ES-C-1 Task Standard: Utilizing OI-RC-12, determined vent path from the Reactor Vessel Head to the Pressurizer Quench Tank is required.

Required Materials: OI-RC-12, Post Accident Venting of Noncondensable Gases from the Reactor Coolant System, Rev. 11.

Validation Time: 12 minutes Completion Time: ________ minutes Comments:

Result: SAT UNSAT Examiner (Print / Sign): Date:

Page 2 of 9 NRC Admin JPM RA4 Rev. Final

Appendix C JPM WORKSHEET Form ES-C-1 CLASSROOM SETUP EXAMINER:

PROVIDE the examinee with a copy of:

• OI-RC-12, Post Accident Venting of Noncondensable Gases from the Reactor Coolant System.

• Steam Tables Page 3 of 9 NRC Admin JPM RA4 Rev. Final

Appendix C JPM CUE SHEET Form ES-C-1

- Check Mark Denotes Critical Step START TIME:

Examiner Note: The following step is from OI-RC-12, Attachment 1, Prerequisites.

Perform Step: 1 PREREQUISITES:

1, 2, & 3

• Procedure Revision Verification Revision No. _____ Date: _____

• The Reactor is subcritical with a Tave less than 515°F (Ref.

Technical Specification 2.1.8).

• The RCS is being maintained in a stable condition with the following:

• Pressurizer (PZR) Level is between 49% and 93%

• Charging flow is in operation

• RCS subcooling is between 20°F and 200°F Standard: DETERMINED the following per the Initial Conditions:

• Procedure Revision is as provided.

• Reactor Coolant System TAVE is ~402°F.

• Pressurizer Level is 60%.

• Charging flow is in operation (based on HPSI Stop and Throttle has been performed).

• RCS subcooling is ~54°F.

Comment: SAT UNSAT Page 4 of 9 NRC Admin JPM RA4 Rev. Final

Appendix C JPM CUE SHEET Form ES-C-1 Examiner Note: The following steps are from OI-RC-12, Attachment 1, Procedure.

Perform Step: 2 IF one or more of the following conditions are present in the PZR, THEN 1 & all bullets determine the venting path per Attachment 2:

• Figure 1 indicates the presence of non-condensable gases

• Departure from saturation

• Sluggish pressure control

• Sampling results indicate non-condensable gases Standard: IDENTIFIED that the Pressurizer does not indicate a departure from saturated conditions and no other conditions warrant Pressurizer venting.

Comment: SAT UNSAT Perform Step: 3 Determine if bubble exists in the RV Head by monitoring RV level less 2 than 100% via the Reactor Vessel Level Monitoring System (RVLMS),

THEN determine the venting path per Attachment 2.

Standard: IDENTIFIED that a bubble exists in the Reactor Vessel Head and CIRCLED Vent from: Reactor Vessel Head Comment: SAT UNSAT Page 5 of 9 NRC Admin JPM RA4 Rev. Final

Appendix C JPM CUE SHEET Form ES-C-1 Examiner Note: The following step is from OI-RC-12, Attachment 2, Prerequisites.

Perform Step: 4 PREREQUISITES:

1, 2, 3, 4, 5, & 6

• Procedure Revision Verification Revision No. _____ Date: _____

• The reactor is subcritical with a Tave less than 515°F (Ref.

Technical Specification 2.1.8).

• Containment Isolation has been verified per EOP Safety Function Status Check.

• All available Containment Ventilation Units are in operation:

• VA-3A, Cntmt Vent Fan

• VA-3B, Cntmt Vent Fan

• VA-7C, Cntmt Vent Fan

• VA-7D, Cntmt Vent Fan

• The RCS is being maintained in a stable condition with the following:

• Pressurizer (PZR) Level is between 49% and 93%

• Charging flow is in operation

• RCS subcooling is between 20°F and 200°F

• TSC has been activated.

Standard: DETERMINED the following per the Initial Conditions:

• Procedure Revision is as provided.

• Reactor Coolant System TAVE is ~402°F.

• Containment Isolation has been verified.

• All Containment Ventilation Units are in operation.

• Pressurizer Level is 60%.

• Charging flow is in operation (based on HPSI Stop and Throttle has been performed).

• RCS subcooling is ~54°F.

• Technical Support Center has been activated.

Comment: SAT UNSAT Page 6 of 9 NRC Admin JPM RA4 Rev. Final

Appendix C JPM CUE SHEET Form ES-C-1 Examiner Note: The following steps are from OI-RC-12, Attachment 2, Procedure.

Perform Step: 5 IF one or more of the following conditions exist, THEN the Containment 1, 1.a, & all bullets vent path should be used per the following:

• Verify:

• There is no water in the RC-5, PQT AND DC Bus 1 electrical power source is available to RCGVS Valves

• Large quantities of gas need to be vented

• Rapid venting is required

• The potential for loss of core cooling exists

• There is serious interference with the ability to maintain pressure control Standard: IDENTIFIED that the Containment vent path is NOT preferred.

Comment: SAT UNSAT Page 7 of 9 NRC Admin JPM RA4 Rev. Final

Appendix C JPM CUE SHEET Form ES-C-1 Perform Step: 6 IF the following conditions exist, THEN RC-5, Pressurizer Quench Tank 2, 2.a, & all bullets (PQT) vent path should be used per the following:

• Verify:

• There is water in the PQT AND DC Bus 2 electrical power source is available to RCGVS Valves

• Small quantities of gas need to be vented

• Rapid venting is not required Standard: IDENTIFIED the following:

• A bubble exists in the Reactor Vessel Head. (NOT critical)

• Control power exists to the valves. (NOT critical)

• Quench Tank is the preferred venting path. (critical)

• Attachment 4, Venting RV Head to the Pressurizer Quench Tank. (NOT critical)

CIRCLED VENT to: PZR Quench Tank (critical)

Terminating Cue: This JPM is complete.

Comment: SAT UNSAT STOP TIME:

Page 8 of 9 NRC Admin JPM RA4 Rev. Final

Appendix C JPM CUE SHEET Form ES-C-1 INITIAL CONDITIONS: Given the following conditions:

• A Small Break Loss of Coolant Accident (LOCA) has occurred.

• EOP-03, Loss of Coolant Accident, has been implemented.

• Reactor Coolant System (RCS) conditions:

• RCS Pressure = 450 psia.

• RCS TCOLD = 402°F.

• Pressurizer (PZR) conditions:

• PZR Level [actual] = 60% and stable.

• PZR Temperature = 456°F and stable.

• Reactor Vessel Level Monitoring System (RVLMS) is 83%

and stable.

• Containment conditions:

• Containment Safety Function is satisfied.

• All Containment Ventilation Fans are operating.

• Containment Pressure = 1.2 psig.

• Containment Temperature = 118°F.

• Containment Hydrogen concentration = 1.2%.

• RC-5, Pressurizer Quench Tank (PZR QT), conditions:

• PZR QT Level = 70%.

• PZR QT Pressure = 5 psig.

• HPSI Stop and Throttle has been performed for a LOCA.

• RCS and Pressurizer sample results are normal.

• The Technical Support Center (TSC) is activated.

• Use of EOP/AOP Attachment IC-14, RCS Void Elimination, has been unsuccessful in eliminating the RCS voids.

INITIATING CUE: The Control Room Supervisor directs you to PERFORM the following:

• IMPLEMENT OI-RC-12, Post Accident Venting of Noncondensable Gases from the Reactor Coolant System.

• VENT from (CIRCLE one):

Reactor Vessel Head PZR

• VENT to (CIRCLE one):

Containment PZR Quench Tank Page 9 of 9 NRC Admin JPM RA4 Rev. Final

Appendix C JPM WORKSHEET Form ES-C-1 Facility: FCS JPM # NRC SA1 Task # 1363 K/A # 2.1.25 3.9 / 4.2

Title:

Perform an Alternate Decay Heat Removal Method Determination Examinee (Print):

Testing Method:

Simulated Performance: Classroom: X Actual Performance: X Simulator:

Alternate Path: Plant:

Time Critical:

READ TO THE EXAMINEE I will explain the Initial Conditions, which steps to simulate or discuss, and provide an Initiating Cue.

When you complete the task successfully, the objective for this JPM will be satisfied.

Initial Conditions: Given the following conditions:

• Plant was operating for 6 weeks at 100% power when a Reactor Coolant Pump seal failed.

• AOP-19, Loss of Shutdown Cooling, has been entered 7 days after shutdown.

• The Pressurizer manway has been removed.

• HCV-347, Shutdown Cooling Loop 2 Isolation Valve, is closed and cannot be reopened.

• High Pressure Safety Injection (HPSI) Pump SI-2A is available with flow of 300 gpm at 250 psia discharge pressure.

Initiating Cue: The Shift Manager directs you to PERFORM the following:

• PERFORM AOP-19, Loss of Shutdown Cooling, Attachment D, Alternate Decay Heat Removal Method Determination.

• CIRCLE the appropriate Alternate Decay Heat Removal Attachment on Attachment D (indicate your decision path on Attachment D).

Task Standard: Utilizing AOP-19, determined Reactor Coolant System pressure boundary was not intact, Reactor Vessel Head was installed, Shutdown Cooling flow was available, HPSI flow was available but insufficient, and identified Attachment E as the Alternate Decay Heat Removal Method.

Required Materials: AOP-19, Loss of Shutdown Cooling, Rev. 18.

Validation Time: 11 minutes Completion Time: ________ minutes Comments:

Result: SAT UNSAT Examiner (Print / Sign): Date:

Page 1 of 5 NRC Admin JPM SA1 post exam update.docx

Appendix C JPM WORKSHEET Form ES-C-1 CLASSROOM SETUP EXAMINER:

PROVIDE the examinee with a copy of:

• AOP-19, Loss of Shutdown Cooling.

• Attachment D, Alternate Decay Heat Removal Method Determination.

Page 2 of 5 NRC Admin JPM SA1 post exam update.docx

Appendix C JPM STEPS Form ES-C-1

- Check Mark Denotes Critical Step START TIME:

Examiner Note: The following steps are from AOP-19, Attachment D.

Examiner Note: REFER to Answer Key to follow Attachment D flowpath.

Perform Step: 1 IS RCS Pressure Boundary Intact?

1 Standard: DETERMINED answer was NO on Attachment D and PROCEEDED to next box.

Comment: SAT UNSAT Perform Step: 2 IS Reactor Vessel Head on?

2 Standard: DETERMINED answer was YES on Attachment D and PROCEEDED to next box.

Comment: SAT UNSAT Perform Step: 3 Is SDC Discharge Available?

3 Standard: DETERMINED answer was Yes on Attachment D and PROCEEDED to next box.

Comment: SAT UNSAT Perform Step: 4 Is HPSI Discharge Available?

4 Standard: DETERMINED answer was YES on Attachment D and PROCEEDED to next box.

Comment: SAT UNSAT Page 3 of 5 NRC Admin JPM SA1 post exam update.docx

Appendix C JPM STEPS Form ES-C-1 Perform Step: 5 Is Sufficient Injection Available?

5 Standard: PERFORMED the following:

• REFERRED to note (*) and DETERMINED plant operation at 100% power for greater than 30 days.

• DETERMINED Time after Shutdown was 7 days ago.

• DETERMINED Required (gpm) is >310 gpm but < 385 gpm.

• DETERMINED Sufficient Injection Flow NOT available and PROCEEDED to the next box.

Comment: SAT UNSAT Perform Step: 6 GO TO Attachment E Alternate Decay Heat Removal by Boiling.

6 Standard: DETERMINED Attachment E, Alternate Decay Heat Removal by Boiling is the appropriate Attachment.

Terminating Cue: This JPM is complete.

Comment: SAT UNSAT STOP TIME:

Page 4 of 5 NRC Admin JPM SA1 post exam update.docx

Appendix C JPM CUE SHEET Form ES-C-1 INITIAL CONDITIONS: Given the following conditions:

• The plant was operating for 6 weeks at 100% power when a Reactor Coolant Pump seal failed.

• AOP-19, Loss of Shutdown Cooling, has been entered 7 days after shutdown.

• The Pressurizer manway has been removed.

• HCV-347, Shutdown Cooling Loop 2 Isolation Valve, is closed and cannot be reopened.

• High Pressure Safety Injection (HPSI) Pump SI-2A is available with flow of 300 gpm at 250 psia discharge pressure.

INITIATING CUE: The Shift Manager directs you to PERFORM the following:

• PERFORM AOP-19, Loss of Shutdown Cooling, Attachment D, Alternate Decay Heat Removal Method Determination.

• CIRCLE the appropriate Alternate Decay Heat Removal Attachment on Attachment D (indicate your decision path on Attachment D).

Page 5 of 5 NRC Admin JPM SA1 post exam update.docx

Attachment D Alternate Decay Heat Removal Method Determination Determine the method of Alternate Decay Heat Removal from the Flow Chart.

• Appropriate flow required to remove heat by Time after shutdown Required injection, based on 100% full power (days) (gpm) operation for greater than 30 days. 1 575 5 385 10 310 30 230 End of Attachment D

Attachment D Alternate Decay Heat Removal Method Determination Determine the method of Alternate Decay Heat Removal from the Flow Chart.

• Appropriate flow required to remove heat by Time after shutdown Required injection, based on 100% full power (days) (gpm) operation for greater than 30 days. 1 575 5 385 10 310 30 230 End of Attachment D

Appendix C JPM WORKSHEET Form ES-C-1 Facility: FCS JPM # NRC SA2 Task # 1528 K/A # 2.1.43 4.1 / 4.3

Title:

Calculate an Estimated Critical Boron Concentration Examinee (Print):

Testing Method:

Simulated Performance: Classroom: X Actual Performance: X Simulator:

Alternate Path: Plant:

Time Critical:

READ TO THE EXAMINEE I will explain the Initial Conditions, which steps to simulate or discuss, and provide an Initiating Cue.

When you complete the task successfully, the objective for this JPM will be satisfied.

Initial Conditions: Given the following conditions:

• A Reactor Trip from 100% power occurred on 12/03/15 at 0400 following 6 months of full power operation.

• All Control Rods were fully withdrawn at the time of the trip.

• Boron concentration prior to the trip was 400 ppm.

• Average Core Burnup is 6 GWD/MTU.

• Current boron concentration is 500 ppm.

• Criticality is scheduled to occur with Regulating Group 4 at 78 inches.

• Reactor Engineering reports no correction is needed for boron depletion.

• Reactor Startup is scheduled for 12/10/15 at 0600.

Initiating Cue: The Shift Manager directs you to PERFORM the following:

• CALCULATE estimated critical boron concentration per TDB-V.1.B, Estimated Critical Conditions Worksheet.

• COMPLETE data entry through TDB-V.1.B, Step F.4, Estimated Critical Condition Summary.

Task Standard: Utilizing TDB-V.1.B and TDB-II, calculated Estimated Critical Boron Concentration and Minimum and Maximum Critical Rod Position, completed the Estimated Critical Condition Summary.

Required Materials: TDB-V.1.B, Estimated Critical Conditions Worksheet, Rev. 26.

TDB-II, Technical Data Book Reactivity Curves, Rev. 35.

TDB-VI, Core Operating Limits Report, Rev. 42 Validation Time: 37 minutes Completion Time: ________ minutes Comments:

Result: SAT UNSAT Examiner (Print / Sign): Date:

Page 1 of 5 NRC Admin JPM SA2 Rev. Final

Appendix C JPM WORKSHEET Form ES-C-1 CLASSROOM SETUP EXAMINER:

PROVIDE the examinee with a copy of:

• TDB-V.1.B, Estimated Critical Conditions Worksheet

• TDB-II, Technical Data Book Reactivity Curves

• TDB-VI, Core Operating Limits Report

• Calculator

• Straight Edge Page 2 of 5 NRC Admin JPM SA2 Rev. Final

Appendix C JPM STEPS Form ES-C-1

- Check Mark Denotes Critical Step START TIME:

Examiner Note: The following steps are from TDB-V.1.B.

Perform Step: 1 Conditions at Time of Shutdown.

A Standard: ENTERED Conditions at Time of Shutdown in TDB-V.1.B Steps A.1 to A.5.

Examiner Note: Information found on Answer Key.

Comment: SAT UNSAT Perform Step: 2 Conditions at Time of Startup.

B Standard: ENTERED Conditions at Time of Startup in TDB-V.1.B Steps B.1 to B.4.

Examiner Note: Information found on Answer Key.

Comment: SAT UNSAT Perform Step: 3 ECC Applicability.

C Standard: DETERMINED early and late date/time limits for ECC Applicability and entered data in TDB-V.1.B Steps C.1 to C.4.

Examiner Note: Information found on Answer Key.

Comment: SAT UNSAT Perform Step: 4 Reactivity Changes Due To Shutdown.

D Standard: CALCULATED and ENTERED Reactivity Changes Due To Shutdown in TDB-V.1.B Steps D.1 to D.5.

Examiner Note: Information found on Answer Key.

Comment: SAT UNSAT Page 3 of 5 NRC Admin JPM SA2 Rev. Final

Appendix C JPM STEPS Form ES-C-1 Perform Step: 5 Estimated Critical Boron Concentration.

E Standard: CALCULATED and ENTERED Estimated Critical Boron Concentration in TDB-V.1.B Steps E.1 to E.3.

Step E.3.d: Calculated 884 +/- 25 ppm (Critical)

Examiner Note: Information found on Answer Key.

Comment: SAT UNSAT Perform Step: 6 Minimum and Maximum Critical Rod Position.

F Standard: CALCULATED and ENTERED Minimum and Maximum Critical Rod Position in TDB-V.1.B Steps F.1 to F.4.

Step F.4: Estimated Critical Condition Summary:

• Minimum critical position: Group 3 at 75 (+/- 5) inches (Critical)

• Estimated critical position: Group 4 at 78 (+/- 5) inches (Critical)

• Maximum critical position: Group 4 at 126 (- 5) inches (Critical)

Examiner Note: Information found on Answer Key.

Terminating Cue: This JPM is complete.

Comment: SAT UNSAT STOP TIME:

Page 4 of 5 NRC Admin JPM SA2 Rev. Final

Appendix C JPM CUE SHEET Form ES-C-1 INITIAL CONDITIONS: Given the following conditions:

• A Reactor Trip from 100% power occurred on 12/03/15 at 0400 following 6 months of full power operation.

• All Control Rods were fully withdrawn at the time of the trip.

• Boron concentration prior to the trip was 400 ppm.

• Average Core Burnup is 6 GWD/MTU.

• Current boron concentration is 500 ppm.

• Criticality is scheduled to occur with Regulating Group 4 at 78 inches.

• Reactor Engineering reports no correction is needed for boron depletion.

• Reactor Startup is scheduled for 12/10/15 at 0600.

INITIATING CUE: The Shift Manager directs you to PERFORM the following:

• CALCULATE estimated critical boron concentration per TDB-V.1.B, Estimated Critical Conditions Worksheet.

• COMPLETE data entry through TDB-V.1.B, Step F.4, Estimated Critical Condition Summary.

Page 5 of 5 NRC Admin JPM SA2 Rev. Final

Appendix C JPM WORKSHEET Form ES-C-1 Facility: FCS JPM # NRC SA3 Task # 1260 K/A # 2.2.40 3.4 / 4.7

Title:

Determine In-Core Instrumentation Operability Examinee (Print):

Testing Method:

Simulated Performance: Classroom: X Actual Performance: X Simulator:

Alternate Path: Plant:

Time Critical:

READ TO THE EXAMINEE I will explain the Initial Conditions, which steps to simulate or discuss, and provide an Initiating Cue.

When you complete the task successfully, the objective for this JPM will be satisfied.

Initial Conditions: Given the following conditions:

• Plant is at 100% power.

• An In-Core Detector Status Map was just completed for Cycle 28.

Initiating Cue: The Shift Manager directs you to PERFORM the following:

• EVALUATE the In-Core detector system indications and detector status per OI-NI-2, In-Core Instrumentation Operability Requirements.

• In-Core Instrumentation System OPERABILITY per OI-NI-2 (CIRCLE): YES / NO

• IDENTIFY required actions, if any:

Task Standard: Utilizing OI-NI-2, evaluated In-Core Instrumentation Map and determined that the Incore Instrument Instrument system is Operable, and Technical Specification LCO 2.10.4(1)(a)(i) & (ii) actions are required to: Apply an increase of 1% to the total uncertainties for maximum radial peaking factor (FRT) and the total peaking factor (FQT); and Increase the frequency of performing RE-ST-RX-0001 to a minimum of once every 15 days.

Required Materials: OI-NI-2, In-Core Instrumentation Operability Requirements, Rev. 9.

TDB-I.A.7.C, Core Exit Thermocouple Status, Rev. 89.

Fort Calhoun Station Technical Specifications, Amendment #283.

Validation Time: 22 minutes Completion Time: ________ minutes Comments:

Result: SAT UNSAT Examiner (Print / Sign): Date:

Page 1 of 5 NRC Admin JPM SA3 Rev. Final As Run As Run

Appendix C JPM WORKSHEET Form ES-C-1 CLASSROOM SETUP EXAMINER:

PROVIDE the examinee with a copy of:

• OI-NI-2, In-Core Instrumentation Operability Requirements.

• TDB-I.A.7.C, Core Exit Thermocouple Status.

Page 2 of 5 NRC Admin JPM SA3 Rev. Final As Run As Run

Appendix C JPM STEPS Form ES-C-1

- Check Mark Denotes Critical Step START TIME:

Examiner Note: The following steps are from OI-NI-2, Attachment 1.

Perform Step: 1 WHEN either of the following conditions are met, THEN the In-core 1, 1.a, 1.b, & 1.b.1) Detector System is considered operable:

• At least 75% of all In-core Detector Strings are operable and at least two In-core Detector Strings are operable per full Axial Quadrant.

• Between 28% and 75% of all In-core Detector Strings are operable and:

• At least two In-core Detector Strings are operable per full Axial Quadrant Standard: REVIEWED TDB-I.A.&.C, Core Exit Thermocouple Status map and PERFORMED the following:

• DETERMINED 8 of 28 In-Core Detector Strings are inoperable (71.4%), which is between 28% and 75%

• DETERMINED at least two In-core Detector Strings are OPERABLE per full Axial Quadrant.

• CIRCLED YES.

Comment: SAT UNSAT Page 3 of 5 NRC Admin JPM SA3 Rev. Final As Run As Run

Appendix C JPM STEPS Form ES-C-1 Perform Step: 2 WHEN either of the following conditions are met, THEN the In-core 1, 1.b, 1.b.2), & 1.b.3) Detector System is considered operable:

• Between 28% and 75% of all In-core Detector Strings are operable and:

• An increase of 1% to the total uncertainties shall be applied to the maximum radial peaking factor (FRT) and the total peaking factor (FqT) and

• The frequency of performing RE-ST-RX-0001 is changed to a minimum of once every 15 days.

Standard: IDENTIFIED Required Actions and RECORDED the following:

• Apply an increase of 1% to the total uncertainties for maximum radial peaking factor (FRT) and the total peaking factor (FQT), and

• Increase the frequency of performing RE-ST-RX-0001 to a minimum of once every 15 days.

Terminating Cue: This JPM is complete.

Comment: SAT UNSAT STOP TIME:

Page 4 of 5 NRC Admin JPM SA3 Rev. Final As Run As Run

Appendix C JPM CUE SHEET Form ES-C-1 INITIAL CONDITIONS: Given the following conditions:

• Plant is at 100% power.

• An In-Core Detector Status Map, was just completed for Cycle 28.

INITIATING CUE: The Shift Manager directs you to PERFORM the following:

• EVALUATE the In-Core detector system indications and detector status per OI-NI-2, In-Core Instrumentation Operability Requirements.

• In-Core Instrumentation System OPERABILITY per OI-NI-2 (CIRCLE): YES / NO

• IDENTIFY required actions, if any:

Page 5 of 5 NRC Admin JPM SA3 Rev. Final As Run As Run

Appendix C JPM WORKSHEET Form ES-C-1 Facility: FCS JPM # NRC SA4 Task # 0741 K/A # 2.3.7 3.5 / 3.8

Title:

Authorize a Liquid Waste Release Examinee (Print):

Testing Method:

Simulated Performance: Classroom: X Actual Performance: X Simulator:

Alternate Path: Plant:

Time Critical:

READ TO THE EXAMINEE I will explain the Initial Conditions, which steps to simulate or discuss, and provide an Initiating Cue.

When you complete the task successfully, the objective for this JPM will be satisfied.

Initial Conditions: Given the following conditions:

• Raw Water Pumps AC-10A, AC-10B, and AC-10C are operating.

• Monitor Tank A is to be released and was placed on recirculation four hours ago.

• The permit has just been received in the Control Room to release Monitor Tank A.

Initiating Cue: The Shift Manager directs you to PERFORM the following:

• REVIEW the Liquid Release Permit and Plant Conditions and CIRCLE the results:

• Correct Tank is being DISCHARGED? YES / NO

• Maximum Allowable Flow DETERMINED? YES / NO

• Unloader Flow Rate SATISFACTORY? YES / NO

• Dilution Pump alignment SATISFACTORY? YES / NO Task Standard: Utilizing FC-211, determined Monitor Tank A is being discharged, Maximum Allowable Flow Rate identified as 140 gpm, Unloader Flow Rate is unsatisfactory and improper Pump alignment is in service.

Required Materials: FC-211, Waste Liquid Tank Release Permit, Rev. 25.

Validation Time: 13 minutes Completion Time: ________ minutes Comments:

Result: SAT UNSAT Examiner (Print / Sign): Date:

Page 1 of 4 NRC Admin JPM SA4 Rev. Final

Appendix C JPM WORKSHEET Form ES-C-1 CLASSROOM SETUP EXAMINER:

PROVIDE the examinee with a copy of:

• FC-211, Waste Liquid Tank Release Permit.

Page 2 of 4 NRC Admin JPM SA4 Rev. Final

Appendix C JPM STEPS Form ES-C-1

- Check Mark Denotes Critical Step START TIME:

Examiner Note: The following steps are from FC-211.

Perform Step: 1 Is the correct Tank being discharged?

Standard: DETERMINED Monitor Tank A is being released and CIRCLED YES.

Comment: SAT UNSAT Examiner Note: FC-211, Step IV, Maximum Release Rate Calculations - Set Unloader to a flow rate of 130 gpm which is less than or equal to 90% of the maximum release rate listed in Part IV.

Perform Step: 2 Is the maximum allowable flow rate determined?

Standard: DETERMINED Maximum Allowable Flow Rate set at 140 gpm and CIRCLED YES.

Comment: SAT UNSAT Examiner Note: FC-211, Step VII, Special Instructions, Item B. - Set Unloader to a flow rate of 130 gpm which is less than or equal to 90% of the maximum release rate listed in Part IV.

Perform Step: 3 Is the Unloader Flow Rate satisfactory?

Standard: DETERMINED Unloader Flow Rate is NOT satisfactory and CIRCLED NO. (Unloader Flow Rate should be 126 gpm (90% of 140 gpm).)

Comment: SAT UNSAT Examiner Note: FC-211, Step VII, Special Instructions, Item C. - Maintain 2 Circulating Water Pumps in operation.

Perform Step: 4 Is the Pump alignment satisfactory?

Standard: DETERMINED Pump alignment is NOT satisfactory and CIRCLED NO.

(3 Raw Water Pumps are operating and Step VII, Item C calls for 2 Circulating Water Pumps.)

Terminating Cue: This JPM is complete.

Comment: SAT UNSAT STOP TIME:

Page 3 of 4 NRC Admin JPM SA4 Rev. Final

Appendix C JPM CUE SHEET Form ES-C-1 INITIAL CONDITIONS: Given the following conditions:

• Raw Water Pumps AC-10A, AC-10B, and AC-10C are operating.

• Monitor Tank A is to be released and was placed on recirculation four hours ago.

• The permit has just been received in the Control Room to release Monitor Tank A.

INITIATING CUE: The Shift Manager directs you to PERFORM the following:

• REVIEW the Liquid Release Permit and Plant Conditions and CIRCLE the results:

• Correct Tank is being DISCHARGED? YES / NO

• Maximum Allowable Flow DETERMINED? YES / NO

• Unloader Flow Rate SATISFACTORY? YES / NO

• Dilution Pump alignment SATISFACTORY? YES / NO Page 4 of 4 NRC Admin JPM SA4 Rev. Final

Appendix C JPM WORKSHEET Form ES-C-1 Facility: FCS JPM # NRC SA5 Task # 1453 K/A # 2.4.41 2.9 / 4.6

Title:

Classify an Emergency Plan Event Examinee (Print):

Testing Method:

Simulated Performance: Classroom: X Actual Performance: X Simulator:

Alternate Path: Plant:

Time Critical:

READ TO THE EXAMINEE I will explain the Initial Conditions, which steps to simulate or discuss, and provide an Initiating Cue.

When you complete the task successfully, the objective for this JPM will be satisfied.

Initial Conditions: Given the following conditions:

• The Site has received an aircraft attack notification from the Nuclear Regulatory Commission Headquarters Operations Officer that was confirmed by Security at 1200.

• Actions of AOP-37, Security Events, are being implemented.

• Aircraft arrival time is estimated at 60 minutes.

Initiating Cue: The Shift Manager directs you to PERFORM the following:

• DETERMINE the Recognition Category and Event Classification per EPIP-OSC-1, Emergency Plan.

• Emergency Classification Level __________

• IC/EAL Classification __________

Task Standard: Utilizing EPIP-OSC-1 and TDB-EPIP-OSC-1H, determined Recognition Category and classified the event as a Notification of Unusual Event Category HU4.

Required Materials: EPIP-OSC-1, Emergency Plan, Rev. 48b.

TDB-EPIP-OSC-1H, Recognition Category H - Hazards and Other Conditions Affecting Plant Safety, Rev. 3.

Validation Time: 5 minutes Completion Time: ________ minutes Comments:

Result: SAT UNSAT Examiner (Print / Sign): Date:

Page 1 of 4 NRC Admin JPM SA5 Rev. Final

Appendix C JPM WORKSHEET Form ES-C-1 CLASSROOM SETUP EXAMINER:

PROVIDE the examinee with a copy of:

• EPIP-OSC-1, Emergency Plan.

• EPIP-OSC-1A/1C/1E/1F/1H/1S, EPIP Recognition Category Basis Documents.

• FC-EPF-42, Emergency Planning Form NOTE:

PROVIDE the entire EPIP-OSC-1A/1C/1E/1F/1H/1S, EPIP Recognition Category Basis Documents.

Page 2 of 4 NRC Admin JPM SA5 Rev. Final

Appendix C JPM STEPS Form ES-C-1

- Check Mark Denotes Critical Step START TIME:

Examiner Note: The following steps are from Fort Calhoun Station Emergency Action Levels.

Examiner Note: The Applicant may reference TDB-EPIP-OSC-1H which is the EPIP Bases document for HAZARDS.

Perform Step: 1 DETERMINE the Event Category.

Standard: REFERRED to FCS Emergency Action Levels:

• Figure 8.1, Recognition Categories That Apply to Operating Modes Greater Than OR Equal to 210°F.

• Figure 8.1, Recognition Categories That Apply to Operating Modes Less Than to 210°F.

Comment: SAT UNSAT Perform Step: 2 MATCH plant conditions in the Recognition Category.

Standard: IDENTIFIED EAL Recognition Category H - Hazards and Other Conditions Affecting Plant Safety.

Comment: SAT UNSAT Perform Step: 3 Declare the event emergency level.

Standard: IDENTIFIED Emergency level - NOUE (Notification of Unusual Event)

Comment: SAT UNSAT Perform Step: 4 Classify the event.

Standard: CLASSIFIED the event as a NOTIFICATION OF UNUSUAL EVENT (HU4), EAL 3. Confirmed SECURITY CONDITION or threat which indicates a potential degradation in the level of safety of the plant, EAL

1. 3: A validated notification from NRC providing information of an aircraft threat.

Terminating Cue: This JPM is complete.

Comment: SAT UNSAT STOP TIME:

Page 3 of 4 NRC Admin JPM SA5 Rev. Final

Appendix C JPM CUE SHEET Form ES-C-1 INITIAL CONDITIONS: Given the following conditions:

• The Site has received an aircraft attack notification from the Nuclear Regulatory Commission Headquarters Operations Officer that was confirmed by Security at 1200.

• Actions of AOP-37, Security Events, are being implemented.

• Aircraft arrival time is estimated at 60 minutes.

INITIATING CUE: The Shift Manager directs you to PERFORM the following:

• DETERMINE the Recognition Category and Event Classification per EPIP-OSC-1, Emergency Plan.

• Emergency Classification Level __________

• IC/EAL Classification __________

Page 4 of 4 NRC Admin JPM SA5 Rev. Final

Appendix C JPM WORKSHEET Form ES-C-1 Facility: FCS JPM # NRC S-1 Task # 0675 K/A # 001.A2.11 4.4 / 4.7 SF-1

Title:

Perform Control Element Assembly Exercises Examinee (Print):

Testing Method:

Simulated Performance: Classroom:

Actual Performance: X Simulator: X Alternate Path: X Plant:

Time Critical:

READ TO THE EXAMINEE I will explain the Initial Conditions, which steps to simulate or discuss, and provide an Initiating Cue.

When you complete the task successfully, the objective for this JPM will be satisfied.

Initial Conditions: Given the following conditions:

• Maintenance on Shutdown Group A was just completed.

• A partial movement check of Shutdown Group A is required.

• CEAs are in an All-Rods-Out configuration.

Initiating Cue: The Control Room Supervisor directs you to PERFORM the following:

• COMPLETE Control Element Assembly exercise on Shutdown Group A per OP-ST-CEA-0003, Control Element Assembly Partial Movement Check.

Task Standard: Utilizing OP-ST-CEA-0003, exercised Shutdown Group A CEAs then tripped the Reactor when 2 CEAs dropped by opening CEDM Clutch Power Supply Breakers.

Required Materials: OP-ST-CEA-0003, Control Element Assembly Partial Movement Check, Rev. 14.

EOP-00, Standard Post Trip Actions, Rev. 33.

Validation Time: 20 minutes Completion Time: ________ minutes Comments:

Result: SAT UNSAT Examiner (Print / Sign): Date:

Page 1 of 7 NRC JPM S-1 Rev. Final

Appendix C JPM WORKSHEET Form ES-C-1 SIMULATOR SETUP BOOTH OPERATOR:

INITIALIZE to IC-112:

• ENSURE DCS Computer Screen set at CEA ALL.

Type Item Value Condition Event ATWAS_PLUS MALF/CRD ROD_PWR_A30_1 (Rod 30 DE-ENERGIZED When second rod motion is clutch failure) performed MALF/CRD ROD_PWR_A33_1 (Rod 33 DE-ENERGIZED When second rod motion is clutch failure) performed BOOTH OPERATOR NOTE:

• After each JPM, VERIFY all control switches and reactor trip pushbutton cover is restored to normal condition prior to performance by the next examinee.

EXAMINER:

PROVIDE the examinee with a copy of:

• OP-ST-CEA-0003, Control Element Assembly Partial Movement Check.

• INITIALED through Prerequisites.

• N/A all CEAs from 14 to 1 on Attachment 1.

Page 2 of 7 NRC JPM S-1 Rev. Final

Appendix C JPM STEPS Form ES-C-1

- Check Mark Denotes Critical Step START TIME:

Examiner Note: The following steps are from OP-ST-CEA-0003.

NOTE Step 7.1 can be performed at anytime and repeated as necessary.

Perform Step: 1 IF this Surveillance Test is turned over, a prejob briefing must be 7.1 conducted prior to the continuation of this test.

Standard: ACKNOWLEDGED a pre-job brief is required prior to continuing.

Comment: SAT UNSAT CAUTION When the Reactor is critical, this Surveillance Test must be performed within the specified Technical Specification time interval regardless of rod configuration or use.

Perform Step: 2 IF not in an All-Rods-Out configuration, THEN contact the Reactor 7.2 Engineer prior to commencing this test for guidance to ensure the requirements of Technical Specification 3.2, Table 3-5, Item 2 are met.

Standard: DETERMINED CEAs in an All-Rods-Out configuration per Initial Conditions.

Comment: SAT UNSAT Perform Step: 3 Record Initial Position of all CEAs on Attachment 1.

7.3 Standard

RECORDED Initial Position of Shutdown Group A CEAs #30, #31, #32,

1. 33, #34, #35, #36, and #37 on Attachment 1.

Comment: SAT UNSAT Perform Step: 4 Rotate the Mode Selector Switch (M/M) to the Manual Individual (M/I) 7.4 position.

Standard: ROTATED Mode Selector Switch (M/M) to Manual Individual (M/I) position.

Comment: SAT UNSAT Page 3 of 7 NRC JPM S-1 Rev. Final

Appendix C JPM STEPS Form ES-C-1 Perform Step: 5 Rotate the Group Selector Switch (M/G) to the Group containing the 7.5 CEA to be moved.

Standard: ROTATED Group Selector Switch (M/G) to Shutdown Group A.

Comment: SAT UNSAT Perform Step: 6 If available, verify on SCEAPIS (DCS) display CEA_ALL that the group 7.6 button is DARK GREY for the group selected.

Standard: VERIFIED on Secondary Control Element Assembly Position Indicating System (SCEAPIS) Digital Control System (DCS) display CEA_ALL that Shutdown Group A button is DARK GREY.

Comment: SAT UNSAT Examiner Cue: If questioned, REPORT the CRS directs you to start with CEA #30.

Perform Step: 7 Rotate the Rod Selector Switch to the CEA to be moved.

7.7 Standard

ROTATED Rod Selector Switch to any Shutdown Group A CEA.

Comment: SAT UNSAT NOTE If Group 4 CEAs are being used for ASI control, movement of 6 inches in a single direction may be credited. The returned to position may not necessarily be the initial position. Note the time of Group 4 movement for ASI control on the Comment Sheet if applicable.

Examiner Note: When the 2nd CEA is exercised, two CEAs will drop into the Core.

Perform Step: 8 Insert or withdraw the CEA, as applicable, a minimum of six (6) inches, 7.8 THEN return the CEA to its Initial Position.

Standard: INSERTED CEA a minimum of six (6) inches, MONITORED Nuclear Instrumentation and TAVE then WITHDREW CEA to its Initial Position.

Examiner Cue: If the ROD DRIVE POWER INTERRUPT alarm is received (in the event the CEA is moved 8 inches), REPORT as CRS that permission is granted to use the Rod Block Bypass Switch to move the CEA back to its original position.

Examiner Note: Candidate may reposition the Manual Mode Selector switch to off to respond to alarms. If so, the candidate must return the switch to MI to move the selected rod.

Comment: SAT UNSAT Page 4 of 7 NRC JPM S-1 Rev. Final

Appendix C JPM STEPS Form ES-C-1 NOTE Step 7.9 may be completed after all CEAs within a Group have been exercised, after all CEAs have been exercised, OR after exercising each CEA.

Examiner Note: When the 2nd CEA is exercised, two CEAs will drop into the Core.

Perform Step: 9 Record Inserted/Withdrawn To AND Return To information on Test Data Sheet, THEN initial Attachment 1.

Standard: Record Inserted/Withdrawn To AND Return To information on Test Data Sheet, THEN initial Attachment 1.

Comment: SAT UNSAT Examiner Note: The following steps represent the Alternate Path of this JPM.

Perform Step: 10 Determine that 2 CEAs have dropped into the core.

Standard: OBSERVED Annunciator Alarms and DETERMINED two CEAs have dropped.

Comment: SAT UNSAT Examiner Note: The following steps are from EOP-00, Standard Post Trip Actions.

Perform Step: 11 Verify Reactivity Control is established by performing steps a and b:

1

• Verify ALL of the following:

• No more than one Regulating or Shutdown CEA is NOT inserted

• Reactor power is lowering

• Startup rate is negative Standard: DETERMINED Reactor did NOT trip when both CEAs dropped and REFERRED to CONTINGENCY ACTIONS (CA).

Comment: SAT UNSAT Perform Step: 12a IF the reactor did NOT trip, THEN establish Reactivity Control by 1.1 & 1.1.a CA performing step a, b, c or d:

• Manually trip the Reactor (CB-4).

Standard: DEPRESSED REACTOR TRIP pushbutton on CB-4 and DETERMINED Reactor did NOT trip.

Comment: SAT UNSAT Page 5 of 7 NRC JPM S-1 Rev. Final

Appendix C JPM STEPS Form ES-C-1 Perform Step: 12b IF the reactor did NOT trip, THEN establish Reactivity Control by 1.1 & 1.1.b CA performing step a, b, c or d:

• Manually trip the Reactor (AI-31).

Standard: DEPRESSED REACTOR TRIP pushbutton on AI-31 and DETERMINED Reactor did NOT trip.

Comment: SAT UNSAT Perform Step: 12c IF the reactor did NOT trip, THEN establish Reactivity Control by 1.1 & 1.1.c CA performing step a, b, c or d:

• Place the DSS Manual Trip Switches in "TRIP" (AI-66A/B).

Standard: PERFORMED the following:

• PLACED DSS Manual Trip Switch in TRIP position on AI-66A and DETERMINED Reactor did NOT trip.

• PLACED DSS Manual Trip Switch in TRIP position on AI-66B and DETERMINED Reactor did NOT trip.

Comment: SAT UNSAT Perform Step: 12d IF the reactor did NOT trip, THEN establish Reactivity Control by 1.1 & 1.1.d CA performing step a, b, c or d:

• Manually open the CEDM Clutch Power Supply Breakers (AI-57).

Standard: PERFORMED the following:

• OPENED both CLUTCH POWER SUPPLY BREAKER RPS/

CB-A/B on AI-57 (critical).

• OPENED both CLUTCH POWER SUPPLY BREAKER RPS/

CB-C/D on AI-57 (critical)

• OBSERVED all Rod Bottom lights LIT on SCEAPIS (NOT critical).

Terminating Cue: This JPM is complete.

Comment: SAT UNSAT STOP TIME:

Page 6 of 7 NRC JPM S-1 Rev. Final

Appendix C JPM CUE SHEET Form ES-C-1 INITIAL CONDITIONS: Given the following conditions:

• Maintenance on Shutdown Group A was just completed.

• A partial movement check of Shutdown Group A is required.

• CEAs are in an All-Rods-Out configuration.

INITIATING CUE: The Control Room Supervisor directs you to PERFORM the following:

• COMPLETE Control Element Assembly exercise on Shutdown Group A per OP-ST-CEA-0003, Control Element Assembly Partial Movement Check.

Page 7 of 7 NRC JPM S-1 Rev. Final

Appendix C JPM WORKSHEET Form ES-C-1 Facility: FCS JPM # NRC S-2 Task # 1391 K/A # 004.A4.08 3.8 / 3.4 SF-2

Title:

Align Charging Flow Via the HPSI Header Examinee (Print):

Testing Method:

Simulated Performance: Classroom:

Actual Performance: X Simulator: X Alternate Path: Plant:

Time Critical:

READ TO THE EXAMINEE I will explain the Initial Conditions, which steps to simulate or discuss, and provide an Initiating Cue.

When you complete the task successfully, the objective for this JPM will be satisfied.

Initial Conditions: Given the following conditions:

• AOP-33, CVCS Leak, is in progress.

• Leak isolation has restored Charging Pump CH-1C.

• AOP-33, Step 13.d, directs use of Attachment C.

Initiating Cue: The Control Room Supervisor directs you to PERFORM the following:

• ALIGN Charging flow via the High Pressure Safety Injection (HPSI) header per AOP-33, CVCS Leak, Attachment C, Charging Via the HPSI Header Using Only CH-1C.

Task Standard: Utilizing AOP-33, opened HCV-308, opened HCV-312, and started Charging Pump CH-1C to restore Pressurizer level.

Required Materials: AOP-33, CVCS Leak, Rev. 9.

Validation Time: 13 minutes Completion Time: ________ minutes Comments:

Result: SAT UNSAT Examiner (Print / Sign): Date:

Page 1 of 6 NRC JPM S-2 Rev. Final

Appendix C JPM WORKSHEET Form ES-C-1 SIMULATOR SETUP BOOTH OPERATOR:

INITIALIZE to IC-112:

• VERIFY Pressurizer level is lowered to 55%.

• ENSURE all Charging Loop Isolation Valves CLOSED per AOP-33.

• ENSURE all Auxiliary Spray Valves are CLOSED per AOP-33.

• ENSURE all Charging Pumps in PULL-TO-LOCK per AOP-33.

Type Item Value Condition Remote/CVC REM:CVC_CH172 0 Remote/CVC REM:CVC_CH173 0 Remote/CVC REM:CVC_CH191 0 Remote/CVC REM:CVC_CH192 0 Remote/CVC REM:CVC_CH193 0 Remote/CVC REM:CVC_CH194 0 EXAMINER:

PROVIDE the examinee with a copy of:

• AOP-33, CVCS Leak.

• Attachment C, Charging Via the HPSI Header Using Only CH-1C.

Page 2 of 6 NRC JPM S-2 Rev. Final

Appendix C JPM STEPS Form ES-C-1

- Check Mark Denotes Critical Step START TIME:

Examiner Note: The following steps are from AOP-33, Attachment C.

NOTE Charging flow can be verified on the associated HPSI flow indicator(s) for the HPSI Loop Valve(s) in use, or on ERF (Page 323).

Perform Step: 1 Ensure all Charging Pumps are in "PULL-TO-LOCK".

1 Standard: DETERMINED all Charging Pumps in PULL-TO-LOCK.

Comment: SAT UNSAT Perform Step: 2 Unlock and close BOTH of the following valves:

2 & all bullets

• CH-194, "CHARGING PUMPS CH-1A/B/C DISCHARGE HEADER CONTAINMENT OUTBOARD ISOLATION VALVE" (Room 13)

• CH-191, "CHARGING PUMPS CH-1A & B DISCHARGE HEADER TO SAFETY INJECTION ISOLATION VLV." (Charging Pump Valve Room)

Standard: CONTACTED Auxiliary Operator to UNLOCK and CLOSE CH-194 in Room 13 and CH-191 in Charging Pump Valve Room.

Booth Operator: When contacted, UNLOCK and CLOSE CH-194 and CH-191.

REPORT CH-194 and CH-191 UNLOCKED and CLOSED.

Comment: SAT UNSAT Perform Step: 3 Close ALL of the following valves:

3 & all bullets

• CH-192, "CHARGING PUMP CH-1B DISCHARGE VALVE" (Charging Pump Valve Room)

• CH-173, "CHARGING PUMP CH-1B SUCTION VALVE" (Charging Pump Valve Room)

• CH-193, "CHARGING PUMP CH-1A DISCHARGE VALVE" (Charging Pump Valve Room)

• CH-172, "CHARGING PUMP CH-1A SUCTION VALVE" (Charging Pump Valve Room)

Standard: CONTACTED Auxiliary Operator to UNLOCK and CLOSE CH-192, CH-173, CH-193, and CH-172 in Charging Pump Valve Room.

Booth Operator: When contacted, CLOSE CH-192, CH-173, CH-193, and CH-172.

REPORT CH-192, CH-173, CH-193, and CH-172 CLOSED.

Comment: SAT UNSAT Page 3 of 6 NRC JPM S-2 Rev. Final

Appendix C JPM STEPS Form ES-C-1 Perform Step: 4 Open HCV-308, Charging Pump HPSI Header Isolation Valve.

4 Standard: PERFORMED the following:

• PLACED HCV-308, CHARGING PUMP/HPSI HDR ISOLATION VALVE in OPEN (critical).

• OBSERVED red OPEN light lit (NOT critical).

Comment: SAT UNSAT Perform Step: 5 Ensure ALL of the following valves are open:

5 & all bullets

• HCV-2987, HPSI Header Isolation Valve

• HCV-307, HPSI Header Isolation Valve

• HCV-305, SI-2A and SI-2C Discharge Cross-Connect Valve

• HCV-304, SI-2B and SI-2C Discharge Cross-Connect Valve

• HCV-306, HPSI Header Isolation Valve Standard: VERIFIED all of the following valves OPEN and red OPEN lights lit:

• HCV-2987, HPSI HEADER NUMBER 2 DISCHARGE VALVE

• HCV-307, HPSI HEADER NUMBER 2 DISCHARGE VALVE

• HCV-305, HPSI PUMPS SI-2A/SI-2C DISCH CROSSCONNECT VLV

• HCV-304, HPSI PUMPS SI-2B/SI-2C DISCH CROSSCONNECT VLV

• HCV-306, HPSI HEADER NUMBER 1 DISCHARGE VALVE Comment: SAT UNSAT Examiner Note: HCV-312 was selected for consistency of Applicants.

Examiner Cue: The CRS directs you to open HCV-312, HPSI Loop Injection Valve.

Perform Step: 6a Open at least ONE of the following HPSI Loop Injection Valves:

6, 6.a, & 6.a.1)

• Open HCV-312 (Loop 1B) by performing the following:

• Rotate thumbwheel for PCV-2909, "LEAKAGE CLR SI-4A DISCH VLV CNTRLR" fully clockwise to close "C".

Standard: PERFORMED the following:

• VERIFIED thumbwheel for PCV-2909, LEAKAGE CLR SI-4A DISCHARGE VALVE CONTROLLER fully CLOCKWISE in CLOSE (C position).

• OBSERVED needle in C position.

Comment: SAT UNSAT Page 4 of 6 NRC JPM S-2 Rev. Final

Appendix C JPM STEPS Form ES-C-1 Perform Step: 6b Open at least ONE of the following HPSI Loop Injection Valves:

6, 6.a, & 6.a.2)

• Open HCV-312 (Loop 1B) by performing the following:

• Place PCV-2909, "LEAKAGE CLR SI-4A DISCHARGE VALVE" in "MANUAL".

Standard: PERFORMED the following:

• PLACED PCV-2909, LEAKAGE CLR SI-4A DISCHARGE VALVE to MANUAL position (critical).

• OBSERVED switch in MANUAL (NOT critical).

• OBSERVED amber light off and green light lit (NOT critical).

Comment: SAT UNSAT Perform Step: 6c Open at least ONE of the following HPSI Loop Injection Valves:

6, 6.a, & 6.a.3)

• Open HCV-312 (Loop 1B) by performing the following:

• Open HCV-312, "LOOP 1B HPSI INJECTION VALVE".

Standard: PERFORMED the following:

• TURNED and HELD HCV-312, LOOP 1B HPSI INJECTION VALVE to OPEN position (critical).

• OBSERVED red OPEN light lit (NOT critical).

Comment: SAT UNSAT NOTE Charging flow can be verified on the associated HPSI flow indicator(s) for the HPSI Loop Valve(s) in use, or on ERF (Page 323).

Perform Step: 7 Operate CH-1C as necessary to maintain PZR level within 4% of 7 programmed level.

Standard: PERFORMED the following:

• PLACED CH-1C, CHRG PUMP in START (critical).

• OBSERVED red START light lit (NOT critical).

• OBSERVED flow on HCV-312, HPSI Loop Injection Valve or on ERF Computer Page 323 (NOT critical).

• OBSERVED ~75 amps on CH-1C ammeter (NOT critical).

Terminating Cue: This JPM is complete.

Comment: SAT UNSAT STOP TIME:

Page 5 of 6 NRC JPM S-2 Rev. Final

Appendix C JPM CUE SHEET Form ES-C-1 INITIAL CONDITIONS: Given the following conditions:

• AOP-33, CVCS Leak, is in progress.

• Leak isolation has restored Charging Pump CH-1C.

• AOP-33, Step 13.d, directs use of Attachment C.

INITIATING CUE: The Control Room Supervisor directs you to PERFORM the following:

• ALIGN Charging flow via the High Pressure Safety Injection (HPSI) header per AOP-33, CVCS Leak, Attachment C, Charging Via the HPSI Header Using Only CH-1C.

Page 6 of 6 NRC JPM S-2 Rev. Final

Appendix C JPM WORKSHEET Form ES-C-1 Facility: FCS JPM # NRC S-3 Task # 1129 K/A # 009.EA2.34 3.6 / 4.2 SF-3

Title:

Perform HPSI Stop and Throttle Examinee (Print):

Testing Method:

Simulated Performance: Classroom:

Actual Performance: X Simulator: X Alternate Path: X Plant:

Time Critical:

READ TO THE EXAMINEE I will explain the Initial Conditions, which steps to simulate or discuss, and provide an Initiating Cue.

When you complete the task successfully, the objective for this JPM will be satisfied.

Initial Conditions: Given the following conditions:

• A Small Break Loss of Coolant Accident is in progress.

• EOP-03, Loss of Coolant Accident, has been entered.

Initiating Cue: The Control Room Supervisor directs you to PERFORM the following:

• EVALUATE then EXECUTE actions for HPSI Stop and Throttle per EOP/AOP Floating Step F, HPSI Stop and Throttle Criteria.

Task Standard: Utilizing Floating Step F, stopped all but one HPSI Pump and throttled Loop Injection Valves. Upon leak increase, restarted HPSI Pumps and opened Loop Injection Valves as required.

Required Materials: EOP/AOP Floating Steps, Rev. 7.

Validation Time: 10 minutes Completion Time: ________ minutes Comments:

Result: SAT UNSAT Examiner (Print / Sign): Date:

Page 1 of 6 NRC JPM S-3 Rev. Final As Run As Run

Appendix C JPM WORKSHEET Form ES-C-1 SIMULATOR SETUP BOOTH OPERATOR:

INITIALIZE to IC-118:

Type Item Value Condition MALF/RCS RCS01B (RCS-01B - RCS Loop 0.43 Recall/modify after flow Leak - Loop 1B Cold Leg ** Leak is present and level are balanced, Medium) when you restore increase leak rate to value this IC 1.5 EXAMINER:

PROVIDE the examinee with a copy of:

• EOP/AOP Floating Step F, HPSI Stop and Throttle Criteria.

Page 2 of 6 NRC JPM S-3 Rev. Final As Run As Run

Appendix C JPM STEPS Form ES-C-1

- Check Mark Denotes Critical Step START TIME:

Examiner Note: The following steps are from Floating Step FS-F.

CAUTIONS

1. If emergency boration is required then at least one charging pump must remain running.

2. As natural circulation develops, the expected rise in TH will reduce subcooling. This may jeopardize HPSI Stop and Throttle Criteria.

3. Reducing SI flow should be approached cautiously.

4. The purpose of HPSI stop and throttle is to prevent an over pressurization of the RCS and a solid PZR, however, maintaining RCS inventory is more important than pressure control.

Perform Step: 1 Verify ALL of the following stop and throttle criteria are satisfied:

1 & all bullets

• RCS subcooling is greater than or equal to 20°F

• PZR level is greater than or equal to 10% and not lowering

• At least one S/G is available for RCS heat removal

• RVLMS indicates level is at or above the top of the Hot Leg (43%, ERF "I" display)

Standard: OBSERVED the following:

• RCS subcooling greater than 20°F.

• Pressurizer level greater than 10% and not lowering.

• Both Steam Generators are available for RCS heat removal.

• Reactor Vessel Level Monitoring System is greater than 43%.

Comment: SAT UNSAT Examiner Note: Applicant must place any 2 of 3 Charging Pumps in PULL-TO-LOCK otherwise they will AUTO START.

Perform Step: 2 Ensure only ONE Charging Pump is operating.

2 Standard: STOPPED 2 of 3 Charging Pumps by PERFORMING the following (any 2):

• PLACED CH-1A, CHARGING PUMP in PULL-TO-LOCK (critical).

• PLACED CH-1B, CHARGING PUMP in PULL-TO-LOCK (critical).

• PLACED CH-1C, CHARGING PUMP in PULL-TO-LOCK (critical).

• OBSERVED pump breaker lights off (NOT critical).

Comment: SAT UNSAT Page 3 of 6 NRC JPM S-3 Rev. Final As Run As Run

Appendix C JPM STEPS Form ES-C-1 CAUTIONS

1. During a UHE HPSI stop and throttle should be performed before the expansion of the relatively cold SI water overfills the pressurizer.

2. Operators should closely monitor RCS pressure-temperature limits. Pressurizer spray may be required to prevent exceeding the maximum subcooling limit.

3. Allowing the RCS to repressurize to 1300 psia will effectively stop HPSI flow.

Perform Step: 3 IF a UHE is in progress, THEN maintain RCS pressure control by 3 performing the following:

Standard: DETERMINED Uncontrolled Heat Extraction is NOT in progress.

Comment: SAT UNSAT Examiner Note: It is acceptable to stop one or more HPSI Pumps and throttle HPSI Loop Injection Valves to achieve control over Pressurizer level, resulting in stable or slowly rising Pressurizer level. Applicant may stop one HPSI Pump in this step and/or throttle HPSI Loop Injection Valves to achieve this condition.

CAUTIONS

1. LOCAs pose a significant threat to RCS subcooling. Therefore, full SI Flow should be maintained until subcooled margin is stable and natural circulation has developed.

2. During a SGTR, the depressurization of the RCS to less than 1000 psia should be stopped when HPSI flow is initially being stopped and throttled.

Perform Step: 4a IF a LOCA or SGTR is in progress, THEN maintain RCS pressure 4 & 4.a control by performing the following:

• Ensure at least one HPSI Pump is operating.

Standard: DETERMINED SI-2A and SI-2B, HPSI Pumps are running. May stop one HPSI pump or leave both running.

Comment: SAT UNSAT Perform Step: 4b IF a LOCA or SGTR is in progress, THEN maintain RCS pressure 4 & 4.b control by performing the following:

• Throttle HPSI Loop Injection Valve(s).

Standard: THROTTLED CLOSE any or all the following:

• HCV-314, LOOP 1A HPSI INJECTION VALVE.

• HCV-317, LOOP 2A HPSI INJECTION VALVE.

• HCV-311, LOOP 1B HPSI INJECTION VALVE.

• HCV-321, LOOP 2B HPSI INJECTION VALVE.

Comment: SAT UNSAT Page 4 of 6 NRC JPM S-3 Rev. Final As Run As Run

Appendix C JPM STEPS Form ES-C-1 Examiner Note: The following steps represent the Alternate Path of this JPM.

Examiner Note: Once HPSI flow has been throttled, the break size will increase and Stop and Throttle criteria will no longer be met. HPSI Pumps must be restarted and/or Loop Injection Valves reopened. Applicant must recognize this and begin to take action before Reactor Vessel Level Monitoring System (RVLMS) indicates less than 100%.

Perform Step: 5a IF HPSI stop and throttle criteria can NOT be maintained, THEN raise 5 & 5.a HPSI flow by performing the following:

• Start either HPSI Pumps, SI-2A/B or SI-2B/C, as necessary.

Standard: DETERMINED SI-2A and SI-2B, HPSI Pumps are running. If HPSI pumps were secured in step 4, restart HPSI pumps. (critical if one of more HPSI pumps were secured)

Comment: SAT UNSAT Examiner Note: Applicant should throttle open valve(s) closed at Perform Step 4b.

Perform Step: 5b IF HPSI stop and throttle criteria can NOT be maintained, THEN raise HPSI flow by performing the following:

• Open HPSI Loop Injection Valves, as necessary.

Standard: OPEN any or all the following:

• HCV-314, LOOP 1A HPSI INJECTION VALVE.

• HCV-317, LOOP 2A HPSI INJECTION VALVE.

• HCV-311, LOOP 1B HPSI INJECTION VALVE.

• HCV-321, LOOP 2B HPSI INJECTION VALVE.

Terminating Cue: This JPM is complete.

Comment: SAT UNSAT STOP TIME:

Page 5 of 6 NRC JPM S-3 Rev. Final As Run As Run

Appendix C JPM CUE SHEET Form ES-C-1 INITIAL CONDITIONS: Given the following conditions:

• A Small Break Loss of Coolant Accident is in progress.

• EOP-03, Loss of Coolant Accident, has been entered.

INITIATING CUE: The Control Room Supervisor directs you to PERFORM the following:

• EVALUATE then EXECUTE actions for HPSI Stop and Throttle per EOP/AOP Floating Step F, HPSI Stop and Throttle Criteria.

Page 6 of 6 NRC JPM S-3 Rev. Final As Run As Run

Appendix C JPM WORKSHEET Form ES-C-1 Facility: FCS JPM # NRC S-4 Task # 0612 K/A # 003.A2.02 3.7 / 3.9 SF-4P

Title:

Start a Reactor Coolant Pump Examinee (Print):

Testing Method:

Simulated Performance: Classroom:

Actual Performance: X Simulator: X Alternate Path: Plant:

Time Critical:

READ TO THE EXAMINEE I will explain the Initial Conditions, which steps to simulate or discuss, and provide an Initiating Cue.

When you complete the task successfully, the objective for this JPM will be satisfied.

Initial Conditions: Given the following conditions:

• Plant Startup is in progress.

• Reactor Coolant Pumps RC-3A, RC-3B, and RC-3C are running.

Initiating Cue: The Control Room Supervisor directs you to PERFORM the following:

• START Reactor Coolant Pump RC-3D per OI-RC-9, Reactor Coolant Pump Operation, Attachment 1, Starting Reactor Coolant Pumps (Coupled).

• START at Step 11.

Task Standard: Utilizing OI-RC-9, started RC-3D-1 Oil Lift Pump and RCP RC-3D.

Required Materials: OI-RC-9, Reactor Coolant Pump Operation, Rev. 78.

Validation Time: 7 minutes Completion Time: ________ minutes Comments:

Result: SAT UNSAT Examiner (Print / Sign): Date:

Page 1 of 6 NRC JPM S-4 Rev. Final

Appendix C JPM WORKSHEET Form ES-C-1 SIMULATOR SETUP BOOTH OPERATOR:

INITIALIZE to IC-116:

• ENSURE Reactor Coolant Pump RC-3-D is STOPPED.

• ENSURE ERF Computer Page 342 or DCS "RCP Summary" on display.

EXAMINER:

PROVIDE the examinee with a copy of:

• OI-RC-9, Reactor Coolant Pump Operation.

• Attachment 1, Starting Reactor Coolant Pumps (Coupled), INITIALED through Step 10.

Page 2 of 6 NRC JPM S-4 Rev. Final

Appendix C JPM STEPS Form ES-C-1

- Check Mark Denotes Critical Step START TIME:

Examiner Note: The following steps are from OI-RC-9, Attachment 1.

NOTE The Oil Lift Pump shall not be run for longer than 10 minutes before starting the Reactor Coolant Pump.

Perform Step: 1 Announce the Reactor Coolant Pump start on the Gaitronics.

11 Standard: ANNOUNCED start of Reactor Coolant Pump RC-3D on Gaitronics.

Comment: SAT UNSAT Perform Step: 2 Place the Oil Lift Pump for the selected RCP to AFTER START:

12

• RC-3D-1, Oil Lift Pump Standard: PERFORMED the following:

• PLACED RC-3D-1, OIL LIFT PUMP handswitch in AFTER START (critical).

• OBSERVED red indicating light lit (NOT critical).

Comment: SAT UNSAT Examiner Note: ARRD is the Anti-Reverse Rotation Device on the RCP.

Perform Step: 3 Verify adequate ARRD Lube Oil Flow for the selected RCP. ERF/DCS 13 indication shall read NORMAL: (ERF page 342 or DCS "RCP Summary")

• RC-3D F3190 Standard: OBSERVED ERF Computer Page 342 or Digital Computer System RCP Summary and VERIFIED ARRD Lube Oil Flow for RC-3D is NORMAL.

Comment: SAT UNSAT Perform Step: 4 Prior to starting RCP, inform the Radiation Protection Department so it 14 can monitor changing radiological conditions.

Standard: CONTACTED Radiation Protection Department about RCP start.

Examiner Cue: Radiation Protection acknowledges start of RCP.

Comment: SAT UNSAT Page 3 of 6 NRC JPM S-4 Rev. Final

Appendix C JPM STEPS Form ES-C-1 Perform Step: 5a Startup sequence for selected RCP:

15 & 15.1

• Run the oil lift pump for the selected RCP a minimum of 2 minutes.

• RC-3D-1, Oil Lift Pump Standard: DETERMINED RC-3D-1, OIL LIFT PUMP already running.

Examiner Cue: If Applicant begins timing, REPORT two minutes have passed.

Comment: SAT UNSAT Perform Step: 5b Startup sequence for selected RCP:

15 & 15.2

• Place the selected RCP control switch in AFTER START:

• RC-3D, RC Pump Standard: PERFORMED the following:

• PLACED RC-3D, RC PUMP handswitch in AFTER START (critical).

• OBSERVED red START light lit (NOT critical).

Comment: SAT UNSAT Perform Step: 5c Startup sequence for selected RCP:

15 & 15.3

• IF the Reactor Coolant Pump motor amps fail to drop below 425 amps within the time listed below, THEN place the control switch in AFTER STOP:

• RC-3D - seventeen (17) seconds Standard: DETERMINED RC-3D ammeter reads less than 425 amps in less than 17 seconds.

Comment: SAT UNSAT Perform Step: 6a Verify the following for the selected RCP:

16 & 16.1

• Oil Lift Pump stops (Green indicating light ON):

Standard: OBSERVED RC-3D-1, OIL LIFT PUMP green indicating light lit.

Comment: SAT UNSAT Page 4 of 6 NRC JPM S-4 Rev. Final

Appendix C JPM STEPS Form ES-C-1 Perform Step: 6b Verify the following for the selected RCP:

16 & 16.2

• For the selected RCP verify the following Annunciator is clear:

• RC-3D REACTOR COOLANT PUMP RC-3D REVERSE ROTATION (CB-1/2/3, A6, D5)

Standard: OBSERVED CB-1/2/3/A6, Window D REACTOR COOLANT PUMP RC-3D REVERSE ROTATION is CLEAR.

Comment: SAT UNSAT Perform Step: 6c Verify the following for the selected RCP:

16 & 16.3

• For the selected RCP verify the following Annunciator is clear:

• RC-3D REACTOR COOLANT PUMP RC-3D VIBRATION HI (CB-1/2/3, A6, D4)

Standard: OBSERVED CB-1/2/3/A6, Window D REACTOR COOLANT PUMP RC-3D VIBRATION HI is CLEAR.

Comment: SAT UNSAT NOTE At low RCS Pressure, verification of positive Controlled Bleedoff Flow may NOT be possible.

Perform Step: 7 Verify positive Controlled Bleedoff flow for the selected RCP:

17 (ERF page 342 or DCS "RCP Summary")

• RC-3D F3175 Standard: OBSERVED ERF Computer Page 342 or Digital Computer System RCP Summary and VERIFIED Control Bleedoff Flow for RC-3D is POSITIVE.

Comment: SAT UNSAT Perform Step: 8 Monitor the ERF Computer or DCS and verify all parameters are normal 18 for the selected RCP:

• RC-3D Standard: MONITORED ERF Computer or Digital Computer System and VERIFIED RCP RC-3D parameters NORMAL.

Terminating Cue: This JPM is complete.

Comment: SAT UNSAT STOP TIME:

Page 5 of 6 NRC JPM S-4 Rev. Final

Appendix C JPM CUE SHEET Form ES-C-1 INITIAL CONDITIONS: Given the following conditions:

• Plant Startup is in progress.

• Reactor Coolant Pumps RC-3A, RC-3B, and RC-3C are running.

INITIATING CUE: The Control Room Supervisor directs you to PERFORM the following:

• START Reactor Coolant Pump RC-3D per OI-RC-9, Reactor Coolant Pump Operation, Attachment 1, Starting Reactor Coolant Pumps (Coupled).

• START at Step 11.

Page 6 of 6 NRC JPM S-4 Rev. Final

Appendix C JPM WORKSHEET Form ES-C-1 Facility: FCS JPM # NRC S-5 Task # 0033 K/A # G 2.1.19 3.9 / 3.8 SF-4S

Title:

Perform Control Room Evacuation Immediate Actions Examinee (Print):

Testing Method:

Simulated Performance: Classroom:

Actual Performance: X Simulator: X Alternate Path: X Plant:

Time Critical:

READ TO THE EXAMINEE I will explain the Initial Conditions, which steps to simulate or discuss, and provide an Initiating Cue.

When you complete the task successfully, the objective for this JPM will be satisfied.

Initial Conditions: Given the following conditions:

• A toxic gas leak is requiring evacuation of the Control Room.

Initiating Cue: The Control Room Supervisor directs you to PERFORM the following:

• EXECUTE required actions prior to a Control Room Evacuation per AOP-07, Evacuation of Control Room,Section I, Plant to Hot Shutdown.

Task Standard: Utilizing AOP-07, trip the Reactor and Turbine, secured one Main Feedwater, one Condensate, and two Heater Drain Pumps, and started Turbine Lube Oil Pumps.

Required Materials: AOP-07, Evacuation of Control Room, Rev. 17.

Validation Time: 6 minutes Completion Time: ________ minutes Comments:

Result: SAT UNSAT Examiner (Print / Sign): Date:

Page 1 of 6 NRC JPM S-5 Rev. Final

Appendix C JPM WORKSHEET Form ES-C-1 SIMULATOR SETUP BOOTH OPERATOR:

INITIALIZE to IC-111:

• RUN event No Turbine Trip from CR Edit.evt Type Item Value Condition Other THATFS_050A18O_1FREEZE 1 (FREEZE FLAG)

Other THATFS_050A28O_1FREEZE 1 (FREEZE FLAG)

Other THATFS_050A18O_2FREEZE 1 (FREEZE FLAG)

Other THATFS_050A28O_2FREEZE 1 (FREEZE FLAG)

Remote/GE REM:86-1/G1-TRP (86-1/G1 Trip Tripped P10_235SD_1 eq 1 (B EHC N Signal) pump to PTL)

Override P10_102S1_1 0 (FALSE)

Override P10_102S1_1 1 (TRUE) P10_235SD_1 eq 1 (B EHC pump to PTL)

EXAMINER:

PROVIDE the examinee with a copy of:

• AOP-07, Evacuation of Control Room.

• Section I, Plant to Hot Shutdown.

Page 2 of 6 NRC JPM S-5 Rev. Final

Appendix C JPM STEPS Form ES-C-1

- Check Mark Denotes Critical Step START TIME:

Examiner Note: The following steps are from AOP-07,Section I.

Examiner Note: When the Reactor is tripped, an uncontrolled cooldown will begin because the Turbine has NOT tripped.

Perform Step: 1 Perform the following steps prior to evacuating the Control Room:

1 & 1.a

• Manually trip the Reactor.

Standard: PERFORMED the following:

• DEPRESSED REACTOR TRIP pushbutton on CB-4 (critical).

• OBSERVED all Rods inserted, Reactor Power lowering, and Negative Startup Rate. (NOT critical).

Examiner Cue: If cooldown is addressed by applicant, REPORT as Control Room Supervisor that the ATCO will perform any required Emergency Boration. Continue with AOP-07 actions.

Comment: SAT UNSAT Perform Step: 2 Perform the following steps prior to evacuating the Control Room:

1 & 1.b

• Verify the Turbine is tripped as indicated by Stop and Intercept Valves indicating closed.

Standard: OBSERVED Stop and Intercept Valves, DETERMINED Turbine was NOT tripped, and REFERRED to CONTINGENCY ACTIONS (CA).

Comment: SAT UNSAT Examiner Note: The following steps represent the Alternate Path of this JPM.

Perform Step: 2a Trip the Turbine (CB-10, 11) b.1 & b.1.1) CA Standard: PERFORMED the following:

• DEPRESSED TURBINE ST-1 MASTER TRIP PUSHBUTTON A.

• DEPRESSED TURBINE ST-1 MASTER TRIP PUSHBUTTON B.

• OBSERVED all Stop and Intercept Valves OPEN.

Comment: SAT UNSAT Page 3 of 6 NRC JPM S-5 Rev. Final

Appendix C JPM STEPS Form ES-C-1 Perform Step: 2b Trip the Turbine by performing:

b.1 & b.1.2) CA

• Stop the EHC pumps by placing BOTH of the following control switches in "PULL-TO-LOCK":

• EHC-3A

• EHC-3B Standard: PERFORMED the following:

• PLACED EHC-3A, EHC PUMP handswitch in PULL-TO-LOCK (critical).

• PLACED EHC-3B, EHC PUMP handswitch in PULL-TO-LOCK (critical).

• OBSERVED all Stop and Intercept Valves CLOSED and DETERMINED Turbine is tripped (NOT critical).

Comment: SAT UNSAT Perform Step: 3 Perform the following steps prior to evacuating the Control Room:

1 & 1.c

• Place the "43/FW" Switch in "OFF".

Standard: PLACED 43/FW Switch in OFF.

Comment: SAT UNSAT Perform Step: 4 Perform the following steps prior to evacuating the Control Room:

1 & 1.d

• Ensure no more than one Feed Pump, FW-4A/B/C is operating.

Standard: DETERMINED only FW-4C, MFW Pump is running.

Comment: SAT UNSAT Examiner Note: Applicant can stop either Condensate Pump. Operations expectation is to stop FW-2A.

Perform Step: 5 Perform the following steps prior to evacuating the Control Room:

1 & 1.e

• Ensure no more than one Condensate Pump, FW-2A/B/C is operating.

Standard: PERFORMED one of the following:

• PLACED FW-2A, COND PUMP handswitch in STOP (critical).

OR

• PLACED FW-2C, COND PUMP handswitch in STOP (critical).

• OBSERVED green STOP light lit (NOT critical).

Comment: SAT UNSAT Page 4 of 6 NRC JPM S-5 Rev. Final

Appendix C JPM STEPS Form ES-C-1 Perform Step: 6 Perform the following steps prior to evacuating the Control Room:

1 & 1.f

• Stop ALL operating Heater Drain Pumps, FW-5A/B/C.

Standard: PERFORMED the following:

• PLACED FW-5C, HTR DRN PUMP handswitch in STOP (critical).

• OBSERVED green STOP light lit (NOT critical).

Comment: SAT UNSAT Perform Step: 7 Perform the following steps prior to evacuating the Control Room:

1 & 1.g

• Ensure ALL of the following Turbine Lube Oil equipment is running:

• LO-3, Turning Gear Oil Pump

• LO-8, Motor Suction Oil Pump

• LO-4, DC Oil Pump,

• Turbine Lift Pumps, LO-14A/B/C Standard: PERFORMED the following:

• PLACED LO-3, TURNING GEAR OIL PUMP handswitch in START (critical).

• PLACED LO-8, MOTOR SUCTION OIL PUMP handswitch in START (critical).

• PLACED LO-4, EMGY BRG OIL PUMP handswitch in START (critical).

• PLACED LO-14A, TURBINE BEARING LUBE OIL LIFT OIL PUMP handswitch in START (critical).

• PLACED LO-14B, TURBINE BEARING LUBE OIL LIFT OIL PUMP handswitch in START (critical).

• PLACED LO-14C, TURBINE BEARING LUBE OIL LIFT OIL PUMP handswitch in START (critical).

• OBSERVED red START lights lit (NOT critical).

• OBSERVED six white DISCH PRESS lights lit (NOT critical).

Terminating Cue: This JPM is complete.

Comment: SAT UNSAT STOP TIME:

Page 5 of 6 NRC JPM S-5 Rev. Final

Appendix C JPM CUE SHEET Form ES-C-1 INITIAL CONDITIONS: Given the following conditions:

• A toxic gas leak is requiring evacuation of the Control Room.

INITIATING CUE: The Control Room Supervisor directs you to PERFORM the following:

• EXECUTE required actions prior to a Control Room Evacuation per AOP-07, Evacuation of Control Room,Section I, Plant to Hot Shutdown.

Page 6 of 6 NRC JPM S-5 Rev. Final

Appendix C JPM WORKSHEET Form ES-C-1 Facility: FCS JPM # NRC S-6 Task # 0369 K/A # 026.A4.05 3.5 / 3.5 SF-5

Title:

Reset Containment Spray Actuation Signal Examinee (Print):

Testing Method:

Simulated Performance: Classroom:

Actual Performance: X Simulator: X Alternate Path: Plant:

Time Critical:

READ TO THE EXAMINEE I will explain the Initial Conditions, which steps to simulate or discuss, and provide an Initiating Cue.

When you complete the task successfully, the objective for this JPM will be satisfied.

Initial Conditions: Given the following conditions:

• A Main Steam Line Break has occurred inside Containment.

• EOP- 05, Uncontrolled Heat Extraction, is in progress.

• Containment Spray Pumps SI-3A and SI-3B are running.

• Containment Pressure is less than 3 psig.

• All Containment Cooling and Filtering Units are in service.

Initiating Cue: The Control Room Supervisor directs you to PERFORM the following:

• RESET the Containment Spray Actuation Signal and return Containment Spray Pumps to Standby per EOP/AOP Floating Step A, Containment Spray Termination.

Task Standard: Utilizing Floating Step A, reset CPHS, CSAS, and SGLS lockout relays and secured Containment Spray Pumps.

Required Materials: EOP/AOP Floating Steps, Rev. 7.

Validation Time: 15 minutes Completion Time: ________ minutes Comments:

Result: SAT UNSAT Examiner (Print / Sign): Date:

Page 1 of 10 NRC JPM S-6 Rev. Final As Run

Appendix C JPM WORKSHEET Form ES-C-1 SIMULATOR SETUP BOOTH OPERATOR:

INITIALIZE to IC-120:

Type Item Value Condition MALF/SGN SGN01B (Main Steam Line B 0.25 **Simulator is frozen >30 Leak Inside Containment) minutes into a steam header rupture in containment

• PLACE Simulator in RUN then DELETE all CPHS overrides.

BOOTH OPERATOR NOTE:

• VERIFY CPHS overrides are deleted after Simulator is in RUN.

EXAMINER:

PROVIDE the examinee with a copy of:

• EOP/AOP Floating Step A, Containment Spray Termination.

Page 2 of 10 NRC JPM S-6 Rev. Final As Run

Appendix C JPM STEPS Form ES-C-1

- Check Mark Denotes Critical Step START TIME:

Examiner Note: The following steps are from EOP/AOP Floating Steps, FS-A.

NOTE Stopping SI-3A or SI-3B will result in closure of one spray valve, HCV-344 or HCV-345 by interlock which will extend the time to RAS.

CAUTION Containment Spray may affect proper operation of RCPs, non-qualified equipment, Containment Sump, and instrumentation inside the Containment.

When the termination criterion is satisfied, Containment Spray should be promptly secured Perform Step: 1 IF Containment Spray has been initiated AND ALL of the following 1 conditions are satisfied:

• Two CS pumps are operating

• Containment pressure is less than 60 psig and NOT rising

• At least one VA-3A/B in service

• At least one VA-7C/D in service THEN perform the following:

Standard: DETERMINED all conditions are met per Initial Conditions.

Comment: SAT UNSAT Examiner Note: Applicant may go directly to Procedure step 2 (JPM Perform Step

5) and secure BOTH Containment Spray Pumps. If this is done, this critical action is satisfied by the action performed in Procedure step 2.

Applicant may notice a slight rise in Containment temperature when the Containment Spray Pump is secured.

Examiner Cue: If applicant hesitates, REPORT as Control Room Supervisor that the slight temperature rise is expected and they should continue.

Perform Step: 2 Ensure only ONE CS pump is operating.

1.a Standard: PERFORMED the following:

• PLACED SI-3A or SI-3B, CNTMT SPRAY PUMP in PULL-TO-LOCK (critical).

• OBSERVED pump indicating lights off (NOT critical).

Comment: SAT UNSAT Page 3 of 10 NRC JPM S-6 Rev. Final As Run

Appendix C JPM STEPS Form ES-C-1 Examiner Note: Applicant may place HCV-344 or HCV-345, CNTMT SPRAY VLV CONTROL SWITCH in OVERRIDE, but is not required.

Perform Step: 3 Ensure only ONE of the following valves is open:

1.b

• HCV-344

• HCV-345 Standard: PERFORMED the following:

• OBSERVED that red OPEN light lit for one valve and green CLOSED light lit for the other valve.

• HCV-344 is open when SI-3B is running

• HCV-345 is open when SI-3A is running Comment: SAT UNSAT Perform Step: 4 Ensure total CS flow is at least 1800 gpm.

1.c Standard: OBSERVED approximately 2400 gpm of combined flow on FI-343 and FI-342 SPRAY FLOW meters.

Comment: SAT UNSAT NOTE Terminating Containment Spray prior to resetting actuation relays will require increased monitoring of containment parameters.

Examiner Note: Applicant may go directly to Procedure step 2 (JPM Perform Step

5) and secure BOTH Containment Spray Pumps. If this is done, the applicant must secure BOTH Containment Spray Pumps in this step.

Perform Step: 5 IF CS pump(s) are operating, AND ALL of the following conditions are 2 satisfied:

• Containment pressure is less than 30 psig and stable or lowering

• Containment Spray is NOT required for Containment cooling

• At least one VA-3A/B in service

• At least one VA-7C/D in service THEN terminate Containment Spray by performing the following:

Standard: DETERMINED all conditions are met per Initial Conditions.

Comment: SAT UNSAT Page 4 of 10 NRC JPM S-6 Rev. Final As Run

Appendix C JPM STEPS Form ES-C-1 Examiner Note: Applicant will place valve left open at Perform Step 3 in OPEN.

Perform Step: 6 Place the control switches for the open valve(s) in "OPEN":

2.a

• HCV-344

• HCV-345 Standard: PERFORMED the following for the valve that is open:

• PLACED HCV-344 or HCV-345, CNTMT SPRAY VLV CONTROL SWITCH in OPEN (critical).

• OBSERVED red OPEN light lit (NOT critical).

Comment: SAT UNSAT Examiner Note: Applicant will place SI-3A and SI-3B in PULL-TO-LOCK. Due to system configuration, SI-3C is not lined up for auto start, and is NOT required to be placed in PULL-TO-LOCK.

Perform Step: 7 Place all CS pumps in "PULL-TO-LOCK":

2.b

• SI-3A

• SI-3B

• SI-3C Standard: PERFORMED one the following:

• PLACED SI-3A, CNTMT SPRAY PUMP in PULL-TO-LOCK (critical).

• PLACED SI-3B, CNTMT SPRAY PUMP in PULL-TO-LOCK (critical).

• OBSERVED all pump indicating lights off (NOT critical).

Comment: SAT UNSAT Examiner Note: Applicant will place valve left OPEN at Perform Step 6 in CLOSE.

Perform Step: 8 Close BOTH Containment Spray Valves:

2.c Standard: PERFORMED the following:

• PLACED HCV-344 or HCV-345, CNTMT SPRAY VLV CONTROL SWITCH in OVERRIDE or AUTO (critical).

• OBSERVED green CLOSE light lit (NOT critical).

Comment: SAT UNSAT Page 5 of 10 NRC JPM S-6 Rev. Final As Run

Appendix C JPM STEPS Form ES-C-1 Perform Step: 9 Place BOTH control switches in "AUTO":

2.d

• HCV-344

• HCV-345 Standard: PERFORMED the following:

• PLACED HCV-344, CNTMT SPRAY VLV CONTROL SWITCH in AUTO (critical).

• PLACED HCV-345, CNTMT SPRAY VLV CONTROL SWITCH in AUTO (critical)

• OBSERVED white AUTO light off (NOT critical).

Comment: SAT UNSAT NOTES

1. Resetting CPHS and SGLS Lockout Relays may reset SGIS. HCV-1105 and HCV-1106 may reopen.

2. Resetting PPLS, CPHS or SGLS Lockout Relays will reset Containment Spray.

Perform Step: 10 IF resetting actuation relays, THEN perform the following:

3 Standard: DETERMINED actuation relays will be RESET.

Comment: SAT UNSAT Perform Step: 11 IF Containment pressure less than or equal to 3 psig, THEN reset all of 3.a the following relays:

• 86A/CPHS

• 86B/CPHS Standard: PERFORMED the following:

• TURNED 86A/CPHS relay in CLOCKWISE direction until LATCHED (critical).

• TURNED 86B/CPHS relay in CLOCKWISE direction until LATCHED (critical)

• OBSERVED black relay flag and amber light lit (NOT critical).

Comment: SAT UNSAT Page 6 of 10 NRC JPM S-6 Rev. Final As Run

Appendix C JPM STEPS Form ES-C-1 Perform Step: 12 Reset ALL of the following relays:

3.b

• 86A1/CPHS

• 86B1/CPHS Standard: PERFORMED the following:

• TURNED 86A1/CPHS relay in CLOCKWISE direction until LATCHED (critical).

• TURNED 86B1/CPHS relay in CLOCKWISE direction until LATCHED (critical)

• OBSERVED black relay flag and amber light lit (NOT critical).

Comment: SAT UNSAT Perform Step: 13 Reset ALL of the following CSAS relays:

3.c

• 86A/CSAS

• 86B/CSAS

• 86A1/CSAS

• 86B1/CSAS Standard: PERFORMED the following:

• TURNED 86A/CSAS relay in CLOCKWISE direction until LATCHED (critical).

• TURNED 86B/CSAS relay in CLOCKWISE direction until LATCHED (critical)

• TURNED 86A1/CSAS relay in CLOCKWISE direction until LATCHED (critical).

• TURNED 86B1/CSAS relay in CLOCKWISE direction until LATCHED (critical)

• OBSERVED black relay flag and amber light lit (NOT critical).

Comment: SAT UNSAT Page 7 of 10 NRC JPM S-6 Rev. Final As Run

Appendix C JPM STEPS Form ES-C-1 Examiner Note: The following actions are performed at CB-4.

Perform Step: 14 Reset SGLS by performing the following:

3.d

• Block SGLS-A and SGLS-B by performing the following:

• Place the SGLS Block key into the SGLS Block key switch.

• Block SGLS-A and SGLS-B by turning key to "BLOCK".

• Verify at least one of the following SGLS Blocked alarms annunciates (CB-4; A8):

• "SGLS "A" BLOCKED"

• SGLS "B" BLOCKED" Time _____

Standard:

• DETERMINED Steam Generator Low Pressure Signal will be RESET. (NOT critical)

• REMOVED SGLS Block Key from Key Holder and PLACED SGLS Block Key into SGLS Block Key Switch. (critical)

• TURNED Key in SGLS Block Key Switch to BLOCK position.

(critical)

• VERIFIED Annunciator Panel A8 SGLS "A" BLOCKED (Window D-4L) or SGLS "B" BLOCKED (Window D-5U) in alarm and RECORDED time. (NOT critical)

Comment: SAT UNSAT Perform Step: 15 Reset BOTH of the following SGLS relays:

3.d.1).c).2)

• 86A/SGLS

• 86B/SGLS Standard: PERFORMED the following:

• TURNED 86A/SGLS relay in CLOCKWISE direction until LATCHED (critical).

• TURNED 86B/SGLS relay in CLOCKWISE direction until LATCHED (critical).

• OBSERVED black relay flag (NOT critical).

Comment: SAT UNSAT Page 8 of 10 NRC JPM S-6 Rev. Final As Run

Appendix C JPM STEPS Form ES-C-1 Examiner Note: Only SI-3A AND SI-3B will be placed in AFTER STOP.

Perform Step: 16 IF returning CS to standby, THEN perform the following:

3.e.1)

• Place CS Pumps SI-3A/B/C to "AFTER STOP".

Standard: PERFORMED the following:

• PLACED SI-3A, CNTMT SPRAY PUMP in AFTER STOP (critical).

• PLACED SI-3B, CNTMT SPRAY PUMP in AFTER STOP (critical).

• OBSERVED green STOP lights lit (NOT critical).

Comment: SAT UNSAT Perform Step: 17 Place BOTH Containment Spray Valves in "AUTO":

3.e.2)

• HCV-344

• HCV-345 Standard: DETERMINED HCV-344 and HCV-345 already in AUTO.

Terminating Cue: This JPM is complete.

Comment: SAT UNSAT STOP TIME:

Page 9 of 10 NRC JPM S-6 Rev. Final As Run

Appendix C JPM CUE SHEET Form ES-C-1 INITIAL CONDITIONS: Given the following conditions:

• A Main Steam Line Break has occurred inside Containment.

• EOP- 05, Uncontrolled Heat Extraction, is in progress.

• Containment Spray Pumps SI-3A and SI-3B are running.

• Containment Pressure is less than 3 psig.

• All Containment Cooling and Filtering Units are in service.

INITIATING CUE: The Control Room Supervisor directs you to PERFORM the following:

• RESET the Containment Spray Actuation Signal and return Containment Spray Pumps to Standby per EOP/AOP Floating Step A, Containment Spray Termination.

Page 10 of 10 NRC JPM S-6 Rev. Final As Run

Appendix C JPM WORKSHEET Form ES-C-1 Facility: FCS JPM # NRC S-7 Task # 0344 K/A # 064.A4.06 3.9 / 3.9 SF-6

Title:

Parallel and Load Emergency Diesel Generator Examinee (Print):

Testing Method:

Simulated Performance: Classroom:

Actual Performance: X Simulator: X Alternate Path: X Plant:

Time Critical:

READ TO THE EXAMINEE I will explain the Initial Conditions, which steps to simulate or discuss, and provide an Initiating Cue.

When you complete the task successfully, the objective for this JPM will be satisfied.

Initial Conditions: Given the following conditions:

• DG-1 has been manually started and is at idle speed.

• YCV-871 G/H/E Inlet and Exhaust Dampers have been verified OPEN.

• Jacket water temperature is 128°F.

• All Prerequisites are met.

Initiating Cue: The Control Room Supervisor directs you to PERFORM the following:

• PARALLEL and LOAD Emergency Diesel Generator DG-1 with Bus 1A3 per OI-DG-1, Diesel Generator Operation, Attachment 1, Idle Speed Start and Loading.

• START at Step 4.b.

• LOAD DG-1 to 2000 KW.

Task Standard: Utilizing OI-DG-1, raised DG-1 speed, paralleled and loaded to Bus 1A3, then tripped DG-1 Breaker when load rose uncontrollably.

Required Materials: OI-DG-1, Diesel Generator Operation, Rev. 63.

TDB-III.26, Diesel Generator Capability Curve (4160 V), Rev. 5.

TDB-III.26.A, Diesel Generator Loading Curve, Rev. 16.

Validation Time: 15 minutes Completion Time: ________ minutes Comments:

Result: SAT UNSAT Examiner (Print / Sign): Date:

Page 1 of 7 NRC JPM S-7 Rev. Final

Appendix C JPM WORKSHEET Form ES-C-1 SIMULATOR SETUP BOOTH OPERATOR:

INITIALIZE to IC-111. Load and execute scenario NRC JPM S7 Type Item Value Condition Override P20_185_3 (DG/1 Governor Sel 1 Condition: H_P20_033_1 SW Raise Position) ge 800 (When DG Watts are greater than 800, override governor switch to raise)

BOOTH OPERATOR NOTE:

• After each JPM, VERIFY Synchroscope Switch is moved from the D1/BUS 1A3 Sync Switch position prior to performance by the next examinee.

EXAMINER:

PROVIDE the examinee with a copy of:

• OI-DG-1, Diesel Generator Operation.

• Attachment 1, Idle Speed Start and Loading, INITIALED through Step 4.a.

• TDB-III.26, Diesel Generator Capability Curve (4160 V).

• TDB-III.26.A, Diesel Generator Loading Curve.

Page 2 of 7 NRC JPM S-7 Rev. Final

Appendix C JPM STEPS Form ES-C-1

- Check Mark Denotes Critical Step START TIME:

Examiner Note: The following steps are from OI-DG-1, Attachment 1.

Perform Step: 1 Place CS-65/D1, Diesel Generator D1 Governor, to Raise until the 4.b Diesel Speed is 900 rpm.

Standard: PERFORMED the following:

• PLACED CS-65/D1, DIESEL GENERATOR D1 GOVERNOR, to RAISE position until Diesel Speed is 900 rpm. (critical).

• OBSERVED Diesel speed rising to 900 rpm on Diesel Generator DG-1 Engine Tachometer (NOT critical).

Comment: SAT UNSAT Perform Step: 2 Verify the Generator Field flashed by performing one of the following:

4.c

• Ready to Load light is ON (AI-30A)

OR

• Generator frequency is responding Standard: OBSERVED the following:

• Ready to Load red light is ON at Panel AI-30A.

• Generator frequency is 60 Hz at 900 rpm.

Examiner Note: Automatic field flashing occurs at approximately 700 rpm.

Comment: SAT UNSAT Perform Step: 3 (LOCAL) Inspect field flash circuitry by performing the following:

4.d

• Verify that Control Relay 2CR in Panel AI-133A is not energized.

• Verify that Field Flash Current Limiting Resistors (1R4, 1R5, 1R6, and 1R7) in Panel AI-133A are not damaged due to overheating.

Contact System Engineer if damage is suspected.

Standard: CONTACTED Auxiliary Operator at DG-1 to verify Field Flash Circuitry.

Examiner Cue: Auxiliary Operator reports Field Flash Circuitry is satisfactory.

Comment: SAT UNSAT Page 3 of 7 NRC JPM S-7 Rev. Final

Appendix C JPM STEPS Form ES-C-1 Perform Step: 4 (LOCAL) Place AI-133A-S4, Diesel Generator DG-1 Electronic Droop 4.e Control Switch, in ENABLED. (AI-133A)

Standard: CONTACTED Auxiliary Operator at DG-1 to PLACE AI-133A-S4, Diesel Generator DG-1 Electronic Droop Control Switch, in ENABLED position.

Examiner Cue: Auxiliary Operator reports Electronic Droop Control Switch in ENABLED position.

Comment: SAT UNSAT Examiner Note: The panel holding the Synchroscope and Running and Incoming Volts indications is hinged and can be rotated for better viewing.

Perform Step: 5 Place D1/BUS 1A3 Sync Switch to ON.

4.f Standard: PERFORMED the following:

• LOCATED and INSERTED Synchroscope Switch into D1/BUS 1A3 SYNC SWITCH position and TURNED to ON (critical).

• OBSERVED Synchroscope rotation (NOT critical).

Comment: SAT UNSAT Perform Step: 6 Adjust CS-90/D1, Diesel Generator D1 Voltage Regulator, until the 4.g RUNNING VOLTS is approximately matched to the INCOMING VOLTS on the Synchroscope or the ERF DGD Display.

Standard: PERFORMED the following:

• ADJUSTED CS-90/D1, DIESEL GENERATOR D1 VOLTAGE REGULATOR switch until RUNNING VOLTS is approximately MATCHED (within ~100 volts) to INCOMING VOLTS (critical).

Comment: SAT UNSAT NOTE Recommended synchroscope speed is less than 1 revolution per 10 seconds.

Perform Step: 7 Adjust CS-65/D1 until the Synchroscope is rotating slowly in the FAST 4.h direction.

Standard: ADJUSTED CS-65/D1, DIESEL GENERATOR D1 GOVERNOR until Synchroscope is ROTATING SLOWLY in FAST direction and less than 1 revolution per 10 seconds.

Comment: SAT UNSAT Page 4 of 7 NRC JPM S-7 Rev. Final

Appendix C JPM STEPS Form ES-C-1 NOTE Steps 4.i and 4.j may be performed without the procedure in hand.

Sign-offs may be completed after these steps are performed.

CAUTIONS

1. Load must be immediately picked up following closure of 1AD1 to prevent motorizing the Diesel Generator.

2. Governor controls are extremely sensitive.

Perform Step: 8 WHEN the Synchroscope is between 11 and 12 O'CLOCK, THEN close 4.i 1AD1 BREAKER.

Standard: PERFORMED the following:

• When Synchroscope was between 11 and 12 oclock, PLACED 1AD1 BREAKER in CLOSE position (critical).

• OBSERVED red CLOSE light lit (NOT critical).

Comment: SAT UNSAT Perform Step: 9 Place CS-65/D1 to Raise to pick up 250-350 KW.

4.j

• Time _____

Standard: PERFORMED the following:

• PLACED CS-65/D1, DIESEL GENERATOR D1 GOVERNOR in RAISE and PICKED UP 250 KW to 350 KW load (critical).

• OBSERVED load rising on DG-1 WATT METER (NOT critical).

• RECORDED Time at Step 4.j (NOT critical).

Comment: SAT UNSAT Perform Step: 10 Place D1/BUS 1A3 Sync Switch to OFF.

4.k Standard: PLACED D1/BUS 1A3 SYNC SWITCH in OFF.

Comment: SAT UNSAT Perform Step: 11 IF the Diesel is loaded AND Y3287A, ERF 1A3 Bus Voltage, is greater 4.l than 4375 VAC, THEN immediately notify the System Engineer.

Standard: DETERMINED Bus 1A3 voltage is normal.

Comment: SAT UNSAT Page 5 of 7 NRC JPM S-7 Rev. Final

Appendix C JPM STEPS Form ES-C-1 Examiner Note: When DG-1 load is > 800 KW and CS-65/D1 is placed in RAISE, DG-1 load will continue to rise to ~3200 KW.

Examiner Note: The following steps represent the Alternate Path of this JPM.

NOTE

1. Load should be maintained below the 2000 hr Rating vs Ambient Temp curve per TDB-III.26A Figure 1, DG-1 Output Power Rate.

2. Power factor may be determined by using TDB-III.26, Diesel Generator Capability Curve.

3. Current is normally limited to 400 amps at 2500 KW.

4. Diesel Generator manual loading and unloading rates should be maintained at less than 500 KW per minute.

5. Steps 4.m and 4.n may be repeated as necessary while the diesel is loaded. Sign-offs may be completed after these steps are performed.

Examiner Note: Applicant must OPEN the Diesel Generator Output Breaker within 2 minutes of exceeding a loading rate of 500 KW per minute.

Perform Step: 12 Place CS-65/D1 to RAISE picking up the required DG-1 Load.

4.m Standard: PERFORMED the following:

• PLACED CS-65/D1, DIESEL GENERATOR D1 GOVERNOR in RAISE and PICKED UP load (critical).

• DETERMINED DG-1 load rising out of control and PLACED 1AD1 BREAKER in TRIP position (critical).

• OBSERVED green TRIP light lit (NOT critical).

Comment: SAT UNSAT Perform Step: 13 Inform Control Room Supervisor of problem.

Standard: INFORMED Control Room Supervisor DG-1 Output Breaker tripped due to excessive Diesel loading.

Terminating Cue: The CRS has been notified. This JPM is complete.

Comment: SAT UNSAT STOP TIME:

Page 6 of 7 NRC JPM S-7 Rev. Final

Appendix C JPM CUE SHEET Form ES-C-1 INITIAL CONDITIONS: Given the following conditions:

• DG-1 has been manually started and is at idle speed.

• YCV-871 G/H/E Inlet and Exhaust Dampers have been verified OPEN.

• Jacket water temperature is 128°F.

• All Prerequisites are met.

INITIATING CUE: The Control Room Supervisor directs you to PERFORM the following:

• PARALLEL and LOAD Emergency Diesel Generator DG-1 with Bus 1A3 per OI-DG-1, Diesel Generator Operation, Attachment 1, Idle Speed Start and Loading.

• START at Step 4.b.

• LOAD DG-1 to 2000 KW.

Page 7 of 7 NRC JPM S-7 Rev. Final

PAGE 1 OF 2 Fort Calhoun Station Unit 1 TDB-III.26 TECHNICAL DATA BOOK DIESEL GENERATOR CAPABILITY CURVE (4160 VOLTS)

Change No. EC 38104 Reason for Change Correct the Note associated with Diesel Generator Capability Curve.

Requestor Richard Ronning Preparer Daniel A Hochstein Issue Date 03-28-06 3:00 pm R5

FORT CALHOUN STATION TDB-III.26 TECHNICAL DATA BOOK PAGE 2 OF 2 Diesel Generator Capability Curve (4160 Volts)

NOTE: Safe operating area is between the 0.5 and 1.0 power factor lines and less then the 450 amp line.

R5

PAGE 1 OF 9 Fort Calhoun Station Unit 1 TDB-III.26.A TECHNICAL DATA BOOK DIESEL GENERATOR LOADING CURVE Change No. EC 62189 Reason for Change The changes were made as a result of revising EA-FC-92-072.

Adding additional figures to TDB-III.26.

Requestor E. Noseir Preparer K. Bessey Issue Date 10-02-13 1628 R16

FORT CALHOUN STATION TDB-III.26.A TECHNICAL DATA BOOK PAGE 2 OF 9 Figure 1 - (LOCA) DG-1 Output Power Rating Ethylene Glycol Coolant (110.0 Deg. F Maximum Ambient Limit)

R16

FORT CALHOUN STATION TDB-III.26.A TECHNICAL DATA BOOK PAGE 3 OF 9 Figure 2 - (MSLB) DG-1 Output Power Rating Ethylene Glycol Coolant (110.0 Deg. F Maximum Ambient Limit)

R16

FORT CALHOUN STATION TDB-III.26.A TECHNICAL DATA BOOK PAGE 4 OF 9 Figure 3 - (LOCA) DG-2 Output Power Rating Ethylene Glycol Coolant (109 Deg. F Maximum Ambient Limit)

R16

FORT CALHOUN STATION TDB-III.26.A TECHNICAL DATA BOOK PAGE 5 OF 9 Figure 4 - (MSLB) DG-2 Output Power Rating Ethylene Glycol Coolant (105 Deg. F Maximum Ambient Limit)

R16

FORT CALHOUN STATION TDB-III.26.A TECHNICAL DATA BOOK PAGE 6 OF 9 Figure 5 - (LOCA) DG-1 Output Power Rating Water Coolant (110 Deg. F Maximum Ambient Limit)

R16

FORT CALHOUN STATION TDB-III.26.A TECHNICAL DATA BOOK PAGE 7 OF 9 Figure 6 - (MSLB) DG-1 Output Power Rating Water Coolant (110 Deg. F Maximum Ambient Limit)

R16

FORT CALHOUN STATION TDB-III.26.A TECHNICAL DATA BOOK PAGE 8 OF 9 Figure 7 - (LOCA) DG-2 Output Power Rating Water Coolant (114 Deg. F Maximum Ambient Limit)

R16

FORT CALHOUN STATION TDB-III.26.A TECHNICAL DATA BOOK PAGE 9 OF 9 Figure 8 - (MSLB) DG-2 Output Power Rating Water Coolant (105 Deg. F Maximum Ambient Limit)

R16

Appendix C JPM WORKSHEET Form ES-C-1 Facility: FCS JPM # NRC S-8 Task # 0778 K/A # 012.A4.02 3.3 / 3.4 SF-7

Title:

Adjust Reactor Protection System TCOLD Calibration Examinee (Print):

Testing Method:

Simulated Performance: Classroom:

Actual Performance: X Simulator: X Alternate Path: Plant:

Time Critical:

READ TO THE EXAMINEE I will explain the Initial Conditions, which steps to simulate or discuss, and provide an Initiating Cue.

When you complete the task successfully, the objective for this JPM will be satisfied.

Initial Conditions: Given the following conditions:

• Plant is operating at 100% power.

• Channel D TCOLD calibration is indicating high.

Initiating Cue: The Shift Manager directs you to PERFORM the following:

• ADJUST Reactor Protection System (RPS) TCOLD calibration on Channel D per OI-RPS-2, Reactor Protective System-TM/LP TCOLD CAL Calibration.

Task Standard: Utilizing OI-RPS-2, bypassed Channel D TM/LP Trip Unit, adjusted TCOLD CAL Calibration, then returned Channel D TM/LP Trip Unit to service.

Required Materials: OI-RPS-2, Reactor Protective System-TM/LP TCOLD CAL Calibration, Rev. 10.

TM/LP Trip Unit # 9 Bypass Key Validation Time: 19 minutes Completion Time: ________ minutes Comments:

Result: SAT UNSAT Examiner (Print / Sign): Date:

Page 1 of 7 NRC JPM S-8 Rev. Final

Appendix C JPM WORKSHEET Form ES-C-1 SIMULATOR SETUP BOOTH OPERATOR:

INITIALIZE to IC-111 or any at power Initial Condition:

• ENSURE TCOLD Calibrate Potentiometer on Channel D is set to greater than 5.20.

BOOTH OPERATOR NOTE:

• After each JPM, VERIFY Channel D TM/LP Trip Unit # 9 Bypass Key is removed from AI-31D prior to performance by the next examinee.

EXAMINER:

PROVIDE the examinee with a copy of:

• OI-RPS-2, Reactor Protective System-TM/LP TCOLD CAL Calibration.

• INITIALED through Prerequisites.

• PROVIDE the TM/LP Trip Unit # 9 Bypass Key.

EXAMINER NOTE: Only SROs can check out keys from the Key Locker at FCS.

Page 2 of 7 NRC JPM S-8 Rev. Final

Appendix C JPM STEPS Form ES-C-1

- Check Mark Denotes Critical Step START TIME:

Examiner Note: The following steps are from OI-RPS-2.

Perform Step: 1 Record Tcold DVM readings on all four RPS channels.

1.a

• AI-31A °F

• AI-31B °F

• AI-31C °F

• AI-31D °F Standard: SELECTED Channel A/B/C/D TCOLD on Digital Voltmeters (DVM) and RECORDED temperatures at Step 1.a.

Comment: SAT UNSAT Perform Step: 2 Record Tcold cal DVM readings on all four RPS channels.

1.b

• AI-31A °F

• AI-31B °F

• AI-31C °F

• AI-31D °F Standard: SELECTED Channel A/B/C/D TCOLD CAL on Digital Voltmeters (DVM) and RECORDED temperatures at Step 1.b.

Comment: SAT UNSAT Perform Step: 3 Record Tcold cal POT settings.

1.c

• AI-31A _____

• AI-31B _____

• AI-31C _____

• AI-31D _____

Standard: RECORDED Channel A/B/C/D TCOLD CAL POT settings at Step 1.c.

Comment: SAT UNSAT Examiner Note: PROVIDE RPS TM/LP Trip Unit # 9 Bypass Key.

Perform Step: 4 Obtain the RPS TM/LP Trip Unit # 9 Bypass Key.

1.d Standard: OBTAINED RPS TM/LP Trip Unit # 9 Bypass Key from Key Locker.

Comment: SAT UNSAT Page 3 of 7 NRC JPM S-8 Rev. Final

Appendix C JPM STEPS Form ES-C-1 CAUTION Only ONE channel shall be adjusted at a time.

Perform Step: 5 Log into Technical Specification 2.15.1(1) 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> LCO for selected 1.e channel:

• AI-31D Standard: INFORMED CRS of entry into Technical Specification LCO 2.15.1 for Reactor Protection System Channel D.

Examiner Cue: CRS logs entry into Technical Specification LCO.

Comment: SAT UNSAT Perform Step: 6 Bypass TM/LP trip unit # 9 on the selected channel using Bypass Key:

1.f

• AI-31D Standard: PERFORMED the following:

• INSERTED key into Channel D TM/LP Trip Unit # 9 and TURNED to BYPASS Channel D (critical).

• OBSERVED Channel D Trip Unit amber light lit (NOT critical).

Comment: SAT UNSAT Perform Step: 7 Adjust Tcold cal POT on the selected channel until the Tcold cal DVM 1.g reading equals highest RPS channel Tcold recorded in Step a.

• AI-31D Standard: ADJUSTED T COLD CAL POT on Channel D until T COLD CAL DVM reading equals highest RPS channel T COLD recorded in Step a.

Comment: SAT UNSAT Perform Step: 8 Ensure selected TM/LP Trip Unit #9 is RESET by depressing T/U Alarm 1.h Reset.

• AI-31D Standard: PERFORMED the following:

• DEPRESSED T/U Alarm Reset on Channel D TM/LP Trip Unit #9.

• OBSERVED Channel D Trip Unit #9 alarm RESET light off.

Comment: SAT UNSAT Page 4 of 7 NRC JPM S-8 Rev. Final

Appendix C JPM STEPS Form ES-C-1 Perform Step: 9 Verify TM/LP Trip Unit #9 Lights are reset.

1.i Standard: VERIFIED Channel D Trip Unit #9 alarm RESET light off.

Comment: SAT UNSAT Perform Step: 10 Remove Bypass Key for selected TM/LP Trip Unit.

1.j

• AI-31D Standard: PERFORMED the following:

• REMOVED Channel D Trip Unit #9 Bypass Key (critical).

• OBSERVED Channel D Trip Unit amber light off (NOT critical).

Comment: SAT UNSAT Perform Step: 11 Exit Technical Specification 2.15.1(1) for the selected channel.

1.k

• AI-31D Standard: INFORMED CRS of exit from Technical Specification LCO 2.15.1 for Reactor Protection System Channel D.

Examiner Cue: CRS logs exit from Technical Specification LCO.

Comment: SAT UNSAT Perform Step: 12 Repeat Steps e through k for any remaining channels out of 1.l specification.

Standard: DETERMINED there are NO remaining Channels out of specification.

Comment: SAT UNSAT Perform Step: 13 Record Tcold cal DVM readings.

1.m

• AI-31A °F

• AI-31B °F

• AI-31C °F

• AI-31D °F Standard: SELECTED Channel A/B/C/D TCOLD CAL on Digital Voltmeters (DVM) and RECORDED temperatures at Step 1.m.

Comment: SAT UNSAT Page 5 of 7 NRC JPM S-8 Rev. Final

Appendix C JPM STEPS Form ES-C-1 Perform Step: 14 Record Tcold cal POT settings.

1.n

• AI-31A _____

• AI-31B _____

• AI-31C _____

• AI-31D _____

Standard: RECORDED Channel A/B/C/D TCOLD CAL POT settings at Step 1.n.

Terminating Cue: This JPM is complete.

Comment: SAT UNSAT STOP TIME:

Page 6 of 7 NRC JPM S-8 Rev. Final

Appendix C JPM CUE SHEET Form ES-C-1 INITIAL CONDITIONS: Given the following conditions:

• Plant is operating at 100% power.

• Channel D TCOLD calibration is indicating high.

INITIATING CUE: The Shift Manager directs you to PERFORM the following:

• ADJUST Reactor Protection System (RPS) TCOLD calibration on Channel D per OI-RPS-2, Reactor Protective System-TM/LP TCOLD CAL Calibration.

Page 7 of 7 NRC JPM S-8 Rev. Final

Appendix C JPM WORKSHEET Form ES-C-1 Facility: FCS JPM # NRC P-1 Task # 1398 K/A # 033.A2.02 2.7 / 3.0 SF-8

Title:

Spent Fuel Pool Cooling Restoration with SIAS Examinee (Print):

Testing Method:

Simulated Performance: X Classroom:

Actual Performance: Simulator:

Alternate Path: Plant: X Time Critical:

READ TO THE EXAMINEE I will explain the Initial Conditions, which steps to simulate or discuss, and provide an Initiating Cue.

When you complete the task successfully, the objective for this JPM will be satisfied.

Initial Conditions: Given the following conditions:

• AOP-36, Loss of Spent Fuel Pool Cooling, has been entered following a Safety Injection Actuation Signal.

• SIAS and CIAS have been RESET.

• 480 Volt Bus 1B4C is energized and available for Spent Fuel Pool Cooling Pump AC-5B.

• HC-478, STORAGE POOL HX AC-8 AC OUTL HCV-478 at AI-45 is OPEN.

Initiating Cue: The Control Room Supervisor directs you to PERFORM the following:

• RESTORE Spent Fuel Pool Cooling using Spent Fuel Pool Cooling Pump AC-5B per AOP-36, Loss of Spent Fuel Pool Cooling, Attachment H, Spent Fuel Pool Cooling Restoration with SIAS.

• START at Step 6.

Task Standard: Utilizing AOP-36, Attachment H, aligned valves, started Spent Fuel Pool Cooling Pump AC-5B, and restored SFP cooling flow.

Required Materials: AOP-36, Loss of Spent Fuel Pool Cooling, Rev. 11.

Validation Time: 15 minutes Completion Time: ________ minutes Comments:

Result: SAT UNSAT Examiner (Print / Sign): Date:

Page 1 of 6 NRC JPM P-1 Rev. Final

Appendix C JPM WORKSHEET Form ES-C-1 PLANT SETUP EXAMINER:

PROVIDE the examinee with a copy of:

• AOP-36, Loss of Spent Fuel Pool Cooling.

• Attachment H, Spent Fuel Pool Cooling Restoration with SAIS.

• INITIALED/PLACE KEEPING through Step 5.

Page 2 of 6 NRC JPM P-1 Rev. Final

Appendix C JPM STEPS Form ES-C-1

- Check Mark Denotes Critical Step START TIME:

Examiner Note: The following steps are from AOP-36, Attachment H, Step 6.

Examiner Note: MCC 4C2 is located in Corridor 4, 989' elev. of Auxiliary Building.

Examiner Note: MCC-4C2-F05 is a bucket style breaker. Turn to the left to RESET, then turn to right to ON.

Perform Step: 1 Verify breaker MCC-4C2-F05, "AC-5B FUEL STORAGE POOL CIRC 6 PUMP" is reset and closed (Corridor 4)

Standard: PERFORMED the following:

• OBSERVED breaker in the ON position with Green light LIT

• Candidate *may* (not required) reset breaker by: TURN handle on MCC-4C2-F05, AC-5B FUEL STORAGE POOL CIRC PUMP to LEFT to RESET breaker, then TURN handle on MCC-4C2-F05, AC-5B FUEL STORAGE POOL CIRC PUMP to RIGHT to ON.

Examiner Cue: Breaker handle pointing to ON. Green light is lit.

Comment: SAT UNSAT Examiner Note: Room 5 is located in Corridor 4, 989' elev. of Auxiliary Building.

Examiner Note: Radiation levels are somewhat elevated in Room 5. By selecting SFP AC-5B, the Applicant can remain in a LOW DOSE WAITING AREA while describing manipulations to be performed.

Perform Step: 2 Contact Shift RP for Room 5 entry.

7 Standard: CONTACTED Shift Radiation Protection for Room 5 entry.

Examiner Cue: Shift Radiation Protection has been contacted and approves entry.

Comment: SAT UNSAT Page 3 of 6 NRC JPM P-1 Rev. Final

Appendix C JPM STEPS Form ES-C-1 Examiner Note: AC-191 is a 90-degree butterfly valve. The operating handle has notches to keep the valve in position. The specific throttle position in this step is not critical, but throttling the valve is critical to prevent pump runout on start. Applicant may determine appropriate throttle position, or the Examiner may provide a cue.

Examiner Cue: If requested: The CRS directs you to place AC-191 at approximately 50% open.

Perform Step: 3 Throttle open the selected SFP pump discharge valve (Room 5):

8

• AC-191, "SPENT FUEL POOL CIRC PUMP AC-5B DISCHARGE VALVE" Standard: PERFORMED the following:

• SQUEEZED handle on AC-191, SPENT FUEL POOL CIRC PUMP AC-5B DISCHARGE VALVE and PLACED in a throttled position less than full open.

• OBSERVED valve handle at 45° from piping.

Examiner Cue: Valve handle is 45° offset from piping (or throttled appropriately).

Comment: SAT UNSAT CAUTION If the Discharge Valve for the non-running Spent Fuel Pool Circ Pump is Open, the pump will windmill backwards. The discharge valve must be Closed while starting the pump to prevent excessive starting current.

Perform Step: 4 Close the non-selected SFP pump discharge valve:

9

• AC-192, AC-5A Standard: PERFORMED the following:

• SQUEEZED handle on AC-192, SPENT FUEL POOL CIRC PUMP AC-5A DISCHARGE VALVE and PLACED in CLOSED position then RELEASED (critical).

• OBSERVED valve handle perpendicular to piping (NOT critical).

Examiner Cue: Valve handle is perpendicular to piping.

Comment: SAT UNSAT Page 4 of 6 NRC JPM P-1 Rev. Final

Appendix C JPM STEPS Form ES-C-1 Examiner Note: START pushbutton is located on wall behind pump.

Perform Step: 5 Start the selected SFP pump (Room 5):

10

• AC-5B, "SPENT FUEL POOL COOLING PUMP" Standard: PERFORMED the following:

• DEPRESSED black START pushbutton for AC-5B, SPENT FUEL POOL COOLING PUMP (critical).

• OBSERVED red START light lit and GREEN stop light off (NOT critical).

Examiner Cue: RED light is lit and GREEN light is off. Noise emanating from pump.

Comment: SAT UNSAT Perform Step: 6 Open the selected SFP pump discharge valve:

11

• AC-191 Standard: PERFORMED the following:

• SQUEEZED handle on AC-191, SPENT FUEL POOL CIRC PUMP AC-5B DISCHARGE VALVE and PLACED in OPEN position then RELEASED (critical).

• OBSERVED valve handle parallel with piping (NOT critical).

Examiner Cue: Valve handwheel is parallel with piping.

Comment: SAT UNSAT Examiner Note: Pressure gauge ranges from 0 to 300 psig.

Perform Step: 7 Verify SFP pump discharge pressure 40-60 psig.

12 Standard: OBSERVED pressure gauge on discharge of AC-5B between 40 psig and 60 psig.

Terminating Cue: Pressure gauge needle positioned 1/5th upscale. Another operator will throttle CCW flow to minimize dose rate. This JPM is complete.

Comment: SAT UNSAT STOP TIME:

Page 5 of 6 NRC JPM P-1 Rev. Final

Appendix C JPM CUE SHEET Form ES-C-1 INITIAL CONDITIONS: Given the following conditions:

• AOP-36, Loss of Spent Fuel Pool Cooling, has been entered following a Safety Injection Actuation Signal.

• SIAS and CIAS have been RESET.

• 480 Volt Bus 1B4C is energized and available for Spent Fuel Pool Cooling Pump AC-5B.

• HC-478, STORAGE POOL HX AC-8 AC OUTL HCV-478 at AI-45 is OPEN.

INITIATING CUE: The Control Room Supervisor directs you to PERFORM the following:

• RESTORE Spent Fuel Pool Cooling using Spent Fuel Pool Cooling Pump AC-5B per AOP-36, Loss of Spent Fuel Pool Cooling, Attachment H, Spent Fuel Pool Cooling Restoration with SIAS.

• START at Step 6.

Page 6 of 6 NRC JPM P-1 Rev. Final

Appendix C JPM WORKSHEET Form ES-C-1 Facility: FCS JPM # NRC P-2 Task # 0809 K/A #061.A2.05 3.1 / 3.4 SF-4S

Title:

Locally Start FW-54, Diesel Driven AFW Pump Examinee (Print):

Testing Method:

Simulated Performance: X Classroom:

Actual Performance: Simulator:

Alternate Path: X Plant: X Time Critical:

READ TO THE EXAMINEE I will explain the Initial Conditions, which steps to simulate or discuss, and provide an Initiating Cue.

When you complete the task successfully, the objective for this JPM will be satisfied.

Initial Conditions: Given the following conditions:

• EOP-06, Loss of All Feedwater, is in progress.

• FW-54, Diesel Driven Auxiliary Feedwater Pump, failed to start from the Control Room.

• Feedwater flow CANNOT be aligned to the Feed Ring.

• Control Room has provided AI-114, FW-54 Control Panel keys.

Initiating Cue: The Control Room Supervisor directs you to PERFORM the following:

• Locally START FW-54, Diesel Driven AFW Pump per EOP/AOP Attachments-HR Heat Removal, HR-16, FW-54 Operation.

• ESTABLISH Auxiliary Feedwater flow.

Task Standard: Utilizing HR-16, started FW-54 then deenergized and opened HCV-1384.

Required Materials: EOP/AOP Attachments-HR Heat Removal, Rev. 1.

Validation Time: 15 minutes Completion Time: ________ minutes Comments:

Result: SAT UNSAT Examiner (Print / Sign): Date:

Page 1 of 6 NRC JPM P-2 Rev. Final

Appendix C JPM WORKSHEET Form ES-C-1 PLANT SETUP EXAMINER:

PROVIDE the examinee with a copy of:

• EOP/AOP Attachments-HR Heat Removal.

• HR-16, FW-54 Operation.

Page 2 of 6 NRC JPM P-2 Rev. Final

Appendix C JPM STEPS Form ES-C-1

- Check Mark Denotes Critical Step START TIME:

Examiner Note: The following steps are from EOP/AOP Attachments, HR-16.

Examiner Note: FW-54 Room is located in Southeast corner of Turbine Building basement.

Perform Step: 1 Start FW-54, Diesel AFW Pump, by performing step a or b.

1, 1.a, & 1.b

• Start FW-54 from the Control Room by placing HC/FW-54, "AFW PUMP FW-54" in "START".

• (LOCAL) Start FW-54 by performing the following:

Standard: DETERMINED from Initial Conditions that Local Start of FW-54 is required.

Comment: SAT UNSAT Perform Step: 2 Obtain AI-114, FW-54 Control Panel keys, from one of the following:

b.1)

• EONT key ring

• Control Room Standard: DETERMINED from Initial Conditions that Control Room has provided AI-114, FW-54 Control Panel keys.

Comment: SAT UNSAT Perform Step: 3 Using key, place HC/FW-54-1, "LOCAL CONTROL SWITCH" in b.2) "STOP".

Standard: INSERTED key into 2 position switch and TURNED HC/FW-54-1, FW-54 LOCAL CONTROL SWITCH to STOP (left) position.

Examiner Cue: Key is in STOP position.

Comment: SAT UNSAT Perform Step: 4 Using key, place 43/FW-54, "CONTROL TRANSFER SWITCH" in b.3) "RESET".

Standard: INSERTED key into 3 position switch and TURNED 43/FW-54, FW-54 CONTROL TRANSFER SWITCH to RESET (center) position.

Examiner Cue: Key is in RESET position.

Comment: SAT UNSAT Page 3 of 6 NRC JPM P-2 Rev. Final

Appendix C JPM STEPS Form ES-C-1 Perform Step: 5 Using key, place "43/FW-54" in "LOCAL".

b.4)

Standard: INSERTED key into 3 position switch and TURNED 43/FW-54, FW-54 CONTROL TRANSFER SWITCH to LOCAL (right) position.

Examiner Cue: Key is in LOCAL position.

Comment: SAT UNSAT Perform Step: 6 Using key, place "HC/FW-54-1" in "RUN".

b.5)

Standard: PERFORMED the following:

• INSERTED key into 2 position switch and TURNED HC/FW-54-1, FW-54 LOCAL CONTROL SWITCH to RUN (right) position.

(critical).

• OBSERVED engine speed rising (NOT critical).

• OBSERVED engine noise rising (NOT critical).

• OBSERVED Diesel Room area for leaks (NOT critical).

Examiner Cue: Key is in RUN position. Engine noise and speed are rising.

Comment: SAT UNSAT Perform Step: 7 Feed through the Feed Ring by performing the following:

2 & 2.1 CA

• IF the Feed Ring is NOT available, THEN GO TO Step 0.

Standard: DETERMINED Feed Ring is NOT available per Initial Conditions and TRANSITIONED to Step 4.

Comment: SAT UNSAT Examiner Note: The following steps represent the Alternate Path of this JPM.

Perform Step: 8 Feed through the AFW Nozzles by performing the following:

4 & 4.a

• Open HCV-1384, FW/AFW Header Cross-Connect Valve.

Standard: CONTACTED Control Room to open HCV-1384, FW/AFW Header Cross-Connect Valve.

Examiner Cue: Control Room reports HCV-1384, FW/AFW Header Cross-Connect Valve will NOT open.

Comment: SAT UNSAT Page 4 of 6 NRC JPM P-2 Rev. Final

Appendix C JPM STEPS Form ES-C-1 Examiner Note: West Upper Electrical Penetration Room is located in the Auxiliary Building directly below Control Room.

Perform Step: 9 (LOCAL) IF valve will NOT open, THEN perform the following:

4.a.1 CA & 4.a.1.1) CA

• Place Breaker MCC-4C1-E03, "FW AND AUX FEED WATER CROSS CONNECTION VALVE" in "OFF" (West Upper Electrical Penetration Room).

Standard: ROTATED switch for MCC-4C1-E03, HCV-1384, FW AND AUX FEEDWATER CROSSCONNECTION VALVE breaker COUNTERCLOCKWISE to OFF position.

Examiner Cue: Breaker switch rotated then stopped.

Comment: SAT UNSAT Examiner Note: Room 81 is located on 1036' elev. of the Auxiliary Building and is accessed from the Turbine Deck.

Perform Step: 10 Manually open HCV-1384 (Room 81).

4.a.1.2) CA Standard: DEPRESSED clutch arm and ROTATED handwheel for HCV-1384, FW-AFW MAIN AND AUXILIARY FEEDWATER CROSSCONNECT VALVE in COUNTERCLOCKWISE direction until stopped.

Terminating Cue: Valve handwheel rotated then stopped. This JPM is complete.

Comment: SAT UNSAT STOP TIME:

Page 5 of 6 NRC JPM P-2 Rev. Final

Appendix C JPM CUE SHEET Form ES-C-1 INITIAL CONDITIONS: Given the following conditions:

• EOP-06, Loss of All Feedwater, is in progress.

• FW-54, Diesel Driven Auxiliary Feedwater Pump, failed to start from the Control Room.

• Feedwater flow CANNOT be aligned to the Feed Ring.

• Control Room has provided AI-114, FW-54 Control Panel keys.

INITIATING CUE: The Control Room Supervisor directs you to PERFORM the following:

• Locally START FW-54, Diesel Driven AFW Pump per EOP/AOP Attachments-HR Heat Removal, HR-16, FW-54 Operation.

• ESTABLISH Auxiliary Feedwater flow.

Page 6 of 6 NRC JPM P-2 Rev. Final

Appendix C JPM WORKSHEET Form ES-C-1 Facility: FCS JPM # NRC P-3 Task # 0735 K/A # 071 G2.1.30 4.4 / 4.0 SF-9

Title:

Terminate Release of Radioactive Gas Examinee (Print):

Testing Method:

Simulated Performance: X Classroom:

Actual Performance: Simulator:

Alternate Path: Plant: X Time Critical:

READ TO THE EXAMINEE I will explain the Initial Conditions, which steps to simulate or discuss, and provide an Initiating Cue.

When you complete the task successfully, the objective for this JPM will be satisfied.

Initial Conditions: Given the following conditions:

• A release from WD-29A, Waste Gas Decay Tank is in progress.

• RM-052, Auxiliary Building Ventilation Stack Radiation Monitor has gone into HIGH alarm.

• AOP-09, High Radioactivity is in progress.

• When HC-532, Waste Gas Release Control Switch was placed in CLOSE, FCV-532A and FCV-532C on AI-100 did NOT close.

• Independent Verification has been waived by the Shift Manager due to AOP-09 entry and high radiation in the Auxiliary Building.

Initiating Cue: The Control Room Supervisor directs you to PERFORM the following:

• TERMINATE release from WD-29A per OI-WDG-2, Waste Gas Disposal System Release, Attachment 3, Manual Waste Gas Release with FE-532 Unavailable.

• START at Step 2.17.

Task Standard: Utilizing OI-WDG-2, closed WD-158, isolated WD-29A via WD-132 terminating release from Waste Gas Decay Tank WG-29A.

Required Materials: AOP-09, High Radioactivity, Rev. 11.

OI-WDG-2, Waste Gas Disposal System Release, Rev. 30.

Validation Time: 15 minutes Completion Time: ________ minutes Comments:

Result: SAT UNSAT Examiner (Print / Sign): Date:

Page 1 of 6 NRC JPM P-3 Rev. Final

Appendix C JPM WORKSHEET Form ES-C-1 PLANT SETUP EXAMINER:

PROVIDE the examinee with a copy of:

• OI-WDG-2, Waste Gas Disposal System Release.

• Attachment 3, Manual Waste Gas Release with FE-532 Unavailable.

• Attachment 3 is INITIALED through Step 2.16.

Page 2 of 6 NRC JPM P-3 Rev. Final

Appendix C JPM STEPS Form ES-C-1

- Check Mark Denotes Critical Step START TIME:

Examiner Note: The following steps are from OI-WDG-2, Attachment 3.

Examiner Note: Panel AI-100 and Room 16 are adjacent to each other and located in Corridor 4, 989' elev. of Auxiliary Building.

Perform Step: 1 Verify the following Gas Release Control Valves closed:

2.17

• FCV-532A (AI-100)

• FCV-532C (AI-100)

• FCV-532B (Room 16)

Standard: NOTED the following:

• DETERMINED FCV-532A and FCV-532C did NOT close from Initial Conditions, or

• OBSERVED red OPEN lights lit and green CLOSE lights off at AI-100, and

• OBSERVED FCV-532B open in Room 16.

Examiner Cue: Red lights are lit on AI-100 for FCV-532A & FCV-532C. In Room 16, FCV-532B indicates mid position (between open and close position discs).

Comment: SAT UNSAT Examiner Note: All valves are located on the East wall of Room 16.

Perform Step: 2 Close WD-158.

2.18 Standard: ROTATED WD-158, WASTE GAS RELEASE HEADER FLOW ELEMENT FE-532 BYPASS LINE ISOLATION VALVE handwheel in CLOCKWISE direction until stopped.

Examiner Cue: Valve handwheel rotated then stopped.

Comment: SAT UNSAT Page 3 of 6 NRC JPM P-3 Rev. Final

Appendix C JPM STEPS Form ES-C-1 Examiner Note: Valve located about 6 feet above the floor.

Perform Step: 3 Close the selected WGDT Outlet to Gas Release Header Valve (Rm 16):

2.19

• WD-132, WD-29A Standard: ROTATED WD-132, GAS DECAY TANK WD-29A OUTLET VALVE handwheel in CLOCKWISE direction until stopped.

Examiner Cue: Valve handwheel rotated then stopped. If asked, REPORT as Control Room Supervisor to continue in procedure.

Comment: SAT UNSAT Perform Step: 4 Close and lock the following Gas Release Header Isolation Valves 2.20 & 1st bullet (Rm 16):

• WD-150 Standard: PERFORMED the following:

• ROTATED WD-150, WASTE GAS DECAY TANKS WD-29A, B, C

& D GAS RELEASE HEADER ISOLATION VALVE handwheel in CLOCKWISE direction until stopped (critical).

• INSTALLED chain and LOCKED valve (NOT critical).

• INFORMED Control Room WD-150 is LOCKED per SO-O-44 (NOT critical).

Examiner Cue: Valve handwheel rotated then stopped.

If Control Room is contacted, ACKNOWLEDGE locking of WD-150.

Comment: SAT UNSAT Examiner Note: Valve located > 8 feet above the floor. May require a ladder that can be obtained from the Corridor (West) just beyond AI-100, and notification to RP that they are working above 7 ft.

Perform Step: 5 Close and lock the following Gas Release Header Isolation Valves 2.20 & 2nd bullet (Rm 16):

• WD-167 Standard: PERFORMED the following:

• If needed, OBTAINED a ladder, simulated notifying RP

• ROTATED WD-167, WASTE GAS DECAY TANKS WD-29A, B, C

& D GAS RELEASE HEADER ISOLATION VALVE handwheel in CLOCKWISE direction until stopped.

Examiner Cue: Valve handwheel rotated then stopped.

Comment: SAT UNSAT Page 4 of 6 NRC JPM P-3 Rev. Final

Appendix C JPM STEPS Form ES-C-1 Examiner Note: Valve located at floor level.

Perform Step: 6 Close WD-165, Gas Release Header Bypass Valve (Rm 16).

2.21 Standard: ROTATED WD-165, GAS RELEASE HEADER BYPASS VALVE handwheel in CLOCKWISE direction until stopped.

Terminating Cue: Valve handwheel rotated then stopped. This JPM is complete.

Comment: SAT UNSAT STOP TIME:

Page 5 of 6 NRC JPM P-3 Rev. Final

Appendix C JPM CUE SHEET Form ES-C-1 INITIAL CONDITIONS: Given the following conditions:

• A release from WD-29A, Waste Gas Decay Tank is in progress.

• RM-052, Auxiliary Building Ventilation Stack Radiation Monitor has gone into HIGH alarm.

• AOP-09, High Radioactivity is in progress.

• When HC-532, Waste Gas Release Control Switch was placed in CLOSE, FCV-532A and FCV-532C on AI-100 did NOT close.

• Independent Verification has been waived by the Shift Manager due to AOP-09 entry and high radiation in the Auxiliary Building.

INITIATING CUE: The Control Room Supervisor directs you to PERFORM the following:

• TERMINATE release from WD-29A per OI-WDG-2, Waste Gas Disposal System Release, Attachment 3, Manual Waste Gas Release with FE-532 Unavailable.

• START at Step 2.17.

Page 6 of 6 NRC JPM P-3 Rev. Final

Appendix D Scenario Outline Form ES-D-1 Facility: Fort Calhoun Station Scenario No.: 1 Op Test No.: Dec 2015 NRC Examiners: Operators:

Initial Conditions: 100% power MOL - RCS Boron is 482 ppm (by sample).

Turnover: Maintain steady-state power conditions. Chemistry requests two Charging Pumps be placed in service per OI-CH-1, CVCS System Normal Operation.

Critical Tasks:

• Manually Actuate Pressurizer Pressure Low Signal (PPLS) when RCS Pressure 1600 psia and before 1350 psia to ensure SIAS / VIAS / CIAS Activation. (Event 8)

• Stop All Reactor Coolant Pumps (RCPs) when Subcooling is approaching or is less than 20°F but before 0°F due to Loss of Net Positive Suction Head (NPSH) per RCP NPSH Curve. (Event 6)

• Commence a Cooldown and Depressurization of the Reactor Coolant System before Reactor Vessel Level Monitoring System (RVLMS) is less than 83%, indicating a bubble has formed in the head, to Reestablish RCS Inventory Control while maintaining RCS Heat Removal.

(Event 6)

Event No. Malf. No. Event Type* Event Description 1 N (ATCO) Raise Charging and Letdown Flow per OI-CH-1, CVCS System

+15 min Normal Operation, Attachment 3.

2 C (ATCO, CRS) Component Cooling Water (CCW) Pump Trip.

+25 min TS (CRS) Start Either Standby CCW Pump.

3 C (BOPO, CRS) Plant Air System Leak @ 15% on 2 minute ramp.

+35 min Start Instrument Air Compressors.

4 I (ATCO, CRS) Pressurizer Pressure Control Channel PT-103X Fails to 2150 psia

+45 min TS (CRS) on 15 Minute Ramp. Transfer Pressure Control to PT-103Y.

5 R (ATCO) Condenser Evacuation Pump Trip with Auto Start Failure.

+55 min C (BOPO, CRS) Partial Loss of Condenser Vacuum. Reduce Turbine Load.

6 M (ATCO, BOPO, Inadvertent Main Turbine Trip.

+55 min CRS) Pressurizer Safety Valve Fails 50% Open on Reactor Trip.

7 C (BOPO) Total Loss of Condenser Vacuum.

+55 min Place HCV-1040, Atmospheric Dump Valve in Service.

8 I (ATCO) Pressurizer Pressure Low Signal Actuation Failure.

+65 min Manually Initiate Safety Injection.

9 C (ATCO) Low Pressure Safety Injection (LPSI) Pumps Start Failure.

+65 min Manually Start LPSI Pumps.

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Technical Specifications Actual Target Quantitative Attributes 3 Malfunctions after EOP entry (1-2) 5 Abnormal events (2-4) 1 Major transients (1-2) 1 EOPs entered/requiring substantive actions (1-2) 0 EOP contingencies requiring substantive actions (0-2) 3 Critical tasks (2-3)

NRC Simulator Scenario 1 Outline Rev. Final

Scenario Event Description NRC Scenario 1 SCENARIO

SUMMARY

NRC 1 The crew will assume the shift at 100% power per OP-4, Load Change and Normal Power Operation.

The scheduled activity is to start a second Charging Pump per OI-CH-1, CVCS System Normal Operation, Attachment 3, Raising Charging and Letdown Flows per Chemistry request.

The next event is a Component Cooling Water Pump Trip with auto start failure of the standby pumps.

The crew enters AOP-11, Loss of Component Cooling Water, and restores flow by starting either CCW Pump AC-3A or AC-3B. The SRO will refer to Technical Specification LCO 2.4(1) - Component Cooling Water Pump.

The next event is a Plant Air System leak and entry into AOP-17, Loss of Instrument Air, is required.

Crew should recognize that the Control Room Standby Instrument Air Compressor is not loading (ammeter at 0) and start a 3rd Air Compressor. Procedure exit occurs when the Plant Air System is locally isolated from the Instrument Air System.

When plant conditions are stable, Pressurizer Pressure Control Channel, PT-103X, will fail to 2150 psia over 15 minutes. Operator actions are per ARP-CB-1/2/3/A4, Windows A-4 & B-4, PRESSURIZER PRESSURE OFF NORMAL HI-LO CHANNEL X & CHANNEL Y. The crew will transfer to the standby channel PT-103Y and restore Reactor Coolant System (RCS) pressure. The SRO will refer to Technical Specification LCO 2.10.4 - DNBR Margin during Power Operation above 15% of Rated Power.

The next event is a partial Loss of Condenser Vacuum. The crew enters AOP-26, Turbine Malfunctions,Section I, Loss of Vacuum. Actions include starting a Condenser Evacuation Pump and transitioning to AOP-05, Emergency Shutdown, to lower Turbine load and restore Condenser vacuum. When power has been reduced 3% to 5%, an inadvertent Main Turbine trip will occur.

The inadvertent Main Turbine trip results in lifting of a Pressurizer Safety Valve resulting in a Small Break Loss of Coolant Accident (Vapor Space LOCA). The crew enters EOP-00, Standard Post Trip Actions, and manually actuates Safety Injection when it is determined that a Pressurizer Pressure Low Signal Actuation failure has occurred. When Diagnostic Actions are completed at the end of EOP-00, a transition will be made to EOP-03, Loss of Coolant Accident. Two Reactor Coolant Pumps are secured while in EOP-00 when pressure drops to 1350 psia. Eventually all RCPs will be secured due to a loss of subcooling (< 20°F). Upon entry into EOP-03, Containment Cooling Fans VA-7C and VA-7D will need to be started. Containment pressure remains less than 3 psig throughout the event.

The event is complicated by total Loss of Condenser Vacuum which will require placing the Atmospheric Steam Dump Valve, HCV-1040 in service and manual starting of the Low Pressure Safety Injection Pumps due to an automatic start failure.

This scenario is terminated when a cooldown and depressurization is commenced while in EOP-03 using HR-12, Secondary Heat Removal Operation, and PC-11, Pressure Control.

Risk Significance:

• Failure of risk important system prior to trip: Loss of CCW Pump

• Risk significant core damage sequence: Small Break LOCA Safety Injection Actuation Failure

• Risk significant operator actions: Manually Actuate Safety Injection Stop RCPs Upon Loss of Subcooling Cooldown and Depressurize RCS NRC Simulator Scenario 1 Outline Rev. Final

Scenario Event Description NRC Scenario 1 BOOTH OPERATOR INSTRUCTIONS for SIMULATOR SETUP RESET to IC-108 and LOAD & EXECUTE NRC 1.sce for NRC Scenario 1.

Preset Item - Event 2 - Block Autostart of Non-running CCW Pumps Type Item Value Condition Expert CCAAFU_STDBY_AC_3BCC 1 Scenario Event: AC-3B (AC-3B standby fuse failure) Stbyfuse blown CCBPFU_STDBY_AC_3ACC 1 Scenario Event: AC-3A (AC-3A standby fuse failure) Stby Fuse blown Preset Item - Event 3 - Block Autostart of CA-1B Type Item Value Condition Remote REM:CA1B_3SS (CA-1B control Off (value = 3) Scenario Event: Block start selector switch) of CA-1B Preset Item - Event 5 - Block Auto Start of Condenser Evacuation Pump FW-8C Type Item Value Condition Expert CEACWL_CLTVSP Triggered Scenario Event: block start FW-8C Preset Item - Event 8 - PPLS Fail to Actuate Type Item Value Condition Malfunction ESF07 (PPLS Actuation - Train A) Block Scenario Event: PPLS auto ESF08 (PPLS Actuation - Train B) Block fail Preset Item - Event 9 - LPSI Pumps Fail to Automatically Start Type Item Value Condition Expert ESEARL62_2_1X_SI_1BTVSP Deenergized Scenario Event: LPSI fail ESEBRL62_2_2X_SI_1BTVSP Deenergized to start ESCBRL62_1_2X_SI_1ATVSP Deenergized ESCARL62_1_1X_SI_1ATVSP Deenergized Event 2 - CCW Pump AC-3C Trips Type Item Value Condition Malfunction BUS_1B3C_4C_4_BKR_TRIP trip When directed by examiner, (CCW pump AC-3C breaker fail to trigger/activate this event.

the trip position) Scenario Event: CCW Pump AC-3C Trip Event 3 - Plant Air Leak Type Item Value Condition Malfunction CAS02C (Plant Air Leak) 15 When directed by examiner, Ramp: 120 seconds trigger/activate this event.

Scenario Event: Plant Air Leak Remote REM:CAS_CA630 0 When directed to close CA-REM:CAS_PCV1753 0 121 to isolate the instrument air leak, trigger/activate this event. Scenario Event:

When directed to close CA121 NRC Simulator Scenario 1 Outline Rev. Final

Scenario Event Description NRC Scenario 1 Event 4 - Pressurizer Pressure Transmitter PT-103X Fails High Type Item Value Condition Transmitter RCS_PT103X 2150 When directed by examiner, Ramp: 900 seconds trigger/activate this event.

Scenario Event: PT-103x fail high Event 5 - Running Condenser Evacuation Pump Trips, Degrading Condenser Vacuum Type Item Value Condition Malfunction CES06 (Condenser Evacuation FW- Trip When directed by examiner, 8B Pump trips) trigger/activate this event.

CND01 (Loss of Main Condenser 3%, ramp = 60 sec Scenario Event: Cond Vacuum) Evac trip Event 6 - Inadvertent Trip, Pressurizer Safety Valve Opens Type Item Value Condition Remote REM:86-1/G1-TRP (relay 86-1/G1 Trip When directed by examiner, fail to trip position) trigger/activate this event.

REM: 86-2/G1-TRP (relay 86-2/G1 Trip Scenario Event: Trip, fail to trip position) safety valve open Malfunction RCS_RC141 (safety valve RC-141) After reactor trip, value = 50, ramp =

15 seconds, delay =

5 seconds Event 7 - Total Loss of Condenser Vacuum Type Item Value Condition Malfunction CND01 (Loss of Main Condenser 100%, 300 second 60 seconds after reactor trip, Vacuum) ramp automatically trigger/activate event:

Complete Loss of Cond Vacuum NRC Simulator Scenario 1 Outline Rev. Final

Scenario Event Description NRC Scenario 1 Booth Operator: INITIALIZE to IC-108 and LOAD NRC 1.sce.

ENSURE all Simulator Annunciator Alarms are ACTIVE.

ENSURE all Control Board Tags are removed.

ENSURE CH-1C, Charging Pump is running.

ENSURE AC-3C, Component Cooling Water Pump running.

ENSURE Channel X Pressurizer Pressure and Level selected.

ENSURE FW-8B, Condenser Evacuation Pump running.

ENSURE Reactivity Briefing Sheet printout provided with Turnover.

ENSURE Middle-of-Life Thumb Rule Sheet provided with Turnover.

ENSURE Containment Pressure Relief (CPR) is secured.

ENSURE procedures in progress provided to crew in Briefing Room:

- COPY of OI-CH-1, Chemical and Volume Control System Normal Operation, Attachment 3, Raising Charging and Letdown Flows, INITIALED through Step 2.i.

Control Room Annunciators in Alarm:

NONE 0B Procedure List Event 1: OP-4, Load Change and Normal Power Operation Event 1: OI-CH-1, CVCS System Normal Operation, Attachment 3, Raising Charging and Letdown Flows Event 2: AOP-11, Loss of Component Cooling Water Event 3: AOP-17, Loss of Instrument Air Event 4: ARP-CB-1/2/3/A4, Windows A-4 & B-4, PRESSURIZER PRESSURE OFF NORMAL HI-LO CHANNEL X & CHANNEL Y Event 5: AOP-26, Turbine Malfunctions,Section I, Loss of Vacuum Event 6: EOP-00, Standard Post Trip Actions Event 6: EOP-03, Loss of Coolant Accident Event 6: HR-12, Secondary Heat Removal Operation Event 6: PC-11, Pressure Control NRC Simulator Scenario 1 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 1 Page 6 of 29 Event

Description:

Raise Charging and Letdown Flow Time Position Applicants Actions or Behavior Booth Operator: When directed, RESPOND to requests from Control Room. REPORT back that plant conditions requested are normal unless otherwise scripted.

Indications Available:

NONE Examiner Note: The following steps are from OI-CH-1, Chemical and Volume Control System Normal Operation, Attachment 3, Raising Charging and Letdown Flows.

+1 min ATCO START the selected Charging Pump CH-1B. [Step 3]

• PLACE CH-1B switch to START.

NOTES

1. PIC-210 Letdown Press Cntrlr should be continuously monitored while adjusting letdown flow.

2. Steps 4 and 5 may be performed concurrently without the procedure in hand. Sign-offs may be completed after these steps are performed.

RAISE bias on HIC-101-1/101-2, Letdown Throttle Valves Controller, and ATCO OBSERVE an increase in Letdown flow. [Step 4]

• ROTATE HIC-101-1/101-2 in COUNTERCLOCKWISE direction to increase Letdown flow.

Examiner Note: It is acceptable to place Letdown Pressure Control and Flow Control in MANUAL or AUTOMATIC control during rotation of Charging Pumps.

ADJUST PIC-210, Letdown Press Controller as necessary to maintain ATCO Letdown pressure approximately 300 psig. [Step 5]

Continue to ADJUST bias on HIC-101-1/101-2 until Pressurizer level is ATCO STABILIZED at the programmed setpoint. [Step 6]

When Letdown flow is stable, PROCEED to Event 2.

NRC Simulator Scenario 1 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 2 Page 7 of 29 Event

Description:

Component Cooling Water Pump Trip Time Position Applicants Actions or Behavior Booth Operator: When directed, EXECUTE Event 2.

- AC-3C, CCW Pump trip.

Indications Available:

CB-1/2/3/A2 - CCW PUMPS TRIP CB-1/2/3/A2 - CC WATER FROM DISCH HEADER FLOW LO CB-1/2/3/A2 - CCW PUMPS AC-3A/B/C STANDBY START CB-1/2/3/A2 - AUXILIARY COOLANT FROM CRDM FLOW LO CCW Pump AC-3C white TRIP and green STOP lights lit Multiple loss of CCW flow alarms Booth Operator: When contacted for pump conditions, REPORT as Auxiliary Building Operator all conditions normal. REPORT as Water Plant Operator that breaker tripped on overcurrent.

+30 sec ATCO RESPOND to Annunciator Response Procedures.

ATCO INFORM CRS of CCW Pump AC-3C trip with NO auto start of standby pump.

Examiner Note: ATCO may Operate to Mitigate per OPD 3-01 and START a CCW Pump.

CRS REFER to AOP-11, Loss of Component Cooling Water.

Examiner Note: The following steps are from AOP-11, Loss of Component Cooling Water.

ATCO VERIFY normal CCW/RW System operation: [Step 4.1]

• START CCW Pump AC-3A or AC-3B. [Step 4.1.a]

• VERIFY CCW System pressure 60 psig. [Step 4.1.b]

• DETERMINE AC-1B, Raw Water CCW Heat Exchanger in service.

[Step 4.1.c]

• DETERMINE RCP Coolers CCW Valves, HCV-438A/B/C/D all OPEN.

[Step 4.1.d]

ATCO VERIFY Raw Water Pump operating. [Step 4.2]

ATCO If CCW Surge Tank level < 42 inches, FILL the CCW Surge Tank: [Step 4.3]

NRC Simulator Scenario 1 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 2 Page 8 of 29 Event

Description:

Component Cooling Water Pump Trip Time Position Applicants Actions or Behavior

• OPEN LCV-2801, CCW Surge Tank Makeup Valve, to refill CCW Surge Tank. [Step 4.3.a]

• PLACE LCV-2801 in CLOSE or AUTO. [Step 4.3.b]

CRS EVALUATE Technical Specification LCO 2.4, Containment Cooling

• LCO 2.4.(1).a - Component Cooling Water Pump AC-3C

• CONDITION 2.4.(1).a - Component Cooling Water Pump AC-3C inoperable.

• ACTION 2.4.(1).b - RESTORE Component Cooling Water Pump AC-3C within 7 days OR PLACE Reactor in HOT SHUTDOWN condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

When Technical Specifications are addressed, PROCEED to Event 3.

NRC Simulator Scenario 1 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 3 Page 9 of 29 Event

Description:

Plant Air System Leak Time Position Applicants Actions or Behavior Booth Operator: When directed, EXECUTE Event 3.

- Plant Air System leak @ 15% on 2 minute ramp.

Indications Available:

CB-10,11/A21 - PLANT AIR PRESS LO PI-1700, Plant Air Press lowering on CB-10,11

+30 sec BOPO RESPOND to Annunciator Response Procedures.

BOPO INFORM CRS of Plant Air System pressure less than 96 psig and lowering.

Examiner Note: BOPO may Operate to Mitigate per OPD 3-01 and START an Air Compressor.

CRS REFER to AOP-17, Loss of Instrument Air.

Examiner Note: The following steps are from AOP-17, Loss of Instrument Air.

BOPO ENSURE all available Air Compressors start. [Step 4.1]

• START Air Compressor CA-1A.

• CONTACT Auxiliary Operator to START Air Compressor CA-1B.

Examiner Note: Air Compressor CA-1B does NOT Auto Start from the Control Room. Control Board indications for CA-1B show the breaker closed but compressor is NOT running or loaded (observe CA-1B amps).

Booth Operator: If contacted, REPORT standby Air Compressor CA-1B switch alignment is normal.

CONTACT Equipment Operator to ensure proper operation of Instrument Air BOPO Compressors, Dryers, and Filters. [Step 4.2]

Booth Operator: If contacted, REPORT Compressors, Dryers, and Filters appear to be operating normally.

ANNOUNCE and REPEAT message using Plant Communication System:

CREW

[Step 4.3]

NRC Simulator Scenario 1 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 3 Page 10 of 29 Event

Description:

Plant Air System Leak Time Position Applicants Actions or Behavior

• "Attention all personnel, attention all personnel; there is a plant air leak in progress. Report any large air usage to the Control Room."

CRS DIRECT available operators to search for source of air leakage. [Step 4.4]

Booth Operator: When contacted, REPORT leak is downstream of PCV-1753. When directed, EXECUTE remote function to isolate leak and report CA-121, Service Air Supply System Manual Isolation Valve is CLOSED.

DETERMINE Instrument Air pressure is < 80 psig, and CONTACT BOPO Equipment Operator to VERIFY PCV-1753, Service Air System Automatic Isolation Valve CLOSED. [Step 4.5]

DETERMINE Instrument Air pressure slowly returning to normal after service CRS air was isolated. [Step 4.6]

• VERIFY CA-121, Service Air Supply System Manual Isolation Valve is closed. [Step 4.6.a]

• GO TO Section 5.0, Exit Conditions. [Step 4.6.b]

Examiner Note: Plant Air System remains isolated for the duration of the Scenario.

When Instrument Air pressure returns to normal, PROCEED to Event 4.

NRC Simulator Scenario 1 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 4 Page 11 of 29 Event

Description:

Pressurizer Pressure Control Channel Failure Time Position Applicants Actions or Behavior Booth Operator: When directed, EXECUTE Event 4.

- Pressurizer Pressure Control Channel PT-103X fails to 2150 psia on 15 minute ramp.

Indications Available:

CB-1/2/3/A4 - PRESSURIZER PRESSURE OFF NORMAL HI-LO CHANNEL Y (1st alarm)

CB-1/2/3/A4 - PRESSURIZER PRESSURE OFF NORMAL HI-LO CHANNEL X (2nd alarm ~ 2 min later)

Examiner Note: Due to the nature of this failure, Channel Y alarm comes in 1st as it senses PZR pressure < 2080 psia (alarm setpoint) even though Channel X is the Controlling Channel. As the Channel X setpoint failure ramps in and reaches

> 2145 psia (alarm setpoint), Channel X annunciator will alarm.

+30 sec ATCO RESPOND to Annunciator Response Procedures.

Examiner Note: ATCO may Operate to Mitigate per OPD 4-09 and TRANSFER to Channel Y.

Examiner Note: The following steps are from ARP-CB-1/2/3/A4, Window A-4 for Channel X.

ATCO VERIFY RCS pressure using all available indications. [Step 1]

• MONITOR Pressurizer Pressure and operation of PC-103X. [Step 1.1]

• DETERMINE PC-103X is not controlling pressure and PLACE HC-103, Pressurizer Pressure Channel Selector Switch to CHAN Y position. [Step 1.1.1]

• MAY REFER to OI-RC-7, Reactor Coolant System Pressure Control CRS Normal Operation. [Step 1.1.2]

ATCO PERFORM the following for the low pressure condition: [Step 2]

• REFER to Technical Specification LCO 2.10.4.(5) if pressure 2075 CRS psia. [Step 2.1]

• DETERMINE Pressurizer Spray Valves PCV-103-1 and PCV-103-2 are ATCO CLOSED. [Step 2.2]

• ENSURE all Pressurizer Heater Control Switches in AUTO or ON.

ATCO

[Step 2.3]

ATCO

• ENERGIZE additional Pressurizer Heaters as required. [Step 2.4]

• DETERMINE Pressurizer level NOT lowering on LR-101X/LR-101Y.

ATCO

[Step 2.5]

NRC Simulator Scenario 1 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 4 Page 12 of 29 Event

Description:

Pressurizer Pressure Control Channel Failure Time Position Applicants Actions or Behavior ATCO

• VERIFY VCT level trend on LI-219. [Step 2.6]

CRS EVALUATE Technical Specification LCO 2.10.4, Power Distribution Limits

• LCO 2.10.4.(5) - DNBR Margin during Power Operation above 15% of Rated Power

• CONDITION 2.10.4.(5).(a).(ii) - Pressurizer Pressure < 2075 psia.

• ACTION 2.10.4.(5).(b) - RESTORE Pressurizer Pressure within its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or REDUCE power to less than 15% of rated power within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

When Technical Specifications have been addressed, PROCEED to Event 5.

NRC Simulator Scenario 1 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 5 Page 13 of 29 Event

Description:

Partial Loss of Condenser Vacuum / Condenser Evacuation Pump Trip With Auto Start Failure Time Position Applicants Actions or Behavior Booth Operator: When directed, EXECUTE Event 5

- Partial Loss of Condenser vacuum @ 5% on 3 minute ramp.

- Condenser Evacuation Pump FW-8B trip.

- Condenser Evacuation Pump FW-8C Auto Start failure.

Indications Available:

CB-10,11/A9 - VACUUM PUMP B STOPPED OR SEAL WATER TEMP HI Emergency Response Facility Computer System (ERFCS) Alarm on Low Condenser Vacuum Condenser Evacuation Pump FW-8B green STOP light lit Lowering Condenser Vacuum on PI-925A/B or P0976A/B Examiner Note: Rate of lowering Condenser vacuum may be modified at your discretion to advance or retard the pace of this and the next event.

+30 sec BOPO RESPOND to Annunciator Response Procedures.

INFORM CRS of lowering Condenser vacuum and Condenser Evacuation BOPO Pump FW-8B trip.

Examiner Note: BOPO may Operate to Mitigate per OPD 4-09 and START FW-8C.

CRS REFER to AOP-26, Turbine Malfunctions,Section I, Loss of Vacuum.

Examiner Note: The following steps are from AOP-26, Turbine Malfunctions,Section I, Loss of Vacuum.

MONITOR Condenser vacuum on ERF Computer System/PI-925A/

BOPO PI-925B/P0976A/P0976B. [Step 4.1]

BOPO ENSURE all Condenser Evacuation Pumps are running. [Step 4.2]

• START FW-8C, Condenser Evacuation Pump.

NRC Simulator Scenario 1 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 5 Page 14 of 29 Event

Description:

Partial Loss of Condenser Vacuum / Condenser Evacuation Pump Trip With Auto Start Failure Time Position Applicants Actions or Behavior CAUTION The Turbine should not be operated with a Generator load of less than 150 MW when vacuum is less than or equal to 23.85" Hg (ERF, P0976A/B) or 6.07" Hg absolute (PI-925A/B) due to possible overheating of final stage blades.

If Condenser vacuum is < 25" Hg or 4.92" Hg Absolute, COMMENCE a plant CRS shutdown to restore vacuum per AOP-05 Emergency Shutdown. [Step 4.3]

Examiner Note: The following steps are from AOP-05, Emergency Shutdown.

NOTE TDB-III-23a and the Power Ascension/Power Reduction Strategy (PAPRs) provide guidance for the shutdown.

CRS CONTACT Reactor Engineer if additional guidance is required. [Step 4.1]

NOTE Operation of more than one Charging Pump will raise the rate of the power reduction.

Examiner Note: Unless directed, boration will occur from the Safety Injection Refueling Water Tank (SIRWT) when in AOP-05 to avoid time constraints.

If borating from SIRWT, COMMENCE boration by performing the following:

CRS

[Step 4.2]

• DETERMINE Charging Pumps CH-1B & CH-1C RUNNING.

ATCO

[Step 4.2.a]

• OPEN LCV-218-3, Charging Pump Suction SIRWT Isolation Valve.

ATCO

[Step 4.2.b]

ATCO

• CLOSE LCV-218-2, VCT Outlet Valve. [Step 4.2.c]

CRS DETERMINE Boration alignment from CVCS NOT required. [Step 4.3]

CRS NOTIFY Energy Marketing of power reduction. [Step 4.4]

NRC Simulator Scenario 1 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 5 Page 15 of 29 Event

Description:

Partial Loss of Condenser Vacuum / Condenser Evacuation Pump Trip With Auto Start Failure Time Position Applicants Actions or Behavior NOTE During the power reduction, maintain TC PER TDB Figure III.1, Tave Program.

MAINTAIN RCS Temperature Control via Turbine Load per HR-12, BOPO Secondary Heat Removal Operation: [Step 4.5]

• MAINTAIN TCOLD 527°F to 547°F AND

• MAINTAIN TCOLD +0°F to -1°F of program.

MAINTAIN Pressurizer Level via Charging and Letdown per IC-11, Inventory ATCO Control: [Step 4.6]

• MAINTAIN Pressurizer Level 45% to 60% AND

• MAINTAIN Pressurizer Level within 4% of program.

When power has been lowered 3% to 5%, PROCEED to Events 6, 7, 8, and 9.

NRC Simulator Scenario 1 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 6, 7, 8, & 9 Page 16 of 29 Event

Description:

Inadvertent Main Turbine Trip / Pressurizer Safety Valve Failure / Loss of Condenser Vacuum /

Pressurizer Pressure Low Signal Actuation Failure / Low Pressure Safety Injection Pumps Start Failure Time Position Applicants Actions or Behavior Booth Operator: When directed, EXECUTE Events 6, 7, 8, and 9.

- Inadvertent Turbine Trip.

- Pressurizer Safety Valve fails 50% open on Reactor Trip.

- Loss of Condenser Vacuum @ 100%.

- Pressurizer Pressure Low Signal (PPLS) Actuation failure.

- Low Pressure Safety Injection Pumps start failure.

Indications Available:

Numerous Reactor Trip and Turbine Trip Alarms.

+10 sec ATCO RECOGNIZE Reactor Trip due to Turbine Trip.

CRS DIRECT performance of EOP-00, Standard Post Trip Actions.

Examiner Note: The following steps are from EOP-00, Standard Post Trip Actions.

ATCO VERIFY Reactivity Control: [Step 1]

• VERIFY ALL of the following: [Step 1.a]

• VERIFY no more than one Regulating or Shutdown CEA NOT inserted.

• VERIFY Reactor Power is LOWERING.

• VERIFY Startup Rate is NEGATIVE.

• MONITOR plant for an uncontrolled RCS Cooldown. [Step 1.b]

CRS DETERMINE Reactivity Control criteria SATISFIED.

BOPO VERIFY Turbine Trip: [Step 2]

• VERIFY HP & LP Stop and Intercept Valves CLOSED.

BOPO ENSURE all Generator Breakers are tripped: [Step 3]

• DETERMINE Generator Output Breaker 3451-4 tripped.

• DETERMINE Generator Output Breaker 3451-5 tripped.

• DETERMINE Generator Field Breaker 41E/G1F tripped.

BOPO VERIFY 4160 V Safeguards Buses 1A3 and 1A4 are ENERGIZED. [Step 4]

NRC Simulator Scenario 1 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 6, 7, 8, & 9 Page 17 of 29 Event

Description:

Inadvertent Main Turbine Trip / Pressurizer Safety Valve Failure / Loss of Condenser Vacuum /

Pressurizer Pressure Low Signal Actuation Failure / Low Pressure Safety Injection Pumps Start Failure Time Position Applicants Actions or Behavior CRS DETERMINE Maintenance of Vital Auxiliaries criteria SATISFIED.

Examiner Note: Diesel Generators only start after Safeguards (PPLS) actuation.

VERIFY both Diesel Generators RUNNING on Safety Injection Actuation BOPO Signal. [Step 5]

VERIFY 4160 V Non-Safeguards Buses 1A1 and 1A2 are ENERGIZED.

BOPO

[Step 6]

BOPO VERIFY 125 VDC Buses 1 and 2 are ENERGIZED. [Step 7]

BOPO VERIFY Instrument Air is AVAILABLE: [Step 8]

• DETERMINE Instrument Air pressure 90 psig.

• DETERMINE Instrument Air Compressor CA-1A RUNNING.

ATCO VERIFY Component Cooling Water System operation NORMAL: [Step 9]

• DETERMINE at least one CCW pump RUNNING. [Step 9.a]

• DETERMINE CCW Pump discharge pressure 60 psig. [Step 9.b]

• DETERMINE HCV-438A/B/C/D, CCW to RCP Coolers OPEN. [Step 9.c]

• DETERMINE at least one Raw Water Pump RUNNING. [Step 9.d]

CRS VERIFY RCS Inventory Control criteria satisfied: [Step 10]

• DETERMINE PZR level NOT between 30% and 70% and NOT ATCO TRENDING to between 45% and 60%.

• DETERMINE RCS subcooling 20°F:

• [CA] RESTORE Inventory Control by manually controlling Charging and Letdown. [Step 10.1.a]

CRS DETERMINE RCS Inventory Control criteria NOT SATISFIED.

CRS VERIFY RCS Pressure Control criteria satisfied: [Step 11]

NRC Simulator Scenario 1 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 6, 7, 8, & 9 Page 18 of 29 Event

Description:

Inadvertent Main Turbine Trip / Pressurizer Safety Valve Failure / Loss of Condenser Vacuum /

Pressurizer Pressure Low Signal Actuation Failure / Low Pressure Safety Injection Pumps Start Failure Time Position Applicants Actions or Behavior

• DETERMINE RCS pressure NOT between 1800 psia and 2300 psia and ATCO NOT trending to between 2050 psia and 2150 psia.

• [CA] DETERMINE RCS pressure < 2300 psia and PORV NOT open.

[Step 11.1]

• [CA] DETERMINE RCS pressure 1350 psia and TRIP one RCP in each loop. [Step 11.2]

• [CA] DETERMINE RCS pressure 1600 psia and ENSURE PPLS actuated. [Step 11.3]

Manually Actuate Pressurizer Pressure Low Signal (PPLS) when RCS Pressure CRITICAL TASK 1600 psia and before 1350 psia to ensure SIAS / VIAS / CIAS Activation.

STATEMENT Pressure at Time of PPLS Trip: ______ psia.

CRITICAL TASK ATCO DETERMINE PPLS relays NOT tripped and manually ACTUATE PPLS.

• [CA] INSERT and TURN keys at 86A/PPLS Test Switch & 86B/PPLS ATCO Test Switch on AI-30A & AI-30B. [Step 11.3.a]

• [CA] DETERMINE PPLS relays 86A/PPLS / 86B/PPLS / 86A1/PPLS /

86B1/PPLS have TRIPPED. [Step 11.3.b]

• [CA] DETERMINE VIAS relays 86A/VIAS / 86B1/VIAS / 86B/VIAS /

86A1/VIAS have TRIPPED. [Step 11.3.c]

• [CA] DETERMINE SIAS relays 86A/SIAS / 86AX/SIAS / 86B1/SIAS /

86B1X/SIAS / 86B/SIAS / 86BX/SIAS / 86A1/SIAS / 86A1X/SIAS have TRIPPED. [Step 11.3.d]

• [CA] DETERMINE CIAS relays 86A/CIAS / 86B1/CIAS / 86B/CIAS /

86A1/CIAS have TRIPPED. [Step 11.3.e]

• [CA] ENSURE HPSI / LPSI / Charging Pumps RUNNING. [Step 11.3.f]

• DETERMINE HPSI Pumps SI-2A & SI-2B or SI-2B & SI-2C RUNNING.

• DETERMINE LPSI Pumps NOT RUNNING and manually ATCO START SI-1A and SI-1B.

• DETERMINE Charging Pumps CH-1A, CH-1B, & CH-1C RUNNING.

• [CA] ENSURE adequate SI flow per IC-13, Safety Injection Flow vs.

Pressurizer Pressure. [Step 11.3.g]

• [CA] DETERMINE Emergency Boration in progress. [Step 11.3.h]

NRC Simulator Scenario 1 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 6, 7, 8, & 9 Page 19 of 29 Event

Description:

Inadvertent Main Turbine Trip / Pressurizer Safety Valve Failure / Loss of Condenser Vacuum /

Pressurizer Pressure Low Signal Actuation Failure / Low Pressure Safety Injection Pumps Start Failure Time Position Applicants Actions or Behavior CRS DETERMINE RCS Pressure Control criteria NOT SATISFIED.

Examiner Note: The following steps are from RC-11, Emergency Boration Verification.

ATCO ENSURE the following valves are CLOSED: [Step 1]

• FCV-269X, Demin Water Makeup Valve

• FCV-269Y, Boric Acid Makeup Valve

• HCV-264, CH-4A Recirc Valve

• HCV-257, CH-4B Recirc Valve ATCO VERIFY all the following valves OPEN: [Step 2]

• HCV-268, Boric Acid Pump Header to Charging Pumps Isolation Valve

• HCV-265, CH-11A Gravity Feed Valve

• HCV-258, CH-11B Gravity Feed Valve ATCO ENSURE all available Boric Acid Pumps RUNNING: [Step 3]

• CH-4A, Boric Acid Pump

• CH-4B, Boric Acid Pump ATCO ENSURE all available Charging Pumps RUNNING: [Step 4]

• CH-1A, Charging Pump

• CH-1B, Charging Pump

• CH-1C, Charging Pump ATCO ENSURE the following valves are CLOSED: [Step 5]

• LCV-218-2, VCT Outlet Valve

• LCV-218-3, Charging Pump Suction SIRWT Isolation Valve Examiner Note: The following steps continue from EOP-00, Standard Post Trip Actions.

CRS VERIFY Core Heat Removal criteria satisfied: [Step 12]

NRC Simulator Scenario 1 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 6, 7, 8, & 9 Page 20 of 29 Event

Description:

Inadvertent Main Turbine Trip / Pressurizer Safety Valve Failure / Loss of Condenser Vacuum /

Pressurizer Pressure Low Signal Actuation Failure / Low Pressure Safety Injection Pumps Start Failure Time Position Applicants Actions or Behavior ATCO

• DETERMINE RCP NPSH requirements met.

• DETERMINE at least one RCP operating.

• DETERMINE Core T 10°F.

Examiner Note: Depending on Crew actions, RCS subcooling will be lost in either EOP-00, SPTAs or EOP-03, LOCA.

Stop All Reactor Coolant Pumps (RCPs) when Subcooling is approaching or is CRITICAL TASK less than 20°F but before 0°F due to Loss of Net Positive Suction Head (NPSH)

STATEMENT per RCP NPSH Curve.

Subcooling at Time of RCP Trip: ______ °F.

CRITICAL TASK ATCO DETERMINE RCP subcooling < 20°F and PERFORM the following:

ATCO

• STOP all RCPs.

• [CA] PLACE TCV-909, Temperature Controller in MANUAL on DCS.

BOPO

[Step 12.2.a]

• [CA] ENSURE TCV-909, Temperature Controller OUTPUT is zero BOPO (0). [Step 12.2.b]

CRS * [CA] VERIFY Natural Circulation in at least one Loop. [Step 12.2.c]

• [CA] DETERMINE Core T 50°F.

• [CA] DETERMINE difference between CETs and RCS THOT is 10°F on ERF "CHR" display.

• [CA] DETERMINE RCS subcooling is 20°F.

• [CA] DETERMINE THOT and TCOLD are stable or lowering.

CRS DETERMINE Core Heat Removal criteria NOT SATISFIED.

NOTE If Instrument Air to valves HCV-1105, HCV-1106, HCV-1107A/B and HCV-1108A/B is not available, throttling of these valves is not possible.

Open or close operation of these valves is possible for a minimum of three cycles.

CRS VERIFY RCS Heat Removal criteria satisfied:

NRC Simulator Scenario 1 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 6, 7, 8, & 9 Page 21 of 29 Event

Description:

Inadvertent Main Turbine Trip / Pressurizer Safety Valve Failure / Loss of Condenser Vacuum /

Pressurizer Pressure Low Signal Actuation Failure / Low Pressure Safety Injection Pumps Start Failure Time Position Applicants Actions or Behavior VERIFY Main Feedwater is restoring SG levels to 35% to 80% NR and 73%

BOPO to 94% WR. [Step 13]

• DETERMINE FCV-1101 & FCV-1102 Feed Regulating Valves CLOSED.

[Step 13.a]

• DETERMINE HCV-1105 & HCV-1106 Feed Regulating Bypass Valves ramped to between 40% & 45% OPEN. [Step 13.b]

BOPO

• PLACE 43/FW switch in OFF. [Step 13.c]

• ENSURE no more than one Main Feedwater Pump RUNNING.

[Step 13.d]

• ENSURE no more than one Condensate Pump RUNNING. [Step 13.e]

• STOP running Heater Drain Pumps FW-5A, FW-5B, and/or FW-5C.

BOPO

[Step 13.f]

• ENSURE both sets of SG Blowdown Isolation Valves CLOSED.

[Step 13.g]

VERIFY Steam Dump and Bypass Valves controlling both of the following:

BOPO

[Step 14]

• VERIFY RCS TCOLD between 525°F and 535°F.

• VERIFY Steam Generator pressure between 850 psia & 925 psia.

• [CA] DETERMINE loss of Condenser vacuum and PLACE HCV-1040, BOPO Atmosphere Dump Valve in service.

• SELECT HCV-1040 on DCS Secondary Screen.

CRS DETERMINE RCS Heat Removal criteria SATISFIED.

CRS VERIFY Containment Isolation criteria satisfied:

ATCO VERIFY Normal Containment conditions exist: [Step 15]

• DETERMINE unexpected rise in Containment Sump level. [Step 15.a]

• DETERMINE Containment Area Radiation Monitors in ALARM.

ATCO

[Step 15.b]

• DETERMINE Containment Ventilation Radiation Monitors RM-050 and ATCO RM-051 in ALARM. [Step 15.c]

• [CA] ENSURE VIAS has ACTUATED and 86A/VIAS, 86A1/VIAS, 86B/VIAS, & 86B1/VIAS relays TRIPPED.

NRC Simulator Scenario 1 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 6, 7, 8, & 9 Page 22 of 29 Event

Description:

Inadvertent Main Turbine Trip / Pressurizer Safety Valve Failure / Loss of Condenser Vacuum /

Pressurizer Pressure Low Signal Actuation Failure / Low Pressure Safety Injection Pumps Start Failure Time Position Applicants Actions or Behavior

• [CA] ENSURE RM-050 & RM-051 Containment Radiation Monitor Sample Pump STOPPED.

• [CA] ENSURE RM-065, Post Accident Control Room Iodine Monitor RUNNING.

• DETERMINE SG Blowdown & Condenser Off Gas Radiation Monitors (RM-054A / RM-054B / RM-057) NOT in alarm. [Step 15.d]

• DETERMINE SG Blowdown & Condenser Off Gas Radiation Monitors (RM-054A / RM-054B / RM-057) NOT trending to alarm. [Step 15.e]

ATCO

• VERIFY Containment conditions: [Step 15.f]

• DETERMINE Containment pressure < 3 psig.

• DETERMINE Containment temperature > 120°F.

CRS DETERMINE Containment Integrity criteria NOT SATISFIED.

CRS DIAGNOSE event in progress: [Step 16]

• DETERMINE Reactivity Control requirements met.

• DETERMINE both DC buses energized.

• DETERMINE at least one Vital 4160 V Bus energized.

• DETERMINE at least one Non-Vital 4160 V Bus energized.

• VERIFY at least one RCP running.

• If not, CONSIDER EOP-02, Loss of Offsite Power/Forced Circulation.

• DETERMINE adequate Feedwater flow to at least one SG.

• VERIFY Pressurizer pressure > 1800 psia with high subcooled margin, normal SG pressure, and no indications of primary to secondary leakage.

• If not, CONSIDER EOP-03, Loss of Coolant Accident.

NOTE Certain events (i.e., LOCA, SGTR, UHE and Loss of All Feedwater) do not require offsite power in order to adequately, mitigate the effects of the accident. For this reason, the LOCA, SGTR, UHE or Loss of All Feedwater procedure may be implemented even if a Loss of Offsite Power has also occurred.

• DETERMINE all Safety Function Acceptance Criteria NOT SATISFIED.

NRC Simulator Scenario 1 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 6, 7, 8, & 9 Page 23 of 29 Event

Description:

Inadvertent Main Turbine Trip / Pressurizer Safety Valve Failure / Loss of Condenser Vacuum /

Pressurizer Pressure Low Signal Actuation Failure / Low Pressure Safety Injection Pumps Start Failure Time Position Applicants Actions or Behavior

• DETERMINE single event in progress and transition to EOP-03, Loss of Coolant in Accident.

Examiner Note: The following steps are from EOP-03, Loss of Coolant Accident.

CRS CONFIRM Standard Post Trip Actions have been performed. [Step 1]

CRS CONFIRM Loss of Coolant Accident Diagnosis: [Step 2]

• VERIFY Safety Function Status Check Acceptance Criteria being satisfied. [Step 2.a]

• VERIFY CIAS is NOT present and SAMPLE both SGs. [Step 2.b]

CRS IMPLEMENT the Emergency Plan. [Step 3]

• Time: __________

NOTE Floating Step BB, Minimizing DC Loads, requires operator action within 15 minutes of loss of either battery charger.

CREW MONITOR the Floating Steps. [Step 4]

DETERMINE RCS pressure 1600 psia and Containment pressure 5 psig CRS and CSAS NOT present. [Step 5]

DETERMINE RCS pressure 1600 psia and VERIFY Engineered ATCO Safeguards Actuation: [Step 6]

• DETERMINE PPLS relays 86A/PPLS / 86B/PPLS / 86A1/PPLS /

86B1/PPLS have TRIPPED. [Step 6.a]

• DETERMINE VIAS relays 86A/VIAS / 86B1/VIAS / 86B/VIAS /

86A1/VIAS relays TRIPPED. [Step 6.b]

• DETERMINE SIAS relays 86A/SIAS / 86AX/SIAS / 86B1/SIAS /

86B1X/SIAS / 86B/SIAS / 86BX/SIAS / 86A1/SIAS / 86A1X/SIAS have TRIPPED. [Step 6.c]

• DETERMINE CIAS relays 86A/CIAS / 86B1/CIAS / 86B/CIAS /

86A1/CIAS relays TRIPPED. [Step 6.d]

NRC Simulator Scenario 1 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 6, 7, 8, & 9 Page 24 of 29 Event

Description:

Inadvertent Main Turbine Trip / Pressurizer Safety Valve Failure / Loss of Condenser Vacuum /

Pressurizer Pressure Low Signal Actuation Failure / Low Pressure Safety Injection Pumps Start Failure Time Position Applicants Actions or Behavior CRS DETERMINE Containment pressure 5 psig. [Step 7]

ATCO DETERMINE SIAS actuated and OPTIMIZE SI flow. [Step 8]

• ENSURE HPSI / LPSI / Charging Pumps RUNNING. [Step 8.a]

• DETERMINE HPSI Pumps SI-2A & SI-2B RUNNING.

• DETERMINE LPSI Pumps SI-1A and SI-1B RUNNING.

• DETERMINE Charging Pumps CH-1A, CH-1B, & CH-1C RUNNING.

• DETERMINE Emergency Boration in progress per RC-11, Emergency Boration Verification. [Step 8.b]

• DETERMINE adequate SI flow per IC-13, Safety Injection Flow vs.

Pressurizer Pressure. [Step 8.c]

CRS VERIFY RCP operating parameters: [Step 9]

ATCO

• ENSURE at least one RCP stopped if TCOLD < 500°F. [Step 9.a]

• ENSURE one RCP stopped in each loop if RCS pressure 1350 psia.

ATCO

[Step 9.b]

• ENSURE all RCPs STOPPED if RCS pressure < NPSH requirements ATCO per PC-12, RCS Pressure-Temperature Limits. [Step 9.c]

CRS RECORD time of SIAS initiation. [Step 10]

• Time: __________

VERIFY normal Component Cooling Water (CCW) and Raw Water (RW)

ATCO System operation: [Step 11]

• ENSURE at least 2 CCW Pumps RUNNING. [Step 11.a]

• VERIFY CCW Pump discharge pressure 60 psig. [Step 11.b]

• ENSURE at least 2 RW Pumps RUNNING. [Step 11.c]

• ENSURE at least 3 CCW Heat Exchangers in service. [Step 11.d]

• ENSURE all RCP Coolers CCW Valves OPEN. [Step 11.e]

NRC Simulator Scenario 1 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 6, 7, 8, & 9 Page 25 of 29 Event

Description:

Inadvertent Main Turbine Trip / Pressurizer Safety Valve Failure / Loss of Condenser Vacuum /

Pressurizer Pressure Low Signal Actuation Failure / Low Pressure Safety Injection Pumps Start Failure Time Position Applicants Actions or Behavior NOTE Do NOT isolate a PORV if the pressurizer is water solid.

ATCO VERIFY PORVs and PZR Code Safety Valves are CLOSED. [Step 12]

• DETERMINE Quench Tank temperature, pressure, and level in ALARM.

[Step 12.a]

• DETERMINE PZR Safety Valve discharge temperature high in ALARM.

[Step 12.b]

ATCO

• NOTIFY CRS that a PZR Safety Valve is OPEN.

• DETERMINE PORV Acoustic Flow Alarms are CLEAR. [Step 12.c]

NOTE Rising Radiation Monitor RM-053 count rate, rising CCW surge tank level or rising CCW surge tank pressure may be indications of a RCS-to-CCW leak.

ATCO DETERMINE RCS to CCW leak is NOT in progress. [Step 13]

CRS DETERMINE LOCA is inside Containment. [Step 14]

ATCO PERFORM the following for a LOCA inside Containment: [Step 15]

• PLACE HC-504A, CNTMT SUMP PUMP WD-3A CONTROL SWITCH, in PULL-TO-LOCK. [Step 15.a]

• PLACE HC-504B, CNTMT SUMP PUMP WD-3B CONTROL SWITCH, in PULL-TO-LOCK. [Step 15.a]

• CLOSE HCV-506A, Containment Sump Isolation Valve. [Step 15.b]

• CLOSE HCV-506B, Containment Sump Isolation Valve. [Step 15.b]

ATCO VERIFY all the following conditions exist: [Step 16]

• DETERMINE all HPSI Pumps are operating.

• DETERMINE SI flowrate is acceptable per IC-13 SI Flow vs. PZR Pressure.

• DETERMINE Representative CET temperature less than superheat.

• DETERMINE Reactor Vessel Level Monitoring System > 43% and NOT lowering.

NRC Simulator Scenario 1 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 6, 7, 8, & 9 Page 26 of 29 Event

Description:

Inadvertent Main Turbine Trip / Pressurizer Safety Valve Failure / Loss of Condenser Vacuum /

Pressurizer Pressure Low Signal Actuation Failure / Low Pressure Safety Injection Pumps Start Failure Time Position Applicants Actions or Behavior ATCO ENSURE SI-2C, HPSI Pump Control Switch in PULL-TO-LOCK.

ATCO DETERMINE NONE of the following conditions exist: [Step 17]

• SI flowrate is less than IC-13 SI Flow vs. PZR Pressure.

• Representative CET temperature greater than superheat.

• Reactor Vessel Level Monitoring System < 43% and lowering.

CRS DETERMINE RCS leak is NOT isolated. [Step 18]

DETERMINE Steam Generator Isolation Signal (SGIS) NOT actuated.

BOPO

[Step 19]

DETERMINE SG levels between 35% and 85% NR using Main Feedwater.

BOPO

[Step 20]

• MAINTAIN Feedwater flow per HR-15, Main Feed Pump Operation.

[Step 20.a]

• CONTROL Feedwater flow per HR-11, Manual Feet Control (DCS).

[Step 20.b]

CAUTION Failure to place the Containment Spray Pumps to Pull to Lock may allow actuation of Spray into Containment. This can lead to Containment Sump Blockage.

ATCO SECURE all Containment Spray flow: [Step 21]

• PLACE SI-3A, Containment Spray Pump in PULL-TO-LOCK.

• PLACE SI-3B, Containment Spray Pump in PULL-TO-LOCK.

• PLACE SI-3C, Containment Spray Pump in PULL-TO-LOCK.

CAUTION

1) When TC is 178°F or greater, the maximum RCS cooldown rate is 100°F/hr. When TC is less than 178°F, the maximum RCS cooldown rate is 50°F/hr.

2) No more than three RCPs shall be in operation when RCS temperature is less than 500°F.

COMMENCE a Steam Generator cooldown per HR-12, Secondary Heat CRS Removal Operation. [Step 22]

NRC Simulator Scenario 1 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 6, 7, 8, & 9 Page 27 of 29 Event

Description:

Inadvertent Main Turbine Trip / Pressurizer Safety Valve Failure / Loss of Condenser Vacuum /

Pressurizer Pressure Low Signal Actuation Failure / Low Pressure Safety Injection Pumps Start Failure Time Position Applicants Actions or Behavior

• Time: __________

MAINTAIN RCS pressure per PC-12, Pressure-Temperature Limits.

CRS

[Step 23]

• CONTROL RCS heat removal per HR-12, Secondary Heat Removal BOPO Operation. [Step 23.a]

ATCO

• CONTROL RCS pressure per PC-11, Pressure Control. [Step 23.b]

• If HPSI Stop and Throttle criteria are met, CONTROL Charging, ATCO Letdown, and HPSI flow per IC-11, Inventory Control. [Step 23.c]

NOTE Voiding of the RCS is indicated by the inability to depressurize to SDC entry pressure.

Attachment IC-14, RCS Void Elimination, provides guidance to correct this condition.

COMMENCE depressurizing RCS to 300 psia using any of the following CRS per PC-11, Pressure Control: [Step 24]

• CONTROL Pressurizer Spray flow.

• CONTROL Charging and Letdown flow.

• THROTTLE HPSI Pumps.

• Time: __________

Commence a Cooldown and Depressurization of the Reactor Coolant System before Reactor Vessel Level Monitoring System (RVLMS) is less than 83%,

CRITICAL TASK indicating a bubble has formed in the head, to Reestablish RCS Inventory STATEMENT Control while maintaining RCS Heat Removal.

RVLMS at start of Cooldown: ______

CRITICAL IMPLEMENT HR-12, Secondary Heat Removal Operation, to lower RCS TASK BOPO temperature.

Examiner Note: The following steps are from HR-12, Secondary Heat Removal Operation.

BOPO ENSURE Turbine Control is in MANUAL. [Step 1]

BOPO * [CA] DETERMINE Turbine NOT online and GO TO Step 4.

NRC Simulator Scenario 1 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 6, 7, 8, & 9 Page 28 of 29 Event

Description:

Inadvertent Main Turbine Trip / Pressurizer Safety Valve Failure / Loss of Condenser Vacuum /

Pressurizer Pressure Low Signal Actuation Failure / Low Pressure Safety Injection Pumps Start Failure Time Position Applicants Actions or Behavior NOTES

1. In MANUAL, single arrows are 1% / double arrows are 5% change in OUTPUT value.

2. While Steam Dump and Bypass Control is in Temperature/Pressure Mode, the controllers PC0910 and TC0909_PI will alternate between controls, depending on the higher output signal. A red square outlining the controlling function signify which parameter is in control.

3. Steps 4 through 11 may be performed as needed, and in any order.

CAUTIONS When TC is 178°F or greater, the maximum RCS cooldown rate is 100°F/hr.

When TC is less than 178°F, the maximum RCS cooldown rate is 50°F/hr.

If Steam Dump and Bypass (SD&B) is available, CONTROL RCS BOPO temperature with a single SD&B Valve. [Step 4]

• [CA] DETERMINE Steam Dump and Bypass is NOT available and GO BOPO TO Step 9.

Examiner Note: HCV-1040, Atmospheric Dump Valve, may already be in service following the Loss of Condenser Vacuum that occurred on Reactor Trip.

BOPO If HCV-1040, is available, CONTROL RCS temperature as follows: [Step 9]

• DEPRESS the valve toggle to SELECT HCV-1040. [Step 9.a]

• PUSH UP and DOWN arrows as required to ADJUST HCV-1040 output as needed. [Step 9.b]

CRITICAL TASK ATCO IMPLEMENT PC-11, Pressure Control, to lower RCS pressure.

Examiner Note: The following steps are from PC-11, Pressure Control. PC-12, RCS Pressure-Temperature Limits (graph), is maintained on a Control Room hardcopy.

CAUTION A charging header flow path must be maintained at all times.

MAINTAIN RCS pressure per the PC-12, RCS Pressure-Temperature Limits ATCO graph: [Step 1]

• DETERMINE Steps 1.a through 1.d N/A. [Step 1.e]

NRC Simulator Scenario 1 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 6, 7, 8, & 9 Page 29 of 29 Event

Description:

Inadvertent Main Turbine Trip / Pressurizer Safety Valve Failure / Loss of Condenser Vacuum /

Pressurizer Pressure Low Signal Actuation Failure / Low Pressure Safety Injection Pumps Start Failure Time Position Applicants Actions or Behavior

• CONTROL Auxiliary Spray flow as necessary by operating the following:

[Step 1.e]

• HCV-240, PZR Auxiliary Spray Isolation Valve

• HCV-249, PZR Auxiliary Spray Isolation Valve

• HCV-238, Loop 1 Charging Isolation Valve

• HCV-239, Loop 2 Charging Isolation Valve

• If HPSI Stop and Throttle criteria is met, CONTROL Pressurizer level using Charging, Letdown, and/or HPSI flow per IC-11, Inventory Control.

[Step 1.f]

• CONTROL RCS heat removal per HR-12, Secondary Heat Removal Operation. [Step 1.g]

When RCS Cooldown and Depressurization is in progress, TERMINATE the scenario.

NRC Simulator Scenario 1 Outline Rev. Final

Appendix D Scenario Outline Form ES-D-1 Facility: Fort Calhoun Station Scenario No.: 3 Op Test No.: Dec 2015 NRC Examiners: Operators:

Initial Conditions: 100% power MOL - RCS Boron is 482 ppm (by sample).

Turnover: Maintain steady-state power conditions. Rotate Heater Drain Pumps FW-5B and FW-5C per OI-VD-1, Feedwater Heater Vents and Drains Normal Operation. Charging Pump CH-1C out of service for packing repair.

Critical Tasks:

• Stop One Reactor Coolant Pump in Each Loop when Reactor Coolant System Pressure is

< 1350 psia, Prior to losing Reactor Coolant Pump Net Positive Suction Head. (Event 7)

• Reduce and Maintain RCS THOT 510°F to Maintain SG Pressure Below Safety Valve Setpoint of 1000 psia Prior to Isolating SG RC-2B. (Event 7)

• Isolate the Affected Steam Generator with a Tube Rupture to Minimize Spread of Contamination Prior to Exiting EOP-04, Steam Generator Tube Rupture. (Event 7)

Event No. Malf. No. Event Type* Event Description 1 N (BOPO) Rotate Heater Drain Pumps per OI-VD-1, Feedwater Heater Vents

+10 min and Drains Normal Operation, Attachment 2.

2 I (ATCO, CRS) Pressurizer Level Channel Transmitter LT-101X Fails Low.

+20 min Transfer Pressurizer Level Control to LT-101Y.

3 I (BOPO, CRS) Steam Generator RC-2A Steam Flow Transmitter FT-907 Fails

+30 min High. Bypass Affected Transmitter.

4 C (ATCO, CRS) Charging Pump CH-1A Trip.

+40 min TS (CRS) Restore Letdown and Charging Flow.

5 C (ATCO,BOPO, Steam Generator RC-2B Tube Leak Greater Than 150 GPD.

+50 min CRS) TS (CRS) Isolate Blowdown Flow.

6 R (ATCO) Commence Plant Shutdown per AOP-05, Emergency Shutdown.

+60 min N (BOPO, CRS) 7 M (ATCO, BOPO, Steam Generator RC-2B Tube Rupture at 500 GPM on 10 Minute

+70 min CRS) Ramp Upon 3% to 5% Load Reduction.

8 I (BOPO) Diesel Generator DG-01 Start Failure on SIAS.

+70 min Manual Start Required.

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Technical Specifications Actual Target Quantitative Attributes 1 Malfunctions after EOP entry (1-2) 4 Abnormal events (2-4) 1 Major transients (1-2) 1 EOPs entered/requiring substantive actions (1-2) 0 EOP contingencies requiring substantive actions (0-2) 3 Critical tasks (2-3)

NRC Simulator Scenario 3 Outline Rev. Final

Scenario Event Description NRC Scenario 3 SCENARIO

SUMMARY

NRC 3 The crew will assume the shift at 100% power per OP-4, Load Change and Normal Power Operation.

The scheduled activity is to rotate Heater Drain Pumps by starting FW-5C and securing FW-5B per OI-VD-1, Feedwater Heater Vents and Drains Normal Operation, Attachment 2, Rotating Operating Heater Drain Pumps.

The next event is a low failure of Pressurizer Level Control Channel, LT-101X. Operator actions are per ARP-CB-1/2/3/A4, Window C PRESSURIZER LEVEL LO-LO CHANNEL X. The crew will transfer to the standby channel LT-101Y and restore Letdown per OI-RC-8, Reactor Coolant System Level Control Normal Operation, Attachment 8, Transferring Pressurizer Level Control Channel in CASCADE and Attachment 4, Transferring Letdown Controller from AUTOMATIC to MANUAL.

When plant conditions are stable, a high failure of Steam Generator RC-2A Steam Flow Transmitter FT-907 will occur. Initial operator actions are per ARP-DCS-FW, Feedwater DCS Annunciator Response Procedure and include verifying Feedwater Control is in Single Element Control, bypassing the failed input, and determining 3 Element Control is restored.

The next event is a trip of the running Charging Pump. Operator actions are per ARP-CB-1/2/3/A2, Window A-6L - CHARGING FLOW LO and include isolating of Letdown and verifying no system leaks exist. Charging Pump CH-1B is placed in service per OI-CH-1, Chemical and Volume Control System Normal Operation, Attachment 1, Startup of Charging and Letdown. The SRO will refer to Technical Specification LCO 2.2.4 - Charging Pumps - Operating.

When Charging flow is restored, a Steam Generator Tube Leak of greater than 150 gallons per day will occur on Steam Generator RC-2B. The crew will enter AOP-22, Reactor Coolant Leak, and implement Attachment B, Primary to Secondary Leak Rate Actions. RM-064, Main Steam Line Radiation Monitor, is placed in service to assist in determining leak size and location. Various Secondary Side valves are closed to minimize system contamination and HR-21, Blowdown Operation is performed to isolate blowdown flow from SG RC-2B. The SRO will refer to Technical Specification LCO 2.1.4 - Reactor Coolant System Leakage Limits.

Once blowdown is isolated, entry into AOP-05, Emergency Shutdown, is performed to bring the plant into MODE 4. When power has been reduced 3% to 5%, a Steam Generator Tube Rupture of 500 gpm will commence on a 10 minute ramp.

The crew enters EOP-00, Standard Post Trip Actions, and then transitions to EOP-04, Steam Generator Tube Rupture. Diesel Generator DG-01 fails to start upon SIAS and must be manually started. While in EOP-04, the Reactor Coolant System is cooled per HR-12, Secondary Heat Removal Operation, and the RCS is depressurized to less than 1000 psia per PC-11, Pressure Control, to allow isolating the affected Steam Generator. When SG RC-2B is isolated, the scenario is terminated.

Risk Significance:

• Failure of risk important system prior to trip: Loss of Charging Pump Steam Generator Tube Leak

• Risk significant core damage sequence: Steam Generator Tube Rupture

• Risk significant operator actions: Stop RCPs Upon Loss of Subcooling Isolate Affected Steam Generator Cooldown and Depressurize RCS NRC Simulator Scenario 3 Outline Rev. Final

Scenario Event Description NRC Scenario 3 BOOTH OPERATOR INSTRUCTIONS for SIMULATOR SETUP RESET to IC-103 and LOAD & EXECUTE NRC 3.sce for NRC Scenario 3.

Preset Item - CH-1C Removed from Service Type Item Value Condition Malfunction BUS_1B3B_4B_5_BKR_Trip True Scenario Event: CH-1C OOS Preset Item - Event 9 - Diesel Generator #1 Auto Start Failure Type Item Value Condition Expert H_PD1_033_3 Reset Scenario Event: DG-1 H_PD1_031_3 Reset Auto Start Failure Event 2 - Pressurizer Level Transmitter LT-101X Fails Low Type Item Value Condition Transmitter RCS_LT101X 0, ramp = 5 seconds When directed by examiner, trigger/activate this event.

Scenario Event: Pzr Level LT-101X Fail Low Event 3 - Steam Generator Flow Transmitter LT-907 Fails High Type Item Value Condition Transmitter FT-907 4000000, ramp = 5 sec When directed by examiner, trigger/activate this event.

FT-907 DCS Fail High Scenario Event: SG Flow FT-907-1 DCS Fail High FT-907 Fail High Event 4 - Charging Pump CH-1A trips Type Item Value Condition Malfunction BUS_1B3A_4_BKR_TRIP True When directed by examiner, trigger/activate this event.

Scenario Event: CH-1A Trip Event 5 - Primary-to-Secondary SG Tube Leak Develops in Steam Generator RC-2B Type Item Value Condition Malfunction RCS04B 0.001 When directed by examiner, trigger/activate this event.

Scenario Event: RC-2B S/G Tube Leak Event 7 - Steam Generator Tube Leak in RC-2B Grows to Tube Rupture Type Item Value Condition Malfunction RCS04B 1.4, ramp = 600 sec When directed by examiner, trigger/activate this event.

Scenario Event: RC-2B S/G Tube Rupture NRC Simulator Scenario 3 Outline Rev. Final

Scenario Event Description NRC Scenario 3 Booth Operator: INITIALIZE to IC-103 and LOAD NRC 3.sce.

ENSURE all Simulator Annunciator Alarms are ACTIVE.

ENSURE all Control Board Tags are removed.

ENSURE Charging Pump CH-1A in service.

ENSURE Charging Pump CH-1C OOS for emulsified oil replacement with Information Tag attached.

ENSURE Channel X Pressurizer Pressure and Level selected.

ENSURE Reactivity Briefing Sheet printout provided with Turnover.

ENSURE Middle-of-Life Thumb Rule Sheet provided with Turnover.

ENSURE ERF Computer System Display set to FWD for BOPO.

ENSURE procedures in progress provided to crew in Briefing Room:

- COPY of OI-VD-1, Feedwater Heater Vents and Drains Normal Operation, Attachment 2, Rotating Operating Heater Drains Pumps, INITIALED through Prerequisites and Procedure Step 2.

Control Room Annunciators in Alarm:

0B AI-30-ESF - CHARGING PUMP CH-1C OFF NORMAL Procedure List Event 1: OP-4, Load Change and Normal Power Operation.

Event 1: OI-VD-1, Feedwater Heater Vents and Drains Normal Operation Event 2: ARP-CB-1/2/3/A4, Window C-8, PRESSURIZER LEVEL LO-LO CHANNEL X Event 3: ARP-DCS-FW, Feedwater DCS Annunciator Response Procedure Event 4: ARP-CB-1/2/3/A2, Window A-6L, CHARGING FLOW LO Event 4: OI-CH-1, Chemical and Volume Control System Normal Operation, Attachment 1, Startup of Charging and Letdown Event 5: AOP-22, Reactor Coolant Leak Event 5: HR-21, Blowdown Operation Event 6: AOP-05, Emergency Shutdown Event 7: EOP-00, Standard Post Trip Actions Event 7: EOP-04, Steam Generator Tube Rupture Event 8: HR-12, Secondary Heat Removal Operation Event 8: PC-11, Pressure Control NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 1 Page 5 of 36 Event

Description:

Rotate Heater Drain Pumps Time Position Applicants Actions or Behavior Booth Operator: When directed, RESPOND to requests from Control Room.

Indications Available:

NONE Examiner Note: The following steps are from OI-VD-1, Feedwater Heater Vents and Drains Normal Operation, Attachment 2.

BOPO PERFORM the following at CB-10, 11: [Step 3]

• PLACE 43/FW Switch in OFF. [Step 3.a]

• VERIFY Annunciator CB-10,11/A10, Window B-6L - 43/FW TRANSFER SWITCH OFF AUTO in ALARM. [Step 3.b]

Examiner Note: XC105 is the Computer (DCS) generated value for Secondary Calorimetric.

CRS DECLARE XC105 invalid. [Step 4]

Make plant announcement, then:

BOPO PLACE FW-5C, Heater Drain Pump control switch to AFTER-START at CB-10, 11. [Step 5]

VERIFY FW-5C, Heater Drain Pump ammeter returns to less than 80 amps BOPO in less than 15 seconds and STABILIZES at ~ 66 amps. [Step 6]

Booth Operator: If contacted, REPORT FCV-1216C is closed.

VERIFY FCV-1216C, Heater Drain Pump FW-5C Recirculation Control Valve BOPO CLOSES. [Step 7]

PLACE FW-5B, Heater Drain Pump control switch to AFTER-STOP at BOPO CB-10, 11. [Step 8]

NOTE Verification of Cooling Water Flow to the Seal cooler will be used to ensure Stuffing Box pressure is < 250 psig when Pressure Gauge PI-1192A, B, or C is out of service.

NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 1 Page 6 of 36 Event

Description:

Rotate Heater Drain Pumps Time Position Applicants Actions or Behavior Booth Operator: If contacted, REPORT FW-5C discharge and stuffing box pressures normal.

BOPO MONITOR the following parameters on Heater Drain Pump FW-5C: [Step 9]

• Motor amperage at ~66 amps.

• PI-1269C, Pump Discharge pressure at ~160 psig on ERF Computer.

• Heater Drain Tank level ~54% on CB-10, 11.

• Bearing temperatures on ERF Display FWD normal.

• PI-1192C, Stuffing Box pressure < 250 psig read locally.

Booth Operator: If contacted, REPORT FW-5B is not rotating in reverse.

CONTACT Auxiliary Operator to VERIFY FW-5B, Heater Drain Pump NOT BOPO ROTATING in reverse direction. [Step 10]

BOPO PERFORM the following at CB-10, 11: [Step 11]

• PLACE 43/FW Switch in AUTO. [Step 11.a]

• VERIFY Annunciator CB-10, 11/A10, Window B-6L - 43/FW TRANSFER SWITCH OFF AUTO is CLEAR. [Step 11.b]

Booth Operator: If contacted, REPORT Shift Technical Advisor will restore GARDEL.

CONTACT Shift Technical Advisor to RESTORE GARDEL data feed per CRS OI-ERFCS-2. [Step 12]

When 12 minute validity period has passed and parameters are steady-state, STA DECLARE XC105 valid and ENTER in Control Room Log. [Step 13]

When restoration of XC105 is discussed, PROCEED to Event 2.

NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 2 Page 7 of 36 Event

Description:

Pressurizer Level Channel Transmitter Failure Time Position Applicants Actions or Behavior Booth Operator: When directed, EXECUTE Event 2.

- Pressurizer Level Channel Transmitter LT-101X fails low.

Indications Available:

CB-1,2,3/A4 - PRESSURIZER LEVEL LO-LO CHANNEL X CB-1,2,3/A4 - PRESSURIZER LEVEL HI-LO CHANNEL X Charging Pump CH-1B starts Letdown flow to minimum (~26 gpm)

+30 sec ATCO RESPOND to Annunciator Response Procedures.

ATCO INFORM CRS of Pressurizer Level Channel LT-101X failure.

Examiner Note: ATCO may Operate to Mitigate per OPD 4-09 and TRANSFER to Channel Y.

REFER to ARP-CB-1,2,3/A4, Window C PRESSURIZER LEVEL LO-LO CRS CHANNEL X.

Examiner Note: During this event, Pressurizer pressure may decrease to less than 2075 psia.

If this occurs, the crew should address Technical Specification LCO 2.10.4.5 for Pressurizer low pressure.

Examiner Note: The following steps are from ARP-CB-1,2,3/A4, Window C-8, PRESSURIZER LEVEL LO-LO CHANNEL X.

ATCO VERIFY Pressurizer Level on LR-101X/LR-101Y. [Step 1]

• If Pressurizer level is NOT low, PERFORM the following: [Step 1.1]

• PLACE HC-101 to Channel Y per OI-RC-8. [Step 1.1.1]

• If desired, PLACE HIC-101-1/101-2, Letdown Throttle Valves Controller to MANUAL per OI-RC-8. [Step 1.1.2]

• PLACE HC-101-1, Pzr Heater Cutout Channel Select Switch, to Channel Y. [Step 1.1.3]

Examiner Note: The following steps are from OI-RC-8, Reactor Coolant System Level Control Normal Operation, Attachment 8, Transferring Pressurizer Level Control Channel (X to Y or Y to X) in CASCADE.

NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 2 Page 8 of 36 Event

Description:

Pressurizer Level Channel Transmitter Failure Time Position Applicants Actions or Behavior ATCO ENSURE both Level Controllers are in (C) CASCADE: [Step 1]

• LC-101X-1, Pressurizer Level Controller

• LC-101Y-1, Pressurizer Level Controller If desired, PLACE Letdown Controller HIC-101-1/101-2, Letdown Throttle ATCO Valves Controller in MANUAL per Attachment 4. [Step 2]

Examiner Note: The following steps are from OI-RC-8, Attachment 4, Transferring Letdown Controller from AUTOMATIC to MANUAL.

ENSURE Letdown Controller HIC-101-1/101-2, Letdown Throttle Valves ATCO Controller in AUTO. [Step 1]

ATCO PLACE HIC-101-1/101-2 Manual/Auto Transfer Switch to BAL. [Step 2]

ADJUST Manual Control Knob on HIC-101-1/101-2 until TOP SCALE ATCO indicates 50% (zero deviation; red pointer aligned with the red dot). [Step 3]

ATCO PLACE HIC-101-1/101-2 Manual/Auto Transfer Switch to MAN. [Step 4]

If necessary, MAKE adjustments to HIC-101-1/101-2 Manual Control Knob to ATCO MAINTAIN desired Pressurizer Level. [Step 5]

Examiner Note: The following steps continue from OI-RC-8, Attachment 8.

CAUTION Transfer from the Selected Controller to the Non-Selected Controller should not be performed until both controller outputs are approximately equal.

VERIFY Controller LR-101Y has INDICATED Pressurizer Level and ATCO PROGRAMMED Pressurizer Level Setpoint MATCHED prior to transfer.

[Step 3]

PLACE HC-101, Pressurizer Level Channel Selector Switch, to Channel Y.

ATCO

[Step 4]

NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 2 Page 9 of 36 Event

Description:

Pressurizer Level Channel Transmitter Failure Time Position Applicants Actions or Behavior ENSURE Controller LC-101Y-1 is controlling INDICATED Pressurizer Level ATCO at PROGRAMMED Setpoint. [Step 5]

PUSH LC-101-1 & LC-101-2, Charging Pump Bistable Reset buttons on ATCO Reactor Regulating System Panel AI-4B and VERIFY all bistables are RESET. [Step 6]

If required, PLACE Letdown Controller HIC-101-1/101-2, Letdown Throttle ATCO Valves Controller, in AUTO per Attachment 3. [Step 7]

Examiner Note: The following steps are from OI-RC-8, Attachment 3, Transferring Letdown Controller from MANUAL to AUTOMATIC.

ENSURE Letdown Controller HIC-101-1/101-2, Letdown Throttle Valves ATCO Controller is in (M) MANUAL. [Step 1]

Manually ADJUST HIC-101-1/101-2, Letdown Throttle Valves Controller and ATCO PIC-210, Letdown Press Controller until following parameters are met:

[Step 2]

• Indicated Pressurizer Level matches the Programmed Pressurizer Level Setpoint on LR-101X or LR-101Y, Pressurizer Level Recorder.

• PIC-210 is maintaining 200 psi to 400 psi.

ADJUST bias knob on HIC-101-1/101-2 until the top scale indicates 50%

ATCO (zero deviation; red pointer aligned with the red dot). [Step 3]

PLACE HIC-101-1/101-2 Manual/Auto Transfer Switch to BAL, then to ATCO AUTO. [Step 4]

If necessary, ADJUST the bias knob of HIC-101-1/101-2 to ENSURE ATCO Indicated Pressurizer Level is maintained at Programmed Pressurizer Level setpoint. [Step 5]

NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 2 Page 10 of 36 Event

Description:

Pressurizer Level Channel Transmitter Failure Time Position Applicants Actions or Behavior Examiner Note: The following steps continue from ARP-CB-1,2,3/A4, Window C-8.

ATCO VERIFY RCS Pressure on PR-103X/PR-103Y > 1600 psia. [Step 2]

ATCO ENSURE all Pressurizer Heaters DEENERGIZED. [Step 3]

DETERMINE RCS Cold Leg temperatures on A-D/TI-112C and A-D/TI-122C ATCO are NOT lowering. [Step 4]

• CHECK VCT level on LI-219, for indication of lowering level. [Step 4.1]

• DETERMINE VCT level is NOT lowering. [Step 4.2]

ATCO VERIFY the following CVCS parameters: [Step 5]

• ENSURE Letdown at minimum flow of 26 gpm on FIC-212. [Step 5.1]

• ENSURE Charging Pumps CH-1A & CH-1B are RUNNING. [Step 5.2]

ATCO NOTIFY Work Week Manager of Pressurizer level instrument failure. [Step 6]

When Pressurizer level is in AUTO, PROCEED to Event 3.

NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 3 Page 11 of 36 Event

Description:

Steam Generator Steam Flow Transmitter Failure Time Position Applicants Actions or Behavior Booth Operator: When directed, EXECUTE Event 3.

- Steam Generator RC-2A Steam Flow Transmitter FT-907 fails high.

Indications Available:

Feedwater Digital Control System Alarm

+30 sec BOPO RESPOND to Annunciator Response Procedures.

INFORM CRS Steam Generator RC-2A Steam Flow Transmitter FT-907 BOPO failed high.

CRS DIRECT actions of ARP-DCS-FW, FT-907.

Examiner Note: The following steps are from ARP-DCS-FW, Feedwater Digital Control System.

BOPO PERFORM the following for Steam Flow Instrument FT-907 failure: [Step 1]

• VERIFY that FORCED TO 1 ELEM and 1 ELEM AUTO is displayed on Feedwater Regulating System display for RC-2A PT-907. [Step 1.1]

• TOUCH display with the BAD process. [Step 1.2]

• DETERMINE BAD input NOT automatically bypassed. [Step 1.3]

• TOUCH Bypass on verification faceplate to BYPASS BAD input.

[Step 1.3.1]

• VERIFY point displays GOOD status. [Step 1.3.2]

• ENSURE control SHIFT to 3 ELEMENT AUTO. [Step 1.3.3]

CRS DETERMINE Steam Generator level instruments NOT affected. [Step 2]

CRS DETERMINE BAD input bypassed MANUALLY. [Step 3]

BOPO MONITOR Steam Generator levels. [Step 4]

CRS VERIFY XC-105, Secondary Calorimetric, is valid. [Step 5]

CRS DETERMINE LT-903 or LT-906 NOT cause of alarm. [Step 6]

NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 3 Page 12 of 36 Event

Description:

Steam Generator Steam Flow Transmitter Failure Time Position Applicants Actions or Behavior BOPO NOTIFY Work Week Manager of FT-907 malfunction. [Step 7]

When Steam Generator levels are normal, PROCEED to Event 4.

NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 4 Page 13 of 36 Event

Description:

Charging Pump Trip Time Position Applicants Actions or Behavior Booth Operator: When directed, EXECUTE Event 4.

- Charging Pump CH-1A trip.

Indications Available:

CB-1,2,3/A2 - CHARGING PUMPS TRIP CB-1,2,3/A2 - CHARGING FLOW LO

+30 sec ATCO RESPOND to Annunciator Response Procedures.

Examiner Note: ATCO may Operate to Mitigate per OPD 4-09 and START CH-1B to avoid losing Letdown flow. Charging Pump CH-1B does not AUTO START until a level deviation exists. If Letdown is lost, steps to restore are included at the end of the event ATCO INFORM CRS of Charging Pump CH-1A trip.

CRS REFER to ARP-CB-1,2,3/A2, Window A-6L - CHARGING FLOW LO.

Examiner Note: The following steps are from ARP-CB-1,2,3/A2, Window A-6L - CHARGING FLOW LO.

ATCO OBSERVE Charging Header flow LOW. [Step 1]

If Charging flow is lost, CLOSE TCV-202 and HCV-204 to ISOLATE ATCO Letdown. [Step 2]

• DETERMINE TCV-202, Letdown to Regenerative Heat Exchanger Isolation Valve AUTO CLOSED or manually CLOSE.

• Manually CLOSE HCV-204, Reactor Coolant to Letdown Heat Exchanger Isolation Valve.

NOTE Based on plant conditions, XC-105 and GARDEL may be invalid.

Booth Operator: When contacted about the status of CH-1A, REPORT a breaker overcurrent trip. Investigation of CH-1A: The pump looks normal locally. If Maintenance or Work Week Manager is contacted, estimated time to restore CH-1C is 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 4 Page 14 of 36 Event

Description:

Charging Pump Trip Time Position Applicants Actions or Behavior If required, ROTATE Charging Pumps per OI-CH-1, CVCS Normal ATCO Operation, Attachment 1, Startup of Charging and Letdown. [Step 5]

EVALUATE Technical Specification LCO 2.2, Chemical and Volume Control CRS System

• LCO 2.2.4 - Charging Pumps - Operating

• CONDITION LCO 2.2.4.(1) - Two Charging Pumps inoperable

• ACTION LCO 2.2.4.(1) - RESTORE to at least two OPERABLE Charging Pumps within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

Examiner Note: The following steps are from OI-CH-1, CVCS Normal Operation, Attachment 1, Startup of Charging and Letdown.

ATCO START CH-1B-1, Packing Cooling Pump. [Step 1]

ATCO DETERMINE boron equalization not required. [Step 2]

ENSURE LCV-218-2, Volume Control Tank Outlet Valve, OPEN and in AUTO.

ATCO

[Step 3]

CAUTION HCV-247, Charg to RC Loop 1A Isolation Valve must remain open to provide an alternate makeup path for charging and ensure CH-202, Ltdn to Regen Ht Exch Isolation Valve will be able to relieve thermal expansion in the Regenerative Heat Exchanger ATCO ENSURE HCV-247, Charging to RC Loop 1A Isolation Valve, OPEN. [Step 4]

ENSURE one of the following combinations of Charging Isolation Valves ATCO OPEN: [Step 5]

• Charging to RC Loop 1A Isolation Valves

• HCV-247, Charging to RC Loop 1A Isolation Valve

• HCV-238, Charging to RC Loop 1A Isolation Valve

• Charging to RC Loop 2A Isolation Valves

• HCV-248, Charging to RC Loop 2A Isolation Valve NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 4 Page 15 of 36 Event

Description:

Charging Pump Trip Time Position Applicants Actions or Behavior

• HCV-239, Charging to RC Loop 2A Isolation Valve ATCO DETERMINE bypass of Ion Exchangers not required. [Step 6]

ATCO PLACE HC-101-3, Limiter Bypass Switch, in BYPASS. [Step 7]

PLACE HIC-101-1/101-2, Letdown Throttle Valve Controller, in MANUAL.

ATCO

[Step 8]

ATCO CLOSE LCV-101-1, Letdown Heat Exchanger Throttle Valve. [Step 9]

DETERMINE HC-101-2, Letdown Heat Exchanger Valves Selector Switch ATCO positioned as required. [Step 10]

ATCO PLACE PIC-210, Letdown Pressure Controller, in MANUAL. [Step 11]

ATCO THROTTLE PIC-210 to approximately 10% open. [Step 12]

PLACE Charging Pumps Mode Select that to CH-1C - CH-1A position.

ATCO

[Step 13]

CAUTION Pressurizer Level deviation will start standby charging pumps not in PULLOUT.

ATCO START CH-1B, Charging Pump. [Step 14]

ATCO OPEN HCV-204, RC to Letdown Heat Exchanger Isolation Valve. [Step 15]

OPEN TCV-202, Letdown to Regen Heat Exchanger Isolation Valve.

ATCO

[Step 16]

Using HIC-101-1/101-2 INITIATE Letdown flow while adjusting PCV-210 to ATCO maintain Letdown pressure approximately 300 psig. [Step 17]

NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 4 Page 16 of 36 Event

Description:

Charging Pump Trip Time Position Applicants Actions or Behavior BALANCE Charging and Letdown flows to maintain Pressurizer level.

ATCO

[Step 18]

ATCO PLACE HC-101-3 in NORMAL. [Step 19]

ATCO PERFORM the following to place PIC-210 in AUTO: [Step 20]

• PLACE HIC-101-1/101-2 Manual/Auto Transfer Switch to BAL, then to AUTO. [Step 20.a]

• PLACE PIC-210 in AUTO. [Step 20.b]

When Charging and Letdown flows are restored, PROCEED to Event 5.

NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 5 Page 17 of 36 Event

Description:

Steam Generator Tube Leak Time Position Applicants Actions or Behavior Booth Operator: When directed, EXECUTE Event 5.

- Steam Generator RC-2B Tube Leak greater than 150 gpd.

Indications Available:

RM-057, Condenser Off Gas Radiation Monitor in alarm and trending up RM-054B, Steam Generator RC-2B Blowdown Radiation Monitor in alarm and trending up

+30 sec ATCO RESPOND to Radiation Monitor Alarms.

ATCO INFORM CRS of indications of the tube leak on Steam Generator RC-2B.

CRS REFER to AOP-22, Reactor Coolant Leak.

Examiner Note: The following steps are from AOP-22, Reactor Coolant Leak,Section I, Leak Rate Determination and Leak Isolation.

CRS DETERMINE Shutdown Cooling is NOT in operation. [Step 4.1]

Booth Operator: When contacted as Shift Chemist, WAIT 2 minutes and REPORT Steam Generator RC-2B has increased activity and RC-2A has normal activity.

DETERMINE CIAS is NOT present and DIRECT Shift Chemist to PERFORM CRS the following: [Step 4.2]

• PERFORM rapid activity analysis of both Steam Generators. [Step 4.2.a]

• SAMPLE both SGs per CH-SMP-SE-0015, Steam Generator Sampling -

Room 60. [Step 4.2.b]

CRS IMPLEMENT the Emergency Plan. [Step 4.3]

CREW MONITOR the Floating Steps. [Step 4.4]

ATCO DETERMINE Pressurizer level is NOT below programmed level. [Step 4.5]

NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 5 Page 18 of 36 Event

Description:

Steam Generator Tube Leak Time Position Applicants Actions or Behavior ATCO DETERMINE RCS leakage rate per IC-17, RCS Manual Leak Rate and/or Calculation. [Step 4.6]

BOPO CRS DETERMINE RCS leak rate is NOT greater than 40 gpm. [Step 4.7]

Booth Operator: When contacted as Shift Chemist, WAIT 10 minutes, then REPORT initial Steam Generator RC-2B leak rate is greater than 150 gallon per day.

DIRECT Shift Chemist to verify primary to secondary leak rate < 1 gpd per CRS CH-AD-0007, Primary to Secondary Leak Rate Determination. [Step 4.8]

• [CA] If primary to secondary leak rate is > 1 gpd, IMPLEMENT Attachment B, Primary to Secondary Leak Rate Actions.

Examiner Note: The following steps are from AOP-22, Reactor Coolant Leak, Attachment B, Primary to Secondary Leak Rate Actions.

CRS IMPLEMENT SO-G-105, Steam Generator Tube Leakage. [Step 1]

Booth Operator: When contacted, REPORT Work Week Manager will implement SO-G-105.

Continuously MONITOR count rate trends for radiation monitors RM-054A, ATCO RM-054B and RM-057 on ERF Computer System. [Step 2]

PERFORM the following to PLACE RM-064, Main Steam Line Radiation ATCO Monitor, in service at AI-33C: [Step 3]

• PLACE Main Steam Line A/B Enable Switch for HCV-921 and HCV-922 in ON. [Step 3.a]

• PLACE Main Steam Line A/B Mode Selector Switch in AUTO. [Step 3.b]

CRS PERFORM the following to IDENTIFY SG with tube leak: [Step 4]

CRS

• DIRECT Shift Chemist to continue sampling. [Step 4.a]

CRS

• REVIEW Steam Generator analysis for activity. [Step 4.b]

• MONITOR RM-057 & RM-064, Steam Line Radiation Monitors and ATCO DETERMINE both radiation levels RISING. [Step 4.c]

NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 5 Page 19 of 36 Event

Description:

Steam Generator Tube Leak Time Position Applicants Actions or Behavior

• MONITOR RM-054A & RM-054B, SG Blowdown Radiation Monitors and ATCO DETERMINE RM-054B is RISING [Step 4.d]

BOPO

• MONITOR SG levels and DETERMINE no apparent change. [Step 4.e]

Booth Operator: When contacted, EXECUTE remote functions to position HC-2509 / HC-2508 /

FW-268 / FW-266 as required.

Direct Equipment Operators to PERFORM the following to MINIMIZE spread CREW of contamination: [Step 5]

• ENSURE HC-2509, SAMPLE DRAIN TO DRAIN HEADER, is OPEN at AI-107 in Room 60. [Step 5.a]

• ENSURE HC-2508, SAMPLE DRAIN TO CONDENSER C.W. TUNNEL, is CLOSED at AI-107 in Room 60. [Step 5.b]

• ENSURE FW-268, CONDENSATE DUMP VALVE LCV-1193 OUTLET ISOLATION VALVE, is CLOSED at Turbine Building Mezzanine.

[Step 5.c]

• ENSURE FW-266, CONDENSATE DUMP VALVE LCV-1193 BYPASS VALVE, is CLOSED at Turbine Building Mezzanine. [Step 5.d]

CRS

• DETERMINE SG RC-2A is least affected Steam Generator. [Step 5.e]

• DETERMINE SG RC-2B is most affected Steam Generator and BOPO PERFORM the following: [Step 5.f]

• PLACE YCV-1045B, RC-2B to FW-10 Isolation Valve in OVERRIDE.

• PLACE YCV-1045B, RC-2B to FW-10 Isolation Valve to CLOSE.

• CONSIDER stopping Turbine Building Sump Pumps VD-1A & VD-1B.

CRS

[Step 5.g]

CRS

• CONSIDER isolating steam generator blowdown. [Step 5.h]

BOPO

• PLACE RCV-978, 6th Stage Extraction Isolation Valve to STOP. [Step 5.i]

Booth Operator: When contacted, EXECUTE remote function to align Condenser Evacuation Discharge to Auxiliary Building Stack.

• CONTACT Auxiliary Operator to ALIGN Condenser Evacuation CRS Discharge to Auxiliary Building stack per OI-CE-1, Condenser Evacuation System Normal Operation. [Step 5.j]

• DIRECT Radiation Protection to develop a method for processing CRS contaminated Condensate. [Step 5.k]

NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 5 Page 20 of 36 Event

Description:

Steam Generator Tube Leak Time Position Applicants Actions or Behavior CRS DETERMINE primary to secondary leakage greater than 5 gpd. [Step 6]

CRS DETERMINE primary to secondary leakage greater than 30 gpd. [Step 7]

DETERMINE primary to secondary leakage greater than 30 gpd independent CRS of Xe-133 concentration. [Step 8]

DETERMINE primary to secondary leakage greater than 75 gpd independent CRS of Xe-133 concentration. [Step 9]

DETERMINE primary to secondary leakage greater than 75 gpd with a rate CRS increase greater than 30 gpd in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period. [Step 10]

DETERMINE primary to secondary leakage greater than 75 gpd with a rate CRS increase greater than 30 gpd in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. [Step 11]

DETERMINE primary to secondary leak rate greater than 150 gpd (0.10 gpm)

CRS and PERFORM the following: [Step 12]

• ISOLATE blowdown from SG RC-2B per HR-21, Blowdown Operation.

[Step 12.a]

• COMMENCE a Plant Shutdown to MODE 4 per AOP-05, Emergency Shutdown. [Step 12.b]

CRS EVALUATE Technical Specification LCO 2.1, Reactor Coolant System.

• LCO 2.1.4 - Reactor Coolant System Leakage Limits

• CONDITION LCO 2.2.4.(1).(d) - RCS operational LEAKAGE shall be limited to 150 gallons per day primary to secondary LEAKAGE through anyone Steam Generator.

• ACTION LCO 2.1.4.(3) - Primary to secondary LEAKAGE is not within limits, then be in MODE 3, Hot Shutdown, within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> AND be in MODE 4, Cold Shutdown within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

When Technical Specifications have been addressed, PROCEED to Event 6.

NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 6 Page 21 of 36 Event

Description:

Commence Plant Shutdown Time Position Applicants Actions or Behavior Booth Operator: When directed, RESPOND to requests from Control Room.

Indications Available:

NONE Examiner Note: The following steps are from AOP-05, Emergency Shutdown.

NOTE TDB-III-23a and the Power Ascension/Power Reduction Strategy (PAPRs) provide guidance for the shutdown.

CRS CONTACT Reactor Engineer if additional guidance is required. [Step 4.1]

NOTE Operation of more than one Charging Pump will raise the rate of the power reduction.

Examiner Note: Unless directed, boration will occur from the Safety Injection Refueling Water Tank (SIRWT) when in AOP-05 to avoid time constraints.

If borating from SIRWT, COMMENCE boration by performing the following:

CRS

[Step 4.2]

• DETERMINE Charging Pump, CH-1B RUNNING.

ATCO

[Step 4.2.a]

• OPEN LCV-218-3, Charging Pump Suction SIRWT Isolation Valve.

ATCO

[Step 4.2.b]

ATCO

• CLOSE LCV-218-2, VCT Outlet Valve. [Step 4.2.c]

CRS DETERMINE Boration alignment from CVCS NOT required. [Step 4.3]

CRS NOTIFY Energy Marketing of power reduction. [Step 4.4]

NOTE During the power reduction, maintain TC PER TDB Figure III.1, Tave Program.

MAINTAIN RCS Temperature Control via Turbine Load per HR-12, BOPO Secondary Heat Removal Operation: [Step 4.5]

NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 6 Page 22 of 36 Event

Description:

Commence Plant Shutdown Time Position Applicants Actions or Behavior

• MAINTAIN TCOLD 527°F to 547°F AND

• MAINTAIN TCOLD +0°F to -1°F of program.

MAINTAIN Pressurizer Level via Charging and Letdown per IC-11, Inventory ATCO Control: [Step 4.6]

• MAINTAIN Pressurizer Level 45% to 60% AND

• MAINTAIN Pressurizer Level within 4% of program.

PERFORM the following to MAINTAIN VCT level between 55% and 85%:

ATCO

[Step 4.7]

• As required, PLACE LCV-218-1, VCT Inlet Valve to RWTS. [Step 4.7.a]

• When diversion is complete, PLACE LCV-218-1, VCT Inlet Valve to AUTO. [Step 4.7.b]

PERFORM the following to MAXIMIZE Pressurizer Heaters and Spray:

ATCO

[Step 4.8]

• As required, PLACE Backup Heater Control Switches to ON. [Step 4.8.a]

• ADJUST PC-103X or PC-103Y, Pressurizer Pressure Controller Setpoint Pushbutton to maintain pressure between 2080 psia and 2145 psia.[Step 4.8.b]

CAUTION Do not insert CEAs below power dependent insertion limit.

As required, ADJUST Regulating Group 4 to CONTROL ASI per OI-RR-1, ATCO Attachment 4, Axial Shape Index (ASI) Control. [Step 4.9]

Examiner Note: The following steps are from HR-12, Secondary Heat Removal Operation.

BOPO ENSURE Turbine Control is in MANUAL. [Step 1]

NOTE Output will be highlighted by a yellow box when selected.

NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 6 Page 23 of 36 Event

Description:

Commence Plant Shutdown Time Position Applicants Actions or Behavior BOPO PUSH the OUT button to select OUTPUT. [Step 2]

NOTES

1. Depressing the single arrow will adjust turbine load by 0.1%. Depressing the double arrow will adjust turbine load by 0.5%.

2. Tc should be maintained within (+)0°F, (-)1°F of program per TDB-III.1, Tave Program.

PRESS single or double UP[] or DOWN[] arrow to maintain Turbine BOPO Load: [Step 3]

• MAINTAIN TCOLD 527°F to 547°F.

• MAINTAIN TCOLD +0°F to -1°F of program.

Examiner Note: Do not proceed to the next event during electrical plant realignment to 161KV.

When Reactor power is reduced 3% to 5%, PROCEED to Events 7 and 8.

NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 7&8 Page 24 of 36 Event

Description:

Steam Generator Tube Rupture / Diesel Generator Start Failure Time Position Applicants Actions or Behavior Booth Operator: When directed, EXECUTE Events 7 and 8.

- Steam Generator RC-2B Tube Rupture @ 500 gpm on 10 minute ramp.

- Diesel Generator DG-01 start failure on SIAS.

Indications Available:

Pressurizer pressure and level lowering.

RECOGNIZE Pressurizer pressure and level lowering, upward trending

+2 min ATCO Radiation Monitors and MANUALLY TRIP Reactor.

CRS DIRECT performance of EOP-00, Standard Post Trip Actions.

Examiner Note: The following steps are from EOP-00, Standard Post Trip Actions.

ATCO VERIFY Reactivity Control: [Step 1]

• VERIFY ALL of the following: [Step 1.a]

• VERIFY no more than one Regulating or Shutdown CEA NOT inserted.

• VERIFY Reactor Power is LOWERING.

• VERIFY Startup Rate is NEGATIVE.

• MONITOR plant for an uncontrolled RCS Cooldown. [Step 1.b]

CRS DETERMINE Reactivity Control criteria SATISFIED.

BOPO VERIFY Turbine Trip: [Step 2]

• VERIFY HP & LP Stop and Intercept Valves CLOSED.

BOPO ENSURE all Generator Breakers are tripped: [Step 3]

• DETERMINE Generator Output Breaker 3451-4 tripped.

• DETERMINE Generator Output Breaker 3451-5 tripped.

• DETERMINE Generator Field Breaker 41E/G1F tripped.

BOPO VERIFY 4160 V Safeguards Buses 1A3 and 1A4 are ENERGIZED. [Step 4]

NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 7&8 Page 25 of 36 Event

Description:

Steam Generator Tube Rupture / Diesel Generator Start Failure Time Position Applicants Actions or Behavior CRS DETERMINE Maintenance of Vital Auxiliaries criteria SATISFIED.

Examiner Note: The following step (Verify Diesel Generators running) is not required until Reactor Coolant System Pressure is less than 1600 psia and PPLS has actuated.

VERIFY both Diesel Generators RUNNING on Safety Injection Actuation BOPO Signal. [Step 5]

• [CA] DEPRESS DG-01 Emergency Start pushbutton and VERIFY DG-01 running at 900 RPM.

VERIFY 4160 V Non-Safeguards Buses 1A1 and 1A2 are ENERGIZED.

BOPO

[Step 6]

BOPO VERIFY 125 VDC Buses 1 and 2 are ENERGIZED. [Step 7]

BOPO VERIFY Instrument Air is AVAILABLE: [Step 8]

• DETERMINE Instrument Air pressure 90 psig.

• DETERMINE Instrument Air Compressor CA-1C RUNNING.

ATCO VERIFY Component Cooling Water System operation NORMAL: [Step 9]

• DETERMINE at least one CCW pump RUNNING. [Step 9.a]

• DETERMINE CCW Pump discharge pressure 60 psig. [Step 9.b]

• DETERMINE HCV-438A/B/C/D, CCW to RCP Coolers OPEN. [Step 9.c]

• DETERMINE at least one Raw Water Pump RUNNING. [Step 9.d]

CRS VERIFY RCS Inventory Control criteria satisfied: [Step 10]

• DETERMINE PZR level NOT between 30% and 70% and NOT TRENDING to between 45% and 60%.

• [CA] RESTORE Inventory Control by manually controlling Charging and Letdown. [Step 10.1.a]

• DETERMINE RCS subcooling > 20°F.

NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 7&8 Page 26 of 36 Event

Description:

Steam Generator Tube Rupture / Diesel Generator Start Failure Time Position Applicants Actions or Behavior CRS DETERMINE RCS Inventory Control criteria NOT SATISFIED.

CRS VERIFY RCS Pressure Control criteria satisfied: [Step 11]

ATCO

• DETERMINE RCS pressure less than 1600 psia.

• [CA] VERIFY RCS pressure < 2300 psia and PORV NOT open.

[Step 11.1]

• [CA] When RCS pressure < 1350 psia, PERFORM the following:

[Step 11.2]

ATCO * [CA] STOP one RCP in each Loop.

• [CA] DETERMINE RCS pressure < 1600 psia and VERIFY Engineered Safeguards ACTUATED. [Step 11.3]

• [CA] DETERMINE PPLS relays 86A/PPLS / 86B/PPLS /

86A1/PPLS / 86B1/PPLS have TRIPPED.

[Step 11.3.a]

• [CA] DETERMINE all PPLS relays have TRIPPED.

[Step 11.3.b]

• [CA] DETERMINE VIAS relays 86A/VIAS / 86B1/VIAS /

86B/VIAS / 86A1/VIAS have TRIPPED. [Step 11.3.c]

• [CA] DETERMINE SIAS relays 86A/SIAS / 86AX/SIAS /

86B1/SIAS / 86B1X/SIAS / 86B/SIAS / 86BX/SIAS /

86A1/SIAS / 86A1X/SIAS have TRIPPED.

[Step 11.3.d]

• [CA] DETERMINE CIAS relays 86A/CIAS / 86B1/CIAS /

86B/CIAS / 86A1/CIAS have TRIPPED. [Step 11.e]

• [CA] ENSURE required pumps RUNNING [Step 11.3.f]

• DETERMINE HPSI Pumps SI-2A & SI-2B RUNNING.

• DETERMINE LPSI Pumps SI-1A and SI-1B RUNNING.

• DETERMINE Charging Pumps CH-1B RUNNING.

• [CA] ENSURE acceptable SI flow per Attachment IC-13, SI Flow vs. Pressurizer Pressure. [Step 11.3.g]

• [CA] ENSURE Emergency Boration in progress.

ATCO

[Step 11.3.h]

CRS DETERMINE RCS Pressure Control criteria NOT SATISFIED.

NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 7&8 Page 27 of 36 Event

Description:

Steam Generator Tube Rupture / Diesel Generator Start Failure Time Position Applicants Actions or Behavior Examiner Note: The following steps are from RC-11, Emergency Boration Verification.

ATCO ENSURE the following valves are CLOSED: [Step 1]

• FCV-269X, Demin Water Makeup Valve

• FCV-269Y, Boric Acid Makeup Valve

• HCV-264, CH-4A Recirc Valve

• HCV-257, CH-4B Recirc Valve ATCO VERIFY all the following valves OPEN: [Step 2]

• HCV-268, Boric Acid Pump Header to Charging Pumps Isolation Valve

• HCV-265, CH-11A Gravity Feed Valve

• HCV-258, CH-11B Gravity Feed Valve ATCO ENSURE all available Boric Acid Pumps RUNNING: [Step 3]

• CH-4A, Boric Acid Pump

• CH-4B, Boric Acid Pump ATCO ENSURE all available Charging Pumps RUNNING: [Step 4]

• CH-1A, Charging Pump is tripped.

• CH-1B, Charging Pump is RUNNING.

• CH-1C, Charging Pump is OOS.

ATCO ENSURE the following valves are CLOSED: [Step 5]

• LCV-218-2, VCT Outlet Valve

• LCV-218-3, Charging Pump Suction SIRWT Isolation Valve

• HCV-257, CH-4B Recirculation Valve

• HCV-264, CH-4A Recirculation Valve ATCO DETERMINE Emergency Boration is in progress. [Step 6]

NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 7&8 Page 28 of 36 Event

Description:

Steam Generator Tube Rupture / Diesel Generator Start Failure Time Position Applicants Actions or Behavior Examiner Note: The following steps continue from EOP-00, Standard Post Trip Actions.

CRS VERIFY Core Heat Removal criteria satisfied: [Step 12]

ATCO

• DETERMINE RCP NPSH requirements met.

• DETERMINE at least one RCP operating.

• DETERMINE Core T 10°F.

Stop One Reactor Coolant Pump in Each Loop when Reactor Coolant System CRITICAL TASK Pressure is < 1350 psia, Prior to losing Reactor Coolant Pump Net Positive STATEMENT Suction Head.

CRITICAL DETERMINE Reactor Coolant System pressure < 1350 psia and PERFORM TASK ATCO the following:

ATCO

• STOP one RCP in each Loop.

CRS DETERMINE Core Heat Removal criteria SATISFIED.

NOTE If Instrument Air to valves HCV-1105, HCV-1106, HCV-1107A/B and HCV-1108A/B is not available, throttling of these valves is not possible. Open or close operation of these valves is possible for a minimum of three cycles.

CRS VERIFY RCS Heat Removal criteria satisfied:

VERIFY Main Feedwater is restoring SG levels to 35% to 80% NR and 73%

BOPO to 94% WR. [Step 13]

• DETERMINE FCV-1101 & FCV-1102 Feed Regulating Valves CLOSED.

[Step 13.a]

• DETERMINE HCV-1105 & HCV-1106 Feed Regulating Bypass Valves ramped to between 40% & 45% OPEN. [Step 13.b]

BOPO

• PLACE 43/FW switch in OFF. [Step 13.c]

• ENSURE no more than one Main Feedwater Pump RUNNING.

[Step 13.d]

• ENSURE no more than one Condensate Pump RUNNING. [Step 13.e]

NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 7&8 Page 29 of 36 Event

Description:

Steam Generator Tube Rupture / Diesel Generator Start Failure Time Position Applicants Actions or Behavior

• STOP running Heater Drain Pumps FW-5A, FW-5B, and/or FW-5C.

BOPO

[Step 13.f]

• ENSURE SG Blowdown Isolation Valves CLOSED. [Step 13.g]

• HCV-1387A & HCV-1387B

• HCV-1388A & HCV-1388B VERIFY Steam Dump and Bypass Valves controlling both of the following:

BOPO

[Step 14]

• DETERMINE RCS TCOLD between 525°F and 535°F.

• DETERMINE Steam Generator pressure between 850 psia & 925 psia.

CRS DETERMINE RCS Heat Removal criteria SATISFIED.

CRS VERIFY Containment Isolation criteria satisfied:

ATCO VERIFY Normal Containment conditions exist: [Step 15]

• DETERMINE no unexpected rise in Containment Sump level. [Step 15.a]

• DETERMINE Containment Area Radiation Monitors NOT in alarm.

[Step 15.b]

• DETERMINE Containment Ventilation Radiation Monitors NOT in alarm.

[Step 15.c]

• DETERMINE RM-054B, SG Blowdown Radiation Monitor ALARMING.

ATCO

[Step 15.d]

• [CA] MINIMIZE spread of contamination: [Step 15.d.1]

• [CA] VERIFY RCV-978, 6th Stage Extraction Isolation Valve BOPO CLOSED. [Step 15.d.1.1)]

• [CA] VERIFY all Blowdown Isolation Valves CLOSED.

[Step 15.d.1.2)]

• [CA] HCV-1387A & HCV-1387B

• [CA] HCV-1388A & HCV-1388B

• DETERMINE RM-054B, SG Blowdown Radiation Monitor and RM-057, ATCO Condenser Off Gas Radiation Monitor TRENDING upward. [Step 15.e]

• [CA] DETERMINE if Steam Generator Tube Rupture is in progress:

CRS

[Step 15.e.1]

NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 7&8 Page 30 of 36 Event

Description:

Steam Generator Tube Rupture / Diesel Generator Start Failure Time Position Applicants Actions or Behavior

• [CA] DIRECT Shift Chemist to perform rapid activity analysis of both SGs. [Step 15.e.1.1)]

• [CA] DETERMINE SG RC-2B has an abnormal rise in level.

BOPO

[Step 15.e.1.2)]

ATCO

• VERIFY Containment conditions: [Step 15.f]

• DETERMINE Containment pressure < 3 psig.

• DETERMINE Containment temperature < 120°F.

CRS DETERMINE Containment Integrity criteria NOT SATISFIED.

CRS DIAGNOSE event in progress: [Step 16]

• DETERMINE Reactivity Control requirements met.

• DETERMINE both DC buses energized.

• DETERMINE at least one Vital 4160 V Bus energized.

• DETERMINE at least one Non-Vital 4160 V Bus energized.

• DETERMINE at least one RCP running.

• DETERMINE adequate Feedwater flow to at least one SG.

• VERIFY Pressurizer pressure > 1800 psia with high subcooled margin, normal SG pressure, and no indications of primary to secondary leakage.

• If not, CONSIDER EOP-04, Steam Generator Tube Rupture.

NOTE Certain events (i.e., LOCA, SGTR, UHE and Loss of All Feedwater) do not require offsite power in order to adequately, mitigate the effects of the accident. For this reason, the LOCA, SGTR, UHE or Loss of All Feedwater procedure may be implemented even if a Loss of Offsite Power has also occurred.

• DETERMINE all Safety Function Acceptance Criteria NOT SATISFIED.

• DETERMINE single event in progress and TRANSITION to EOP-04, Steam Generator Tube Rupture.

Examiner Note: The following steps are from EOP-04, Steam Generator Tube Rupture.

CRS CONFIRM Standard Post Trip Actions have been performed. [Step 1]

NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 7&8 Page 31 of 36 Event

Description:

Steam Generator Tube Rupture / Diesel Generator Start Failure Time Position Applicants Actions or Behavior CRS CONFIRM Steam Generator Tube Rupture Diagnosis: [Step 2]

• VERIFY Safety Function Status Check Acceptance Criteria being satisfied. [Step 2.a]

• VERIFY CIAS is present and SAMPLE both SGs. [Step 2.c]

CRS IMPLEMENT the Emergency Plan. [Step 3]

• Time: __________

CREW MONITOR the Floating Steps. [Step 4]

DETERMINE RCS pressure 1600 psia and VERIFY Engineered CRS Safeguards are ACTUATED: [Step 5]

• DETERMINE PPLS relays 86A/PPLS / 86B/PPLS / 86A1/PPLS /

86B1/PPLS have TRIPPED. [Step 5.a]

• DETERMINE SIAS relays 86A/SIAS / 86AX/SIAS / 86B1/SIAS /

86B1X/SIAS / 86B/SIAS / 86BX/SIAS / 86A1/SIAS / 86A1X/SIAS have TRIPPED. [Step 5.b]

• DETERMINE CIAS relays 86A/CIAS / 86B1/CIAS / 86B/CIAS /

86A1/CIAS relays TRIPPED. [Step 5.c]

• DETERMINE VIAS relays 86A/VIAS / 86B1/VIAS / 86B/VIAS /

86A1/VIAS relays TRIPPED. [Step 5.d]

OPTIMIZE Safety Injection and Charging flow and PERFORM the following:

ATCO

[Step 6]

• ENSURE required Safety Injection Pumps RUNNING: [Step 6.a]

• DETERMINE HPSI Pumps SI-2A & SI-2B RUNNING.

• DETERMINE LPSI Pumps SI-1A and SI-1B RUNNING.

• DETERMINE Charging Pumps CH-1B RUNNING.

• DETERMINE Emergency Boration already in progress per RC-11, ATCO Emergency Boration Verification. [Step 6.b]

• ENSURE adequate SI flow per IC-13, Safety Injection Flow vs.

Pressurizer Pressure. [Step 6.c]

NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 7&8 Page 32 of 36 Event

Description:

Steam Generator Tube Rupture / Diesel Generator Start Failure Time Position Applicants Actions or Behavior NOTE Main PZR Spray flow will be reduced with less than four-pump operation. Pressure should be controlled using Main and Auxiliary PZR Spray whenever the Plant is placed in a two-pump configuration.

ATCO VERIFY RCP operating parameters: [Step 7]

• ENSURE at least one RCP stopped if TCOLD < 500°F. [Step 7.a]

• DETERMINE one RCP stopped in each loop when RCS pressure 1350 psia following SIAS. [Step 7.b]

• DETERMINE all RCPs STOPPED on low subcooling. [Step 7.c]

• Time: _______

DETERMINE Condenser vacuum greater than 10.92 inches Hg absolute or CRS 19 inches Hg. [Step 8]

NOTE Reducing RCS TH to less than or equal to 510°F will maintain adequate RCP NPSH and RCS subcooling when RCS pressure is reduced below SG safety valve setpoint of 1000 psia.

CAUTION When TC is 178°F or greater, the maximum RCS cooldown rate is 100°F/hr.

When TC is less than 178°F, the maximum RCS cooldown rate is 50°F/hr.

COMMENCE a cooldown using both SGs to reduce RCS THOT to 510°F per BOPO Attachment HR-12, Secondary Heat Removal Operation. [Step 9]

COMMENCE a depressurization of RCS to less than 1000 psia per ATCO Attachment PC-11, Pressure Control. [Step 10]

Examiner Note: The following steps are from HR-12, Secondary Heat Removal Operation.

BOPO ENSURE Turbine Control is in MANUAL. [Step 1]

BOPO * [CA] DETERMINE Turbine NOT online and GO TO Step 4.

NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 7&8 Page 33 of 36 Event

Description:

Steam Generator Tube Rupture / Diesel Generator Start Failure Time Position Applicants Actions or Behavior NOTES

1. In MANUAL, single arrows are 1% / double arrows are 5% change in OUTPUT value.

2. While Steam Dump and Bypass Control is in Temperature/Pressure Mode, the controllers PC0910 and TC0909_PI will alternate between controls, depending on the higher output signal. A red square outlining the controlling function signify which parameter is in control.

3. Steps 4 through 11 may be performed as needed, and in any order.

CAUTIONS When TC is 178°F or greater, the maximum RCS cooldown rate is 100°F/hr.

When TC is less than 178°F, the maximum RCS cooldown rate is 50°F/hr.

CRITICAL TASK Reduce and Maintain RCS THOT 510°F to Maintain SG Pressure Below Safety STATEMENT Valve Setpoint of 1000 psia Prior to Isolating SG RC-2B.

CRITICAL DETERMINE Steam Dump and Bypass (SD&B) available and CONTROL TASK BOPO RCS temperature with a single SD&B Valve. [Step 4]

• DEPRESS Valve Toggle to SELECT valve to be operated: [Step 4.a]

• PCV-910 / TCV-909-1 / TCV-909-2 / TCV-909-3 / TCV-909-4

• PLACE Controller for selected valve in MANUAL. [Step 4.b]

• PUSH UP and DOWN arrows to ADJUST Controller Output. [Step 4.c]

• When no longer required, PLACE Controller for selected valve in AUTO.

[Step 4.d]

Examiner Note: The following steps are from PC-11, Pressure Control. PC-12, RCS Pressure-Temperature Limits (graph), is maintained on a Control Room hardcopy.

CAUTION A charging header flow path must be maintained at all times.

MAINTAIN RCS pressure per the PC-12, RCS Pressure-Temperature Limits ATCO graph: [Step 1]

• DETERMINE Steps N/A due to RCS pressure. [Step 1.a to 1.d]

NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 7&8 Page 34 of 36 Event

Description:

Steam Generator Tube Rupture / Diesel Generator Start Failure Time Position Applicants Actions or Behavior CRITICAL

• OPERATE the following to CONTROL Auxiliary Spray flow and TASK ATCO REDUCE RCS pressure to < 1000 psia: [Step 1.e]

• HCV-240, PZR Auxiliary Spray Isolation Valve

• HCV-249, PZR Auxiliary Spray Isolation Valve

• HCV-238, Loop 1 Charging Isolation Valve

• HCV-239, Loop 2 Charging Isolation Valve

• If HPSI Stop and Throttle criteria is met, CONTROL Pressurizer level using Charging, Letdown, and/or HPSI flow per IC-11, Inventory Control.

[Step 1.f]

• CONTROL RCS heat removal per HR-12, Secondary Heat Removal Operation. [Step 1.g]

MAINTAIN RCS Pressure per PC-12, RCS Pressure-Temperature Limits by ATCO performing ANY of the following: [Step 11]

• CONTROL RCS Heat Removal per HR-12, Secondary Heat Removal Operation. [Step 11.a]

• CONTROL Pressurizer Heaters and Spray per PC-11 Pressure Control.

[Step 11.b]

• If HPSI Stop and Throttle criteria are met, CONTROL Charging, Letdown, and/or HPSI flow per IC-11, Inventory Control. [Step 11.c]

If feeding through Feed Ring, MAINTAIN SG levels 44% to 85% NR (77% to BOPO 94% WR) using Main Feedwater or FW-54. [Step 12]

• FEED SGs using HR-15, Main Feed Pump Operation or HR-16, FW-54 Operation. [Step 12.a]

• CONTROL feed flow per HR-11, Manual Feed Control. [Step 12.b]

PERFORM the following to PLACE RM-064, Main Steam Line Radiation ATCO Monitor in service at AI-33C. [Step 13]

• PLACE Main Steam Line A/B Enable Switch for HCV-921 and HCV-922 in ON. [Step 13.a]

• PLACE Main Steam Line A/B Mode Selector Switch in AUTO. [Step 13.b]

CRS DETERMINE Steam Generator RC-2B has the tube rupture. [Step 14]

BOPO PERFORM the following to MINIMIZE spread of contamination: [Step 15]

NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 7&8 Page 35 of 36 Event

Description:

Steam Generator Tube Rupture / Diesel Generator Start Failure Time Position Applicants Actions or Behavior

• CONTACT Auxiliary Operator to POSITION following valves: [Step 15.a]

• ENSURE HC-2509, SAMPLE DRAIN TO DRAIN HEADER is OPEN at AI-107, Room 60. [Step 15.a]

• ENSURE HC-2508, SAMPLE DRAIN TO CONDENSER C.W.

TUNNEL is CLOSED at AI-107, Room 60. [Step 15.b]

• ENSURE FW-268, CONDENSATE DUMP VALVE LCV-1193 OUTLET ISOLATION VALVE, is CLOSED at Turbine Building Mezzanine. [Step 15.c]

• ENSURE FW-266, CONDENSATE DUMP VALVE LCV-1193 BYPASS VALVE, is CLOSED at Turbine Building Mezzanine.

[Step 15.d]

BOPO When RCS THOT is 510°F, ISOLATE SG RC-2B. [Step 16]

• ISOLATE SG RC-2B per HR-20, Isolate/Restore Steam Generator B.

[Step 16.a]

Examiner Note: The following steps are from HR-20, Isolate/Restore Steam Generator B.

NOTE RCS Heat Removal takes precedence over isolation of a S/G with a tube rupture.

CRITICAL TASK Isolate the Affected Steam Generator with a Tube Rupture to Minimize Spread STATEMENT of Contamination Prior to Exiting EOP-04, Steam Generator Tube Rupture.

CRITICAL TASK BOPO PERFORM the following to isolate Steam Generator RC-2B: [Step 1]

• ENSURE all the following valves are CLOSED: [Step 1.a]

• CLOSE HCV-1042A, RC-2B MSIV.

• VERIFY HCV-1042C, RC-2B MSIV Bypass Valve CLOSED.

• VERIFY MS-292, Air Assisted Main Steam Safety Valve CLOSED.

• CLOSE FCV-1102, RC-2B Feed Regulating Valve.

• CLOSE HCV-1106, Feed Regulating Bypass Valve.

BOPO

• CLOSE HCV-1385, RC-2B Feed Header Isolation Valve.

• CLOSE HCV-1104, Feed Regulating Block Valve.

• VERIFY HCV-1387A, Blowdown Isolation Valve CLOSED.

• VERIFY HCV-1387B, Blowdown Isolation Valve CLOSED.

• CLOSE HCV-1108A, AFW Isolation Valve.

• CLOSE HCV-1108B, AFW Isolation Valve.

NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 7&8 Page 36 of 36 Event

Description:

Steam Generator Tube Rupture / Diesel Generator Start Failure Time Position Applicants Actions or Behavior

• CONTACT Auxiliary Operator to CLOSE MS-298, Steam Valves HCV-1041A & 1042A Packing Leakoff Line Isolation Valve in Room 81.

[Step 1.b]

• If sampling is NOT in progress, CLOSE both Sample Valves:

[Step 1.c]

• HCV-2507A, RC-2B Blowdown Sample Isolation Valve

• HCV-2507B, RC-2B Blowdown Sample Isolation Valve BOPO

• PERFORM the following to CLOSE YCV-1045B: [Step 1.d]

• DETERMINE Isolation Valve YCV-1045B OVERRIDE SW in OVERRIDE. [Step 1.d.1)]

• DETERMINE SG RC-2B STM TO FW-10 HDR A ISOLATION VALVE YCV-1045B in CLOSE. [Step 1.d.2)]

NOTE Air accumulators will maintain the valve in a closed position for 30 minutes after a loss of Instrument Air.

• CONTACT Auxiliary Operator to HANDJACK YCV-1045B, MAIN STEAM LINE "B" TO AUX FEEDWATER PUMP FW-10 SUPPLY VALVE to CLOSE in Room 81. [Step 1.d.3)]

CRS

• RECORD time Steam Generator RC-2B ISOLATED. [Step 1.e]

• Time: ________

VERIFY RC-2B is most affected SG per Attachment HR-18, Most Affected CRS Steam Generator Determination. [Step 2]

When Steam Generator RC-2B is isolated, TERMINATE the scenario.

NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Scenario Outline Form ES-D-1 Facility: Fort Calhoun Station Scenario No.: 4 Op Test No.: Dec 2015 NRC Examiners: Operators:

Initial Conditions: MODE 2 at ~1% power - RCS Boron is 959 ppm (by sample).

Turnover: Continue in OP-2A, Plant Startup and OI-RR-1, Reactor Regulating System Operation to raise Reactor power to 7% power. When MODE 1 is entered, place Steam Dump and Bypass Valves in AUTO per OI-MS-1A, Main Steam System Operation.

Critical Tasks:

• Upon Loss of Cooling to Reactor Coolant Pumps (RCPs), Trip the Reactor and Trip Associated RCP(s) within 5 minutes of Total Loss of CCW flow or Motor Bearing Temperatures > 203°F but prior to exceeding 210°F or within 5 minutes of CCW temperature

> 110°F. (Event 2). OR

• Manually Trip Reactor to meet Core Design Criteria of Lowering Reactor Power and Negative Startup Rate to Verify Reactivity Control Established During ATWS Event Prior to Exiting EOP-00, Standard Post Trip Actions. (Event 5)

• Isolate the Affected Steam Generator to Prevent Excess Plant Cooldown and Reactivity Additions Prior to Steam Generator Level = 0% Wide Range Level. (Event 7)

Event No. Malf. No. Event Type* Event Description 1 R (ATCO) Raise Power Using Control Rods to 7% per OP-2A, Plant Startup.

+20 min N (BOPO, CRS) Place Steam Dump and Bypass Valves in AUTO per OI-MS-1A.

2 C (ATCO, CRS) Raw Water Pump Discharge Line Leak Upstream of HCV-2879A in

+30 min TS (CRS) the Auxiliary Building.

3 I (BOPO, CRS) Inadvertent Channel B Auxiliary Feedwater Actuation Signal On

+45 min TS (CRS) Steam Generator RC-2A.

4 C (ATCO, CRS) Loss of Instrument Bus AI-40A.

+60 min TS (CRS) Loss of Letdown and Pressurizer Level Control.

(Alternate Path Event 8) 5 M (ATCO, BOPO, Reactor Coolant Pump RC-3A Trip.

+60 min CRS) Automatic Reactor Trip Failure, Manual Reactor Trip Required.

6 C (BOPO) Instrument Air Compressor CA-1B and CA-1C Trip.

+65 min Bearing Cooling Water Pump AC-9B Trip.

7 M (ATCO, BOPO, Steam Line Break inside Containment on RC-2A @ 0.65% Severity

+70 min CRS) on 5 Minute Ramp.

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Technical Specifications Actual Target Quantitative Attributes 1 Malfunctions after EOP entry (1-2) 3 Abnormal events (2-4) 2 Major transients (1-2) 1 EOPs entered/requiring substantive actions (1-2) 0 EOP contingencies requiring substantive actions (0-2) 2 Critical tasks (2-3)

NRC Simulator Scenario 4 Outline Rev. Final As Run

Scenario Event Description NRC Scenario 4 SCENARIO

SUMMARY

NRC 4 The crew will assume the shift at 1% power and raise Power to ~7% using CEAs per OP-2A, Plant Startup, Attachment 4, Hot Standby, MODE 2 to Minimum Load, MODE 1 and OI-RR-1, Reactor Regulating System Normal Operation, Attachment 7, Manual Sequential Mode Checklist. When MODE 1 is entered, temperature control is placed in AUTO per OI-MS-1A, Main Steam System Operation, , Steam Dump and Bypass Manual Control Function.

The next event is a Raw Water Pump AC-10C discharge line leak in the Auxiliary Building upstream of HCV-2879A. The crew enters AOP-18, Loss of Raw Water, and must observe Raw Water System indications in order to determine the location of the leak. Once identified, the leak is isolated per AOP-18, Attachment C, Equipment Isolation, and Raw Water flow is restored. If the leak is not isolated, the Reactor and affected RCPs will be tripped. The SRO will refer to Technical Specification LCO 2.4(1)

- Raw Water Header.

The next event is an inadvertent Channel B Auxiliary Feedwater Actuation Signal (AFAS) on Steam Generator RC-2A. The crew responds per ARP-AI-66B/A66B, Window 41 and verifies Auxiliary Feedwater Pumps FW-6 and FW-10 are running. Once it is determined the AFAS was inadvertent, AOP-23, Reset of Engineered Safeguards,Section IX, Reset of Inadvertent AFAS, is performed. The SRO will refer to Technical Specification LCO 2.15.1(1) - Automatic Initiation Steam Generator Water Level Logic Subsystem B.

When plant conditions are stable, a loss of Instrument Bus AI-40A occurs. The crew enters AOP-16, Loss of Instrument Bus Power,Section I, Loss of Instrument Bus Power, then Section II, Loss of Instrument Bus AI-40A. Actions include isolating Letdown, transferring Pressurizer Level Control, and operating Charging Pumps as required. Electrical Maintenance is notified and the Plant remains in this configuration through the end of the Scenario. The SRO will refer to Technical Specification LCO 2.15.2

- Reactor Protective System Logic and Trip Initiation and LCO 2.7(1) - 120 VAC Instrument Bus A.

The next event is a trip of Reactor Coolant Pump RC-3A. The crew should recognize failure of the Reactor Protection System Low Flow trips and manually trip the Reactor and enter EOP-00, Standard Post Trip Actions. When the Reactor is tripped, a Steam Line Break inside Containment initiates on a 5 minute ramp. Due to the small size of this break, RCS pressure remains above the SIAS initiation setpoint of 1600 psia. The crew will transition to EOP-05, Uncontrolled Heat Extraction, and identify and isolate the affected Steam Generator RC-2A.

The event is complicated by a trip of the running and standby Instrument Air Compressors CA-1B and CA-1C and a trip of Bearing Water Cooling Pump AC-9B. The crew must restore a Bearing Cooling Water Pump and Instrument Air Compressor while in EOP-00. The scenario is terminated when Steam Generator RC-2A is isolated per HR-19, Isolate/Restore Steam Generator A while in EOP-05.

Risk Significance:

• Failure of risk important system prior to trip: Loss of Raw Water System Header Loss of Instrument Bus

• Risk significant core damage sequence: Automatic Reactor Trip Failure Steam Line Break Inside Containment

• Risk significant operator actions: Isolate Raw Water East Header Manually Trip Reactor Restore Instrument Air Isolate Affected Steam Generator NRC Simulator Scenario 4 Outline Rev. Final As Run

Scenario Event Description NRC Scenario 4 BOOTH OPERATOR INSTRUCTIONS for SIMULATOR SETUP RESET to IC-122 and LOAD & EXECUTE NRC 4.sce for NRC Scenario 4.

Preset item - Event 5 - Reactor Fails to Trip Automatically, CB-4 Trip Button Works Type Item Value Condition Expert RPS02 Energized Scenario Event: Rx Fail to RPS01 Energized Trip, CB-4 works RPS03 Energized RPS04 Energized P6A_026_1 True P6B_028_1 True ANN-P6A_0026R1C_Fail Alarm Off ANN-P6A_0027R1C_Fail Alarm Off ANN-P6B_0026R5C_Fail Alarm Off ANN-P6B_0027R5C_Fail Alarm Off ANN-P6B_0025R5C_Fail Alarm Off ANN-P6A_0025R1C_Fail Alarm Off H_P6A_022A_1 True H_P6B_024A_1 True Event 2 - Raw Water leak in the Auxiliary Building Type Item Value Condition Malfunction RWS02B 25 When directed by examiner, trigger/activate this event.

Scenario Event: Raw Water Leak in Aux Building Event 3 - Inadvertent AFAS on RC-2A Type Item Value Condition Expert B_RC_2A_AFWS True When directed by examiner, trigger/activate this event.

Scenario Event:

Inadvertent AFAS Event 4 - Loss of Instrument Bus AI-40A Type Item Value Condition Malfunction EDA08 10 When directed by examiner, trigger/activate this event.

Scenario Event: Loss of AI-40A Event 5 - A Reactor Coolant Pump Trips Type Item Value Condition Malfunction BUS_1A1_5_BKR_TRIP True When directed by examiner, trigger/activate this event.

Scenario Event: A RCP Trip Event 6 - Following RX Trip, Loss of Instrument Air and Bearing Cooling Water NRC Simulator Scenario 4 Outline Rev. Final As Run

Scenario Event Description NRC Scenario 4 Type Item Value Condition Remote BCW_AC9B_BRKR Trip Event is triggered Malfunction BUS_1B3A_4A_2_BKR_Trip True automatically after reactor BUS_1B4B_4_BKR_TRIP True trip. Scenario Event: Loss of Inst Air and Bearing Water Event 7 - Main Steam Break Inside Containment Type Item Value Condition Malfunction SGN01A 0.65% Event is triggered Ramp = 300 sec automatically after reactor trip. Scenario Event:

Steam Line Break in Containment NRC Simulator Scenario 4 Outline Rev. Final As Run

Scenario Event Description NRC Scenario 4 Booth Operator: INITIALIZE to IC-122 and LOAD NRC 4.sce.

ENSURE all Simulator Annunciator Alarms are ACTIVE.

ENSURE all Control Board Tags are removed.

ENSURE Bearing Water Pump AC-9B running.

ENSURE Charging Mode Select Switch is in CH-1A - CH-1C position.

ENSURE Turbine speed is approximately 3 RPM.

ENSURE Air Compressors CA-1B & CA-1C alignment: 1 in Standby, 1 running.

PLACE Steam Dump & Bypass Controllers in Manual.

ENSURE Lead Examiner has AFAS Keys 55 & 57 for Event 3.

ENSURE Lead Examiner has RPS Trip Unit Keys 1-12 for Event 4.

ENSURE Operator Aid Tags reflect current boron conditions.

ENSURE Reactivity Briefing Sheet printout provided with Turnover.

ENSURE Middle-of-Life Thumb Rule Sheet provided with Turnover.

ENSURE Steam Dump and Turbine Bypass System in MANUAL control.

ENSURE Control Room hard copy for OI-RR-1 is CLEAN.

ENSURE CEA Regulating Group 4 @ 72.

ENSURE procedures in progress provided to crew in Briefing Room:

- COPY of ReMA Data for Reactor Power Ascension.

- COPY of OP-2A, Plant Startup, Attachment 4, Hot Standby, MODE 2 to Minimum Load, MODE 1, INITIALED through Step 6.b.

- COPY of OI-RR-1, Reactor Regulating System Normal Operation, Attachment 7, Manual Sequential Mode Checklist.

- Copy of OI-MS-1A, Main Steam System Operation, Attachment 5, Steam Dump and Bypass Manual Control Function, INITIALED through Prerequisites and Steps 1.a & 2.a.

Control Room Annunciators in Alarm:

A9-B-1(U) - TURBINE DIFFERENTIAL EXPANSION A10-A-1(U) - MOTOR SUCT PUMP RUNNING OR NOT IN AUTO A10-B-6(L) - 43/FW TRANSFER SWITCH OFF-AUTO A11-A-4(U) - HEATER 5A HEATER HI-LO A11-A-4(L) - HEATER 5B HEATER HI-LO A11-B-3(U) - HEATER DRAIN TANK LEVEL HI-LO A20-D LOSS OF LOAD CHANNEL TRIP BYPASSED A20-E HIGH POWER RATE OF CHANGE TRIP ENABLED A21-B-1(U) - HC-909 INHIBIT A21-C-6(U) - HEATING STEAM PRESS LO AI-66B/A66B-Window 3 - FW-10 TURBINE DRIVEN FEEDWATER PUMP RUNNING NRC Simulator Scenario 4 Outline Rev. Final As Run

Scenario Event Description NRC Scenario 4 Procedure List Event 1: OP-2A, Plant Startup, Attachment 4, Hot Standby, MODE 2 to Minimum Load, MODE 1 Event 1: OI-RR-1, Reactor Regulating System Normal Operation, Attachment 7, Manual Sequential Mode Checklist Event 1: OI-MS-1A, Main Steam System Operation, Attachment 5, Steam Dump and Bypass Manual Control Function Event 2: AOP-18, Loss of Raw Water Event 3: ARP-AI-66B/A66B, Window 41, AFWS STEAM GEN RC-2A CHANNEL B ACTUATED Event 3: AOP-23, Reset of Engineered Safeguards,Section IX, Reset of Inadvertent AFAS Event 3: OI-AFW-2, Auxiliary Feedwater System Actuation and Bypass, Attachment 1, Bypass of the Auxiliary Feedwater Actuation Signal (AFAS) (Modes 1 or 2)

Event 3: OI-AFW-2, Auxiliary Feedwater System Bypass, Table 2, AFAS Logic Subsystem Channel Bypass Switch Alignment Event 4: AOP-16, Loss of Instrument Bus Power,Section I - Loss of Instrument Bus Power Event 4: AOP-16, Loss of Instrument Bus Power,Section II, Loss of Instrument Bus AI-40A Event 5: EOP-00, Standard Post Trip Actions Event 7: EOP-05, Uncontrolled Heat Extraction NRC Simulator Scenario 4 Outline Rev. Final As Run

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 1 Page 7 of 37 Event

Description:

Raise Reactor Power Time Position Applicants Actions or Behavior Booth Operator: When directed, RESPOND to requests from Control Room.

Examiner Note: This Scenario Section contains guidance for the following Operator actions:

1. Raising power per OP-2A.

2. Withdrawing Control Rods per OI-RR-1.

3. Control of Steam Dumps and Bypass per OI-MS-1A.

Examiner Note: The following steps are from OP-2A, Plant Startup, Attachment 4, Hot Standby, MODE 2 to Minimum Load, MODE 1, Step 6.

RAISE Reactor power to ~ 10% while performing the following: [Step 6]

• DETERMINE Main Feedwater Pump FW-4B is RUNNING. [Step 6.a]

• DETERMINE Auxiliary Feedwater SECURED. [Step 6.b]

• MAINTAIN RCS temperature 527°F to 535°F using Steam Dump and Bypass Valves. [Step 6.c]

• Prior to exceeding 15% power, VERIFY Secondary Chemistry parameters. [Step 6.d]

• Prior to exceeding 15% power, VERIFY Condensate Pump Discharge Suspended Solids within specification. [Step 6.e]

• PERFORM daily grab samples for Secondary activity or DECLARE RM-057 Radiation Monitor in service. [Step 6.f]

NOTE This step is performed to ensure that the DVM NI indication is greater than or equal to actual power.

• When power is stable at approximately 10% (as indicated by highest of NI and T power), ADJUST RPS power per OI-NI-1. [Step 6.g]

• OPEN MFW Isolation Valves HCV-1103 & HCV-1104. [Step 6.h]

Examiner Note: The following steps are from OI-RR-1, Reactor Regulating System Normal Operation, Attachment 7, Manual Sequential Mode Checklist, and is maintained as a Control Room hard copy.

ENSURE an out-of-scan CEA is NOT selected as Target Rod on CB-4.

ATCO

[Step 1]

VERIFY alarm REGULATING GROUP WITHDRAWAL PROHIBIT is clear.

ATCO

[Step 2]

NRC Simulator Scenario 4 Outline Rev. Final As Run

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 1 Page 8 of 37 Event

Description:

Raise Reactor Power Time Position Applicants Actions or Behavior PLACE Rod Control Mode Selector Switch in Manual Sequential (MS).

ATCO

[Step 3]

NOTE Continuous CEA motion shall be avoided whenever possible. CEA motion should be stopped at least every 33 inches (43 seconds of continuous CEA motion) to check position of CEAs in Group and Reactor response.

MOVE Manual Rod Control Switch to RAISE or LOWER as required.

ATCO

[Step 4]

DETERMINE appropriate Group Overlap during WITHDRAWAL is N/A.

ATCO

[Step 5]

When CEAs are at desired position, RELEASE Manual Rod Control Switch.

ATCO

[Step 6]

ATCO VERIFY all CEA motion has stopped. [Step 7]

ATCO If additional movement is required, GO TO Step 3. [Step 10]

When completed, PLACE Rod Control Mode Selector Switch in OFF.

ATCO

[Step 11]

Examiner Note: The following steps are from OI-MS-1A, Main Steam System Operation, Attachment 5, Steam Dump and Bypass Manual Control Function.

If operating all Steam Dump and Bypass Valves via the Pressure Controller BOPO (PC0910), PERFORM the following (SEC/MS/SD&B Control): [Step 1]

• DETERMINE PC0910, STM DMP & BYP PRESS CONTROL, in MANUAL control. [Step 1.a]

• As required, ADJUST output to Steam Dump and Bypass Valves.

[Step 1.b]

NRC Simulator Scenario 4 Outline Rev. Final As Run

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 1 Page 9 of 37 Event

Description:

Raise Reactor Power Time Position Applicants Actions or Behavior Booth Operator: When power has been raised approximately 3%, and prior to transitioning to the next event, CONTACT the Control Room as the Shift Manager and direct placing Steam Dump and Turbine Bypass System (pressure and temperature control) in AUTO.

• If desired to transfer back to AUTO at Output that has been selected, BOPO COMPLETE the following on Digital Control System: [Step 1.c]

• PLACE PC0910 in LOCAL. [Step 1.c.1)]

• ADJUST PC0910 SPT to approximately match PC0910 MEAS value.

[Step 1.c.2)]

• PLACE PC0910 back in AUTO. [Step 1.c.3)]

If operating all Steam Dump and Bypass Valves via the Temperature BOPO Controller (TC0909_PI), PERFORM the following (SEC/MS/SD&B Control):

[Step 2]

• DETERMINE TC0909_PI, STM DMP & BYP TEMP CONTROL, in MANUAL control. [Step 2.a]

• As required, ADJUST output to Steam Dump and Bypass Valves.

[Step 2.b]

• If desired to transfer back to AUTO at Output that has been selected,

+20 min BOPO COMPLETE the following on Digital Control System: [Step 2.c]

• PLACE TC0909_PI in LOCAL. [Step 2.c.1)]

• ADJUST TC0909_PI SPT to approximately match TC0909_PI MEAS value. [Step 2.c.2)]

• PLACE TC0909_PI back in AUTO. [Step 2.c.3)]

When Reactor power is raised 3% to 5%, PROCEED to Event 2.

NRC Simulator Scenario 4 Outline Rev. Final As Run

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 2 Page 10 of 37 Event

Description:

Raw Water Pump Discharge Line Leak Time Position Applicants Actions or Behavior Booth Operator: When directed, EXECUTE Event 2.

- Raw Water Pump discharge line leak upstream of HCV-2879A.

Indications Available:

CB-1,2,3/A1 - RAW WATER SUPPLY HEADER FLOW LO CB-1,2,3/A1 - RAW WATER SUPPLY HEADER PRESS LO All Raw Water System 10 psig and 25 psig pressure indicating lights OUT

+30 sec ATCO RESPOND to Annunciator Response Procedures.

ATCO INFORM CRS of Raw Water System low pressure and low flow.

Examiner Note: ATCO may Operate to Mitigate per OPD 4-09 and START another Raw Water Pump.

CRS REFER to AOP-18, Loss of Raw Water.

Examiner Note: The following steps are from AOP-18, Loss of Raw Water.

ATCO DETERMINE Raw Water Pump AC-10C is RUNNING. [Step 4.1]

Booth Operator: If not already contacted, 1 minute after Control Room Receipt of alarms, REPORT as Auxiliary Building Operator that he observed water flowing out of Room 18, and he is going in to investigate.

WAIT 30 seconds and REPORT Raw Water System leak in Room 18, upstream of HCV-2879A/B on the header side of the system.

If Raw Water System rupture is indicated, DIRECT Operators to identify ATCO location of leak: [Step 4.2]

• OBSERVE East RW Header Flow FIC-2890 OSCILLATING.

• OBSERVE West RW Header Flow FIC-2891 OSCILLATING.

• OBSERVE RW Pump(s) Current OSCILLATING.

• OBSERVE RW System Pressure PIC-2892 OSCILLATING.

• OBSERVE RW Pump Room Water Level LIC-2889/LC-2825 Level NORMAL.

NRC Simulator Scenario 4 Outline Rev. Final As Run

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 2 Page 11 of 37 Event

Description:

Raw Water Pump Discharge Line Leak Time Position Applicants Actions or Behavior ATCO DETERMINE Raw Water vault flooding is NOT occurring. [Step 4.3]

DETERMINE Raw Water leak in Auxiliary Building and PERFORM the ATCO following: [Step 4.4]

• ENSURE only one Raw Water Pump RUNNING. [Step 4.4.a]

• IMPLEMENT Attachment C, Equipment Isolation. [Step 4.4.b]

Examiner Note: If the leak has NOT been isolated and another Raw Water Pump NOT started, the crew will either continue in AOP-18 (CCW temperature > 110°F) OR transition to AOP-35, RCP Malfunctions (RCP motor bearing temps > 203°F).

ATCO DETERMINE CCW temperature 110°F. [Step 4.5]

CRS * [CA] If CCW temperature > 110°F, GO TO Step 10. [Step 4.5.1]

CRS IMPLEMENT the Emergency Plan. [Step 4.6]

Examiner Note: The following steps are from AOP-18, Attachment C, Equipment Isolation.

CRS If leak is on Raw Water System, GO TO Step 8. [Step 1]

NOTE The leak isolation Steps 8 through 15 may be performed in any logical order.

ATCO DETERMINE leak is NOT on any of the following: [Step 8]

• WEST RW header

• AC-12A, Raw Water Strainer

• AC-1C, RW Heat Exchanger DETERMINE leak is on East Raw Water Header and PERFORM the ATCO following to ISOLATE Header: [Step 9]

• PLACE AC-10D, Raw Water Pump, in PULL-TO-LOCK. [Step 9.a]

NRC Simulator Scenario 4 Outline Rev. Final As Run

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 2 Page 12 of 37 Event

Description:

Raw Water Pump Discharge Line Leak Time Position Applicants Actions or Behavior ATCO

• CLOSE all Raw Water Header Isolation Valves: [Step 9.b]

• CLOSE HCV-2876A.

• CLOSE HCV-2876B.

• CLOSE HCV-2894.

• CLOSE HCV-2879A.

• CLOSE HCV-2879B.

• CLOSE HCV-2883A.

• CLOSE HCV-2883B.

Booth Operator: When contacted, REPORT RW-145 is CLOSED.

When contacted, EXECUTE local actions and REPORT handjacks applied to Raw Water System Valves as directed.

• Locally CLOSE RW-145, RAW WATER STRAINER AC-12B ATCO BACKWASH VALVE HCV-2805B OUTLET ISOLATION VALVE in RW Vault. [Step 9.c]

• DETERMINE leak is isolated and one Raw Water Pump RUNNING.

CRS

[Step 9.d]

Examiner Note: The following steps continue from AOP-18.

CRS DETERMINE Raw Water System restored to service. [Step 4.8]

CRS EVALUATE Technical Specification LCO 2.4, Containment Cooling

• LCO 2.4.(1).a.iv - Raw Water Header

• CONDITION 2.4.(2).d - Raw Water Header inoperable.

• ACTION 2.4.(2).d - RESTORE Raw Water Header within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> OR PLACE Reactor in HOT SHUTDOWN condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Examiner Note: If the leak has been isolated and Raw Water is restored, CONTINUE to the next event.

Examiner Note: If the leak has NOT been isolated and another Raw Water Pump NOT started, the crew will either continue in AOP-18 (CCW temperature > 110°F) OR transition to AOP-35, RCP Malfunctions (RCP motor bearing temps > 203°F).

NRC Simulator Scenario 4 Outline Rev. Final As Run

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 2 Page 13 of 37 Event

Description:

Raw Water Pump Discharge Line Leak Time Position Applicants Actions or Behavior Examiner Note: The following steps continue from AOP-18.

CRS DETERMINE Raw Water System NOT restored to service. [Step 4.9]

CRS If the Reactor is critical, PERFORM the following: [Step 4.10]

• TRIP the Reactor. [Step 4.10.a]

• IMPLEMENT EOP-00, Standard Post Trip Actions. [Step 4.10.b]

Upon Loss of Cooling to Reactor Coolant Pumps (RCPs), Trip the Reactor and Trip Associated RCP(s) within 5 minutes of Total Loss of CCW flow or Motor CRITICAL TASK Bearing Temperatures > 203°F but prior to exceeding 210°F or within 5 minutes STATEMENT of CCW temperature > 110°F.

Time CCW > 110°F: _____ minutes.

CRITICAL TASK ATCO

• Manually TRIP Reactor at CB-4.

Examiner Note: The following steps are from AOP-35, RCP Malfunctions,Section II, Motor Bearing System Failures.

CRS VERIFY none of the following conditions exist: [Step 4.1]

• Motor guide or thrust bearing temperatures > 203°F for RC-3A/3C/3D.

• Motor guide or thrust bearing temperatures > 230°F for RC-3B.

• [CA] If any bearing temperature exceeds its limit and the Reactor ATCO is critical, PERFORM the following: [Step 4.1.1]

• [CA] TRIP the Reactor. [Step 4.1.1.a]

• [CA] IMPLEMENT EOP-00, Standard Post Trip Actions.

[Step 4.1.1.b]

• [CA] STOP the affected RCPS. [Step 4.1.1.c]

• [CA] GO TO Section 5.0, Exit Conditions. [Step 4.1.1.d]

NRC Simulator Scenario 4 Outline Rev. Final As Run

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 2 Page 14 of 37 Event

Description:

Raw Water Pump Discharge Line Leak Time Position Applicants Actions or Behavior Upon Loss of Cooling to Reactor Coolant Pumps (RCPs), Trip Associated RCP(s) within 5 minutes of Total Loss of CCW flow or Motor Bearing CRITICAL TASK Temperatures > 203°F but prior to exceeding 210°F or within 5 minutes of CCW STATEMENT temperature > 110°F.

Time RCPs exceeding > 203°F: _____ minutes CRITICAL TASK ATCO

• Manually TRIP Reactor at CB-4.

CRITICAL TASK ATCO

• Manually TRIP any affected Reactor Coolant Pumps.

When the Reactor and RCPs have been tripped, PROCEED to Events 6 & 7.

NRC Simulator Scenario 4 Outline Rev. Final As Run

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 3 Page 15 of 37 Event

Description:

Inadvertent Auxiliary Feedwater Actuation Signal Time Position Applicants Actions or Behavior Booth Operator: When directed, EXECUTE Event 3.

- Inadvertent Auxiliary Feedwater Actuation Signal.

Indications Available:

AI-66B/A66B - AFWS STEAM GEN RC-2A CHANNEL B ACTUATED AI-66B/A66B - FW-10 TURBINE DRIVEN FEEDWATER PUMP RUNNING (~30 seconds later)

AI-66B/A66B - FW-10 TURBINE OIL PUMP RUNNING (~30 seconds later)

+30 sec BOPO RESPOND to Annunciator Response Procedures.

BOPO INFORM CRS of Auxiliary Feedwater Actuation Signal initiation.

Examiner Note: BOPO may Operate to Mitigate per OPD 4-09 and CLOSE HCV-1107A and HCV-1107B to stop FW-10, Turbine Driven Auxiliary Feedwater Pump.

REFER to ARP-AI-66B/A66B, Window 41, AFWS STEAM GEN RC-2A CRS CHANNEL B ACTUATED.

Examiner Note: The following steps are from ARP-AI-66B/A66B, Window 41 - AFWS STEAM GEN RC-2A CHANNEL B ACTUATED.

CHECK A/B/LI-911, Steam Generator RC-2A Level at AI-66A and AI-66B.

BOPO

[Step 1]

• DETERMINE SG level LI-911A at Panel AI-66A NORMAL.

• DETERMINE SG level LI-911B at Panel AI-66B NORMAL.

Booth Operator: When contacted, REPORT LI-911D, RC-2A level at AI-179 is ~ 64% and LI-911C, RC-2A pressure is ~ 884 psia (or as indicated).

BOPO DISPATCH Operator to check C/D/LI-911, RC-2A Level at AI-179. [Step 2]

BOPO DETERMINE Steam Generator Wide Range level is > 32%. [Step 3]

DETERMINE AFAS initiation is inadvertent and IMPLEMENTS AOP-23, CRS Reset of Engineered Safeguards,Section IX, Reset of Inadvertent AFAS.

[Step 4]

NRC Simulator Scenario 4 Outline Rev. Final As Run

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 3 Page 16 of 37 Event

Description:

Inadvertent Auxiliary Feedwater Actuation Signal Time Position Applicants Actions or Behavior CRS REFER to Technical Specification LCOs 2.14 and 2.15. [Step 5]

EVALUATE Technical Specification LCO 2.15.1, Instrumentation and Control CRS Systems

• LCO 2.15.1.(3) - Automatic Initiation Steam Generator Water Level Logic Subsystem B (Table 2.3 / Function 5).

• CONDITION 2.15.1.(3) - Logic Subsystem B inoperable

• ACTION 2.15.1.(3) - RESTORE inoperable channel within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> OR PLACE Reactor in HOT SHUTDOWN condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Examiner Note: The following steps are from AOP-23, Reset of Engineered Safeguards,Section IX, Reset of Inadvertent AFAS.

CRS DETERMINE the AFAS is inadvertent. [Step 4.1]

CRS REFER to the following Technical Specifications: [Step 4.2]

• LCO 2.5, Steam and Feedwater Systems

• LCO 2.15, Instrumentation and Control Systems Examiner Note: Entry into Technical Specification LCO 2.5.(1).d is required until FW-10, TDAFW Pump is reset and returned to AUTO at the end of this event.

EVALUATE Technical Specification LCO 2.5, Steam and Feedwater CRS Systems

• LCO 2.5.(1) - Two AFW Trains OPERABLE

• CONDITION 2.5.(1).d - Both AFW Trains inoperable

• ACTION 2.5.(1).d - RESTORE one train to OPERABLE status immediately.

BOPO ENSURE both of the following valves in AUTO: [Step 4.3]

• DETERMINE FCV-1368, FW-6 Recirc Valve in AUTO.

• DETERMINE FCV-1369, FW-10 Recirc Valve in AUTO.

NRC Simulator Scenario 4 Outline Rev. Final As Run

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 3 Page 17 of 37 Event

Description:

Inadvertent Auxiliary Feedwater Actuation Signal Time Position Applicants Actions or Behavior PLACE control switches for the following AFW Isolation Valves in CLOSE:

BOPO

[Step 4.4]

• PLACE HCV-1107A in CLOSE.

• PLACE HCV-1107B in CLOSE.

• PLACE HCV-1108A in CLOSE.

• PLACE HCV-1108B in CLOSE.

BYPASS affected logic subsystem per OI-AFW-2, Auxiliary Feedwater CRS System Actuation and Bypass. [Step 4.5]

Examiner Note: The following steps are from OI-AFW-2, Auxiliary Feedwater System Actuation and Bypass, Attachment 1, Bypass of the Auxiliary Feedwater Actuation Signal (AFAS) (Modes 1 or 2).

BOPO DETERMINE AFAS is aligned for automatic initiation. [Step 2]

BOPO DETERMINE plant is in Mode 1. [Step 3]

DETERMINE if an Instrument Channel or a Logic Subsystem Channel is to CRS be bypassed. [Step 1]

• DETERMINE an Instrument Channel will NOT be bypassed. [Step 1.a]

• DETERMINE a Logic Subsystem Channel will be bypassed and GO TO Step 3. [Step 1.b]

If a Logic Subsystem Channel of AFAS is to be bypassed, COMPLETE the CRS following: [Step 3]

SM/CRS

• LOG entry into Technical Specification 2.15.1(3), 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> LCO.

NRC Simulator Scenario 4 Outline Rev. Final As Run

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 3 Page 18 of 37 Event

Description:

Inadvertent Auxiliary Feedwater Actuation Signal Time Position Applicants Actions or Behavior NOTE The following alarms are expected depending on the Logic Subsystem Channel that is bypassed:

• AFWS RC-2A CH A MATRIX TS-A/RC-2A/AFWS TEST SWITCH OFF NORM (AI-66A, Window 24)

• AFWS RC-2B CH A MATRIX TS-A/RC-2B/AFWS TEST SWITCH OFF NORM (AI-66A, Window 25)

• AFWS OVERRIDE SWITCH A/OR-RC-2A/AFWS OFF NORMAL (AI-66A, Window 29)

• AFWS OVERRIDE SWITCH A/OR-RC-2B/AFWS OFF NORMAL (AI-66A, Window 30)

• HCV-1107A & B AFWS OVERRIDE SWITCH CH A OR B OFF NORM (AI-66A, Window 35)

• AFWS RC-2A CH B MATRIX TS-B/RC-2A AFWS TEST SWITCH OFF NORM (AI-66B, Window 21)

• AFWS RC-2B CH B MATRIX TS-B/RC-2B/AFWS TEST SWITCH OFF NORM (AI-66B, Window 22)

• AFWS OVERRIDE SWITCH B/OR-RC-2A/AFWS OFF NORMAL (AI-66B, Window 26)

• AFWS OVERRIDE SWITCH B/OR-RC-2B/AFWS OFF NORMAL (AI-66B, Window 27)

• HCV-1108A & B AFWS OVERRIDE SWITCH CHA OR B OFF NORMAL (AI-66A, Window 32)

BYPASS selected Logic Subsystem using Table 2, AFAS Logic Subsystem BOPO Bypass Switch Alignment, and RECORD as left information in appropriate slots. [Step 3.b]

Examiner Note: The following steps are from OI-AFW-2, Table 2, AFAS Logic Subsystem Channel Bypass Switch Alignment.

Table 2 - AFAS Logic Subsystem Channel Bypass Switch Alignment As-Left Switch Bypassing Channel Panel No. Switch Position Position RC-2A Channel B AI-66B S/G RC-2A Chan. B Auto Sig Bypass (Amber lamps S/G RC- Override Relay Test Sw 2A Chan B/B1)

S/G RC-2A Chan. B Auto Sig Override Override Sw AFW Pumps FW-6/FW-10 Chan. B AFW Auto Sig B/OR -1107 Override S/G Feed Valves AFWS Examiner Note: Acting as Shift Manager, PROVIDE Keys #55 and #57 when requested.

BOPO PERFORM the following at Panel AI-66B for RC-2A Channel B:

• INSERT key #57 and PLACE S/G RC-2A Channel B Auto Signal Override Relay Test Switch in BYPASS.

NRC Simulator Scenario 4 Outline Rev. Final As Run

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 3 Page 19 of 37 Event

Description:

Inadvertent Auxiliary Feedwater Actuation Signal Time Position Applicants Actions or Behavior

• INSERT key #55 and PLACE S/G RC-2A Channel B Auto Signal Override Switch AFW Pumps FW-6/FW-10 in OVERRIDE.

• PLACE Channel B AFW Auto Signal Override S/G Feed Valves to B/OR

-1107 AFWS position.

Examiner Note: The following steps continue from AOP-23,Section IX, Reset of Inadvertent AFAS.

BOPO PERFORM the following to STOP all AFW Pumps: [Step 4.6]

• CLOSE YCV-1045, FW-10 Steam Inlet Valve. [Step 4.6.a]

• PLACE both Override Switches in OVERRIDE: [Step 4.6.b]

• ISOLATION VALVE YCV-1045A OVERRIDE SW.

• ISOLATION VALVE YCV-1045B OVERRIDE SW.

• CLOSE both FW-10 Steam Supply Valves: [Step 4.6.c]

• YCV-1045A, RC-2A to FW-10 Isolation Valve.

• YCV-1045B, RC-2B to FW-10 Isolation Valve.

• ENSURE FIC-1369, AUX FW PUMP FW-10 SUCTION FLOW drops to zero. [Step 4.6.d]

• STOP FW-6, Electric AFW Pump, and PLACE HC-1367, FW-6 Control Switch, in PULL-TO-LOCK. [Step 4.6.e]

• ENSURE FIC-1368, AUX FW PUMP FW-6 SUCTION FLOW drops to zero. [Step 4.6.f]

PERFORM the following to return the AFW System to automatic operation:

BOPO

[Step 4.7]

• PLACE Control Switches for AFW Isolation Valves in RESET:

[Step 4.7.a]

• PLACE HCV-1107A in RESET.

• PLACE HCV-1107B in RESET.

• PLACE HCV-1108A in RESET.

• PLACE HCV-1108B in RESET.

• PLACE Control Switches for AFW Isolation Valves in AUTO: [Step 4.7.b]

NRC Simulator Scenario 4 Outline Rev. Final As Run

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 3 Page 20 of 37 Event

Description:

Inadvertent Auxiliary Feedwater Actuation Signal Time Position Applicants Actions or Behavior

• PLACE HCV-1107A in AUTO.

• PLACE HCV-1107B in AUTO.

• PLACE HCV-1108A in AUTO.

• PLACE HCV-1108B in AUTO.

• PLACE Control Switch for YCV-1045, FW-10 Steam Inlet Valve, in RESET. [Step 4.7.c]

• PLACE Control Switch for YCV-1045, FW-10 Steam Inlet Valve, in AUTO. [Step 4.7.d]

• PLACE both Override Switches in NORMAL. [Step 4.7.e]

• ISOLATION VALVE YCV-1045A OVERRIDE SW

• ISOLATION VALVE YCV-1045B OVERRIDE SW

• PLACE HC-1367, FW-6 Control Switch, in AFTER-STOP. [Step 4.7.f]

Booth Operator: When contacted, EXECUTE remote functions to RESET FW-10 and Trip Latch Clamp is finger tight.

CONTACT Auxiliary Operator ENSURE FW-64-RL, AUX FEED PUMP BOPO FW-10 MANUAL TRIP LATCH RESET LEVER is latched: [Step 4.8]

• VERIFY Reset Lever is seated.

• ENSURE FW-64-C, AUX FEED PUMP FW-10 MANUAL TRIP LATCH CLAMP is installed finger tight.

CRS EXIT Technical Specification LCO 2.5, Steam and Feedwater. [Step 4.9]

When AFAS has been RESET, PROCEED to Event 4.

NRC Simulator Scenario 4 Outline Rev. Final As Run

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 4 Page 21 of 37 Event

Description:

Loss of Instrument Bus Time Position Applicants Actions or Behavior Booth Operator: When directed, EXECUTE Event 4. (Alternate Path Event 8)

- Loss of Instrument Bus AI-40A.

Indications Available:

CB-20/A15 - INVERTER A TROUBLE CB-20/A15 - INSTRUMENT BUS A LOW VOLTAGE/GROUND (~10 seconds later)

Multiple Loss of Instrument Bus alarms

+30 sec BOPO RESPOND to Annunciator Response Procedures.

CREW INFORM CRS of Loss of Instrument Bus AI-40A.

Booth Operator: When contacted, REPORT Inverter A Output Breaker is TRIPPED.

REFER to AOP-16, Loss of Instrument Bus Power,Section I, Loss of CRS Instrument Bus Power.

Examiner Note: The following steps are from AOP-16, Loss of Instrument Bus Power,Section I, Loss of Instrument Power.

CRS DETERMINE a Reactor Trip has NOT occurred: [Step 4.1]

CRS DETERMINE appropriate AOP-16 Section: [Step 4.2]

• OBSERVE an INVERTER A TROUBLE alarm.

• OBSERVE an INSTRUMENT BUS A LOW VOLTAGE/GROUND alarm.

CRS GO TO AOP-16,Section II, Loss of Instrument Bus AI-40A.

Examiner Note: The following steps are from AOP-16, Loss of Instrument Bus Power,Section II, Loss of Instrument Bus AI-40A.

CRS VERIFY Loss of Instrument Bus AI-40A by the following: [Step 4.1]

• INVERTER A TROUBLE alarm.

• INSTRUMENT BUS A LOW VOLTAGE/GROUND alarm.

NRC Simulator Scenario 4 Outline Rev. Final As Run

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 4 Page 22 of 37 Event

Description:

Loss of Instrument Bus Time Position Applicants Actions or Behavior NOTE

1. Upon loss of Instrument Bus A, ALL of the following instrumentation or equipment associated with the Reactivity Control Safety Function is affected as follows:

• All RPS Channel A is in trip

• Channel A "VARIABLE OVER POWER TRIP POWER MARGIN A/JI-007" meter is inoperable

• Channel A Wide Range Log Power Meter and Rate Meter are inoperable

• The Diverse Scram System is in half-trip

• Three RPS Logic Matrix channels are inoperable

2. Loss of more than one RPS Logic Matrix channel requires entry into T.S. 2.15.2.

3. If the associated clutch power supply is selected to Instrument Bus A then two RPS Trip Initiation Logic channels (AB, AC, AD) are inoperable.

DETERMINE clutch power supply selected to AI-40A and VERIFY clutch ATCO power supply is DEENERGIZED: [Step 4.2]

• OBSERVE AI-3-PS1 output current is 0.

• OBSERVE AI-3-PS3 output current is 0.

• OBSERVE AI-3-PS1 Indicating lights are out.

• OBSERVE AI-3-PS3 Indicating lights are out.

• OBSERVE clutch power supply breaker in half trip position.

Examiner Note: Acting as Shift Manager, PROVIDE Trip Unit Keys #1 to #12 when requested.

ATCO INSERT keys and BYPASS all RPS Channel A Bistable Trip Units. [Step 4.3]

CRS COMPLY with Technical Specification 2.15.2(5). [Step 4.4]

EVALUATE Technical Specification LCO 2.15, Instrumentation and Control CRS Systems

• LCO 2.15.2 - Reactor Protective System Logic and Trip Initiation

• CONDITION 2.15.2.(2) - One RPS Trip Initiation Logic channel inoperable.

• ACTION 2.15.2.(2) - Deenergize the affected clutch power supply within one hour (in 1/2 trip).

• ACTION 2.15.2.(5) - With the required actions of (2) not met, be in HOT SHUTDOWN and verify no more than one CEA is capable of being withdrawn within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

NRC Simulator Scenario 4 Outline Rev. Final As Run

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 4 Page 23 of 37 Event

Description:

Loss of Instrument Bus Time Position Applicants Actions or Behavior NOTE Upon loss of Instrument Bus A, ALL of the following instrumentation or equipment associated with the Vital Auxiliaries Safety Function are inoperable:

• "WEST RW SUPPLY HEADER FLOW FIC-2891" indicator

• "CC HT EXCH AC-1A RW OUTLET TEMP TIC-2885"

• "CNTMT CLG COIL VA-1A OUTLT ISOL VLV CNTRLR HCV-400C"

• "CNTMT CLG COIL VA-1B OUTLT ISOL VLV CNTRLR HCV-401C"

• "CNTMT CLG COIL VA-8A OUTLT ISOL VLV CNTRLR HCV-402C"

• "CNTMT CLG COIL VA-8B OUTLT ISOL VLV CNTRLR HCV-403C" ATCO ENSURE CCW System operation satisfactory: [Step 4.5]

• DETERMINE one CCW Pump RUNNING.

• DETERMINE CCW pressure 60 psig.

ATCO DETERMINE one Raw Water Pump RUNNING. [Step 4.6]

BOPO DETERMINE Instrument Air pressure 90 psig. [Step 4.7]

NOTE Upon loss of Instrument Bus A, ALL of the following instrumentation or equipment associated with the RCS Inventory Control Safety Function is affected as follows:

• Letdown is isolated

• Pressurizer Level Channel X is inoperable

• CH-1A, Charging Pump, is inoperable

• Charging Pump Backup Auto starts are disabled MAINTAIN Pressurizer level between 30% and 70% and TRENDING to ATCO between 45% percent by operating Charging Pumps CH-1B and/or CH-1C per IC-11, Inventory Control. [Step 4.8]

ATCO CLOSE TCV-202, Letdown Isolation Valve. [Step 4.9]

Examiner Note: At Chief Examiner discretion, PROCEED to the next event.

NRC Simulator Scenario 4 Outline Rev. Final As Run

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 4 Page 24 of 37 Event

Description:

Loss of Instrument Bus Time Position Applicants Actions or Behavior PLACE HC-101, Pressurizer Level Channel Selector Switch, in CHAN Y ATCO position. [Step 4.10]

NOTE Upon loss of Instrument Bus A, ALL of the following instrumentation or equipment associated with the RCS Pressure Control Safety Function is affected as follows:

• RCS Pressure Control Channel X is inoperable

• "PRESSURIZER PRESSURE A/PIA-102X AND A/PIA-102Y" indicators are inoperable

• PZR Backup Heaters are on

• PZR Heater Cutout is inoperable PLACE HC-103, Pressurizer Pressure Channel Selector Switch in CHAN Y ATCO position. [Step 4.11]

Manually CONTROL Pressurizer Heaters per PC-11, Pressure Control.

ATCO

[Step 4.12]

MAINTAIN RCS pressure per PC-12, RCS Pressure-Temperature Limits.

ATCO

[Step 4.13]

NOTE

1. Only one additional channel trip is needed to actuate the PORVs, even if the channel in trip is bypassed.

2. When RCS Heatup or Cooldown is in progress, the PORVs are the primary means of Low Temperature Overpressure Protection.

3. Closing the PORV block valves requires entry into Tech Spec 2.1.6.

CRS CONSIDER closing PORV Block Valves HCV-150 and HCV-151. [Step 4.14]

NRC Simulator Scenario 4 Outline Rev. Final As Run

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 4 Page 25 of 37 Event

Description:

Loss of Instrument Bus Time Position Applicants Actions or Behavior NOTE Upon loss of Instrument Bus A, ALL of the following instrumentation or equipment associated with the Core Heat Removal Safety Function are inoperable:

• "SUBCOOLED MARGIN MONITOR A-168"

• "RC LOOP TEMPERATURES LOOP 1A "T-COLD" A/TI-112C"

• "RC LOOP TEMPERATURES LOOP 1 "T-HOT" A/TI-112H"

• "RC LOOP TEMPERATURES LOOP 2A "T-COLD" A/TI-122C"

• "RC LOOP TEMPERATURES LOOP 2 "T-HOT" A/TI-122H"

• "SHTDN HT EXCH AC-4A OUTLET VALVE CNTRLR HCV-484" ATCO DETERMINE all RCPs are RUNNING. [Step 4.15]

NOTE Upon loss of Instrument Bus A, ALL of the following instrumentation or equipment associated with the RCS Heat Removal Safety Function is inoperable:

• "EMGY FW STOR TNK LEVEL LIA-1183"

• "STEAM GENERATOR PRESSURE S/G RC-2A A/PIC-902"

• "STEAM GENERATOR PRESSURE S/G RC-2B A/PIC-905"

• "AUX FW PUMP FW-6 SUCTION FLOW FIC-1368" BOPO DETERMIN Steam Generator NR levels steady at ~63%. [Step 4.16]

NOTE

• Upon loss of Instrument Bus A, RM-091A, which is associated with the Containment Integrity Safety Function is inoperable.

ATCO PERFORM the following to CONFIRM Containment Integrity: [Step 4.17]

• DETERMINE no unexpected rise in Containment Sump level.

[Step 4.17.a]

• DETERMINE no Containment Area Radiation Monitor alarms.

[Step 4.17.b]

• DETERMINE Radiation Monitors RM-051 / RM-052 / RM-062 NOT in alarm. [Step 4.17.c]

• DETERMINE SG Blowdown or Condenser off Gas Radiation Monitors RM-054A / RM-054B / RM-057 NOT in alarm. [Step 4.17.d]

• DETERMINE Containment conditions NORMAL. [Step 4.17.e]

• DETERMINE Containment pressure < 3 psig.

NRC Simulator Scenario 4 Outline Rev. Final As Run

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 4 Page 26 of 37 Event

Description:

Loss of Instrument Bus Time Position Applicants Actions or Behavior

• DETERMINE Containment temperature <120°F.

ATCO PLACE the following switches in TEST: [Step 4.18]

• HC-344/TEST, CNTMT SPRAY VLV HCV-344 TEST SWITCH

• HC-345/TEST, CNTMT SPRAY VLV HCV-345 TEST SWITCH NOTE Upon loss of Instrument Bus A, ALL of the following instrumentation or equipment associated with the Engineered Safety Features Systems is affected as follows:

• Safety Injection Tanks 6A and 6C level and pressure indicators are inoperable

• OPLS is in half-trip

• Sequencer S2-2 is inoperable

• PPLS is in a two-out-of-three logic mode

• SGLS is in a two-out-of-three logic mode CRS REFER to all the following Technical Specifications: [Step 4.19]

• 2.1.6, Pressurizer and Steam System Safety Valves

• 2.2, Chemical and Volume Control System

• 2.3, Emergency Core Cooling System

• 2.5, Steam and Feedwater Systems

• 2.7, Electrical Systems

• 2.15, Instrumentation and Control Systems

• 2.21, Post-Accident Monitoring Instrumentation CRS EVALUATE Technical Specification LCO 2.7, Electrical Systems

• LCO 2.7.(1).h - 120 VAC Instrument Bus A (Panel AI-40A).

• CONDITION 2.7.(2).h - 120 VAC Instrument Bus A (Panel AI-40A) inoperable

• ACTION 2.7.(2).h - May remain inoperable for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> provided RPS and ESF instrument channels supplied by the remaining 3 buses are all OPERABLE.

REFER to Electrical Load Distribution Listing Manual for a list of components CREW powered from AI-40A. [Step 4.20]

NRC Simulator Scenario 4 Outline Rev. Final As Run

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 4 Page 27 of 37 Event

Description:

Loss of Instrument Bus Time Position Applicants Actions or Behavior Examiner Note: Instrument Bus IA-40A will remain deenergized for duration of scenario.

Booth Operator: When contacted, REPORT Electrical Maintenance investigating issue with Inverter A.

When cause of power loss has been determined and corrected, RESTORE

+15 min CRS AI-40A to normal per Attachment 1 or 12 of OI-EE-4, 120 Volt AC System Normal Operation. [Step 4.21]

When Technical Specifications have been addressed, PROCEED to Events 5, 6, and 7.

(Alternate Path: If this event was initiated as Alternate Path Event 8, Terminate the Scenario).

NRC Simulator Scenario 4 Outline Rev. Final As Run

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 5, 6, & 7 Page 28 of 37 Event

Description:

RCP Trip / Automatic Reactor Trip Failure / Instrument Air Compressors Trip / Bearing Cooling Water Pump Trip / Steam Line Break inside Containment Time Position Applicants Actions or Behavior Booth Operator: When directed, EXECUTE Events 5, 6, and 7.

- Reactor Coolant Pump RC-3A trip.

- Instrument Air Compressors CA-1B and CA-1C trip.

- Bearing Cooling Water Pump AC-9B trip.

- Steam Line Break inside Containment on RC-2A @ 0.65% severity and 5 minute ramp.

Indications Available:

CB-1,2,3,4/A6 - REACTOR COOLANT PUMP RC-3A BREAKER O/L OR TRIP Low Flow Trip Unit lights lit on all RPS Channels B/C/D ERF Computer System alarms for low RCS flow

+30 sec ATCO RECOGNIZE RPS Low Flow lights lit and MANUALLY trip Reactor.

CRS DIRECT performance of EOP-00, Standard Post Trip Actions.

Examiner Note: The following steps are from EOP-00, Standard Post Trip Actions.

ATCO VERIFY Reactivity Control: [Step 1]

• VERIFY ALL of the following: [Step 1.a]

• DETERMINE more than one Regulating or Shutdown CEA NOT inserted.

• [CA] If Reactor did NOT trip, ESTABLISH Reactivity Control by performing step a, b, c or d: [Step 1.1]

Manually Trip Reactor to meet Core Design Criteria of Lowering Reactor Power CRITICAL TASK and Negative Startup Rate to Verify Reactivity Control Established During STATEMENT ATWS Event Prior to Exiting EOP-00, Standard Post Trip Actions.

CRITICAL TASK ATCO * [CA] Manually TRIP Reactor at CB-4. [Step 1.1.a]

• VERIFY Reactor Power is LOWERING.

• VERIFY Startup Rate is NEGATIVE.

NRC Simulator Scenario 4 Outline Rev. Final As Run

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 5, 6, & 7 Page 29 of 37 Event

Description:

RCP Trip / Automatic Reactor Trip Failure / Instrument Air Compressors Trip / Bearing Cooling Water Pump Trip / Steam Line Break inside Containment Time Position Applicants Actions or Behavior Examiner Note: An Emergency Boration is performed once the cooldown is recognized.

• DETERMINE an uncontrolled RCS Cooldown in progress. [Step 1.b]

• [CA] PERFORM Emergency Boration with uncontrolled cooldown in progress. [Step 1.2]

• [CA] ENSURE both following valves CLOSED: [Step 1.2.a]

• [CA] FCV-269X, Demin Water Makeup Valve

• [CA] FCV-269Y, Boric Acid Makeup Valve

• [CA] OPEN all the following valves: [Step 1.2.b]

• [CA] HCV-268, Boric Acid Pump Header to Charging Pumps Isolation Valve

• [CA] HCV-265, CH-11A Gravity Feed Valve

• [CA] HCV-258, CH-11B Gravity Feed Valve

• [CA] START all the following pumps: [Step 1.2.c]

• [CA] Boric Acid Pump CH-4A

• [CA] Boric Acid Pump CH-4B

• [CA] Charging Pump CH-1A (inoperable)

• [CA] Charging Pump CH-1B

• [CA] Charging Pump CH-1C

• [CA] CLOSE LCV-218-2, VCT Outlet Valve. [Step 1.2.d]

• [CA] ENSURE the following valves CLOSED: [Step 1.2.e]

• [CA] LCV-218-3, Charging Pump Suction SIRWT Isolation Valve

• [CA] HCV-257, CH-4B Recirc Valve

• [CA] HCV-264, CH-4A Recirc Valve

• [CA] BORATE until adequate Shutdown Margin is established.

[Step 1.2.f]

CRS DETERMINE Reactivity Control criteria SATISFIED.

BOPO VERIFY Turbine Trip: [Step 2]

• VERIFY HP & LP Stop and Intercept Valves CLOSED.

NRC Simulator Scenario 4 Outline Rev. Final As Run

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 5, 6, & 7 Page 30 of 37 Event

Description:

RCP Trip / Automatic Reactor Trip Failure / Instrument Air Compressors Trip / Bearing Cooling Water Pump Trip / Steam Line Break inside Containment Time Position Applicants Actions or Behavior Examiner Note: The Generator Output Breakers are CLOSED due to back feeding.

BOPO ENSURE all Generator Breakers are tripped: [Step 3]

• DETERMINE Generator Output Breaker 3451-4 CLOSED.

• DETERMINE Generator Output Breaker 3451-5 CLOSED.

• DETERMINE Generator Field Breaker 41E/G1F tripped.

BOPO VERIFY 4160 V Safeguards Buses 1A3 and 1A4 are ENERGIZED. [Step 4]

CRS DETERMINE Maintenance of Vital Auxiliaries criteria SATISFIED.

DETERMINE Safety Injection Actuation Signal has NOT occurred and both BOPO Diesel Generators are STOPPED. [Step 5]

VERIFY 4160 V Non-Safeguards Buses 1A1 and 1A2 are ENERGIZED.

BOPO

[Step 6]

BOPO VERIFY 125 VDC Buses 1 and 2 are ENERGIZED. [Step 7]

BOPO VERIFY Instrument Air is AVAILABLE: [Step 8]

• DETERMINE Instrument Air pressure < 90 psig.

• DETERMINE Instrument Air Compressors NOT RUNNING.

• [CA] If Instrument Air pressure is < 90 psig, PERFORM the following to restore Instrument Air: [Step 8.1]

BOPO * [CA] START Bearing Water Pump AC-9A.

BOPO * [CA] START Air Compressor CA-1A.

ATCO VERIFY Component Cooling Water System operation NORMAL: [Step 9]

• DETERMINE at least one CCW pump RUNNING. [Step 9.a]

• DETERMINE CCW Pump discharge pressure 60 psig. [Step 9.b]

NRC Simulator Scenario 4 Outline Rev. Final As Run

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 5, 6, & 7 Page 31 of 37 Event

Description:

RCP Trip / Automatic Reactor Trip Failure / Instrument Air Compressors Trip / Bearing Cooling Water Pump Trip / Steam Line Break inside Containment Time Position Applicants Actions or Behavior

• DETERMINE HCV-438A/B/C/D, CCW to RCP Coolers OPEN. [Step 9.c]

• DETERMINE at least one Raw Water Pump RUNNING. [Step 9.d]

CRS VERIFY RCS Inventory Control criteria satisfied: [Step 10]

• DETERMINE PZR level between 30% and 70% and NOT TRENDING to ATCO between 45% and 60%.

• [CA] RESTORE Inventory Control by manually controlling Charging and Letdown. [Step 10.1.a]

• DETERMINE RCS subcooling > 20°F.

CRS DETERMINE RCS Inventory Control criteria NOT SATISFIED.

CRS VERIFY RCS Pressure Control criteria satisfied: [Step 11]

ATCO

• DETERMINE RCS pressure between 1800 psia and 2300 psia.

• DETERMINE RCS pressure NOT TRENDING between 2050 psia and 2150 psia.

• [CA] MANUALLY CONTROL PZR Heaters and Spray to restore RCS pressure.

• DETERMINE PORVs are CLOSED.

CRS DETERMINE RCS Pressure Control criteria NOT SATISFIED.

CRS VERIFY Core Heat Removal criteria satisfied: [Step 12]

ATCO

• DETERMINE RCP NPSH requirements met.

• DETERMINE at least one RCP operating.

• DETERMINE Core T 10°F.

CRS DETERMINE Core Heat Removal criteria SATISFIED.

CRS VERIFY RCS Heat Removal criteria satisfied:

NRC Simulator Scenario 4 Outline Rev. Final As Run

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 5, 6, & 7 Page 32 of 37 Event

Description:

RCP Trip / Automatic Reactor Trip Failure / Instrument Air Compressors Trip / Bearing Cooling Water Pump Trip / Steam Line Break inside Containment Time Position Applicants Actions or Behavior VERIFY Main Feedwater is restoring SG levels to 35% to 80% NR and 73%

BOPO to 94% WR. [Step 13]

• DETERMINE FCV-1101 & FCV-1102 Feed Regulating Valves CLOSED.

[Step 13.a]

• DETERMINE HCV-1105 & HCV-1106 Feed Regulating Bypass Valves ramped to between 40% & 45% OPEN. [Step 13.b]

BOPO

• PLACE 43/FW switch in OFF. [Step 13.c]

• ENSURE no more than one Main Feedwater Pump RUNNING.

[Step 13.d]

• ENSURE no more than one Condensate Pump RUNNING. [Step 13.e]

• STOP running Heater Drain Pumps FW-5A, FW-5B, and/or FW-5C.

[Step 13.f]

BOPO

• ENSURE SG Blowdown Isolation Valves CLOSED. [Step 13.g]

• HCV-1387A & HCV-1387B

• HCV-1388A & HCV-1388B Booth Operator: When contacted, REPORT Air Assisted Main Steam Safety Valves CLOSED.

VERIFY Steam Dump and Bypass Valves controlling both of the following:

BOPO

[Step 14]

• DETERMINE RCS TCOLD NOT between 525°F and 535°F.

• DETERMINE Steam Generator RC-2A pressure < 700 psia.

• [CA] If TCOLD less than 525°F, PERFORM the following: [Step 14.1]

BOPO * [CA] CLOSE Steam Dump and Bypass Valves. [Step 14.1.a]

• [CA] VERIFY HCV-1040, Atmospheric Dump Valve CLOSED.

[Step 14.1.b]

• [CA] CHECK Air Assisted Main Steam Safety Valves CLOSED.

[Step 14.1.c]

• [CA] If Steam Generator pressure < 700 psia, ISOLATE Main Steam Header. [Step 14.1.d]

• [CA] CLOSE HCV-1041A, MSIV. [Step 14.1.d.1)]

• [CA] CLOSE HCV-1042A, MSIV. [Step 14.1.d.1)]

• [CA] VERIFY HCV-1041A, MSIV Bypass CLOSED.

[Step 14.1.d.2)]

NRC Simulator Scenario 4 Outline Rev. Final As Run

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 5, 6, & 7 Page 33 of 37 Event

Description:

RCP Trip / Automatic Reactor Trip Failure / Instrument Air Compressors Trip / Bearing Cooling Water Pump Trip / Steam Line Break inside Containment Time Position Applicants Actions or Behavior

• [CA] VERIFY CLOSE HCV-1042A, MSIV Bypass CLOSED.

[Step 14.1.d.2)]

• [CA] DETERMINE Steam Generator pressure > 500 psia.

[Step 14.1.e]

CRS DETERMINE RCS Heat Removal criteria NOT SATISFIED.

CRS VERIFY Containment Isolation criteria satisfied:

ATCO VERIFY Normal Containment conditions exist: [Step 15]

• DETERMINE rise in Containment Sump level in progress. [Step 15.a]

• DETERMINE Containment Area Radiation Monitors NOT in alarm.

[Step 15.b]

• DETERMINE Containment Ventilation Radiation Monitors NOT in alarm.

[Step 15.c]

• DETERMINE SG Blowdown and Condenser Off Gas Radiation Monitors ATCO NOT alarming. [Step 15.d]

• DETERMINE SG Blowdown and Condenser Off Gas Radiation Monitors ATCO NOT TRENDING to alarm. [Step 15.e]

ATCO

• VERIFY Containment conditions: [Step 15.f]

• DETERMINE Containment pressure > 3 psig.

• DETERMINE Containment temperature > 120°F.

• [CA] INITIATE Containment Cooling. [Step 15.f.1]

ATCO * [CA] ENSURE CCW flow to Containment Vent Fan coils.

• [CA] PLACE HCV-402B/D to OPEN.

• [CA] PLACE HCV-403B/D to OPEN.

• [CA] PLACE HCV-402A/C to OPEN.

• [CA] PLACE HCV-403A/C to OPEN.

ATCO * [CA] START all Containment Vent Fans.

• [CA] VERIFY Containment Vent Fans VA-3A & VA-3B RUNNING.

• [CA] START Containment Vent Fans VA-7C & VA-7D.

• [CA] DETERMINE Containment pressure < 5 psig. [Step 15.f.2]

NRC Simulator Scenario 4 Outline Rev. Final As Run

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 5, 6, & 7 Page 34 of 37 Event

Description:

RCP Trip / Automatic Reactor Trip Failure / Instrument Air Compressors Trip / Bearing Cooling Water Pump Trip / Steam Line Break inside Containment Time Position Applicants Actions or Behavior CRS DETERMINE Containment Integrity criteria NOT SATISFIED.

CRS DIAGNOSE event in progress: [Step 16]

• DETERMINE Reactivity Control requirements met.

• DETERMINE both DC buses energized.

• DETERMINE at least one Vital 4160 V Bus energized.

• DETERMINE at least one Non-Vital 4160 V Bus energized.

• DETERMINE at least one RCP running.

• DETERMINE adequate Feedwater flow to at least one SG.

• VERIFY Pressurizer pressure > 1800 psia with high subcooled margin, normal SG pressure, and no indications of primary to secondary leakage.

• If not, CONSIDER EOP-05, Uncontrolled Heat Extraction.

NOTE Certain events (i.e., LOCA, SGTR, UHE and Loss of All Feedwater) do not require offsite power in order to adequately, mitigate the effects of the accident.

For this reason, the LOCA, SGTR, UHE or Loss of All Feedwater procedure may be implemented even if a Loss of Offsite Power has also occurred.

• DETERMINE all Safety Function Acceptance Criteria NOT SATISFIED.

• DETERMINE single event in progress and TRANSITION to EOP-05, Uncontrolled Heat Extraction.

Examiner Note: The following steps are from EOP-05, Uncontrolled Heat Extraction.

CRS CONFIRM Standard Post Trip Actions have been performed. [Step 1]

CRS CONFIRM Uncontrolled Heat Extraction Diagnosis: [Step 2]

• VERIFY Safety Function Status Check Acceptance Criteria being satisfied. [Step 2.a]

• DETERMINE CIAS is NOT present and DIRECT Shift Chemist to SAMPLE both SGs for activity. [Step 2.c]

NRC Simulator Scenario 4 Outline Rev. Final As Run

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 5, 6, & 7 Page 35 of 37 Event

Description:

RCP Trip / Automatic Reactor Trip Failure / Instrument Air Compressors Trip / Bearing Cooling Water Pump Trip / Steam Line Break inside Containment Time Position Applicants Actions or Behavior CRS IMPLEMENT the Emergency Plan. [Step 3]

• Time: __________

CREW MONITOR the Floating Steps. [Step 4]

DETERMINE RCS pressure > 1600 psia, Containment pressure < 5 psig, CRS with Steam Generator > 500 psia. [Step 5]

CRS DETERMINE RCS pressure 1600 psia. [Step 6]

CRS DETERMINE Containment pressure < 5 psig. [Step 7]

CRS DETERMINE SIAS has NOT actuated. [Step 8]

ATCO VERIFY RCP operating parameters: [Step 9]

• DETERMINE RCP RC-3A TRIPPED and TCOLD < 500°F. [Step 9.a]

• DETERMINE RCS pressure ~1900 psia. [Step 9.b]

• DETERMINE RCPs subcooling > 20°F. [Step 9.c]

ATCO VERIFY normal CCW/RW System operation: [Step 10]

• DETERMINE at least 2 CCW Pumps are RUNNING. [Step 10.a]

• DETERMINE CCW Pump discharge pressure 60 psig. [Step 10.b]

• ENSURE at least two Raw Water Pumps operating. [Step 10.c]

ATCO

• START at least one Raw Water Pump.

• DETERMINE at least three RW/CCW Heat Exchangers in service.

[Step 10.d]

• DETERMINE all RCP cooler CCW Valves OPEN. [Step 10.e]

CRS DETERMINE affected SG is RC-2A and SG pressure is < 700 psia. [Step 11]

CRS DETERMINE Uncontrolled Heat Extraction has NOT been isolated. [Step 12]

NRC Simulator Scenario 4 Outline Rev. Final As Run

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 5, 6, & 7 Page 36 of 37 Event

Description:

RCP Trip / Automatic Reactor Trip Failure / Instrument Air Compressors Trip / Bearing Cooling Water Pump Trip / Steam Line Break inside Containment Time Position Applicants Actions or Behavior

• [CA] DETERMINE Emergency Boration already in progress. [Step 12.1]

BOPO DETERMINE SG RC-2A and SG RC-2B both > 500 psia. [Step 13]

BOPO DETERMINE Steam Generator RC-2A is most affected SG. [Step 14]

CRS DETERMINE Uncontrolled Heat Extraction has NOT been isolated. [Step 15]

IF RC-2A is most affected, ISOLATE RC-2A by performing HR-19, CRS Isolate/Restore Steam Generator A. [Step 16]

Examiner Note: The following steps are from HR-19, Isolate/Restore Steam Generator A.

CRITICAL TASK Isolate the Affected Steam Generator to Prevent Excess Plant Cooldown and STATEMENT Reactivity Additions Prior to Steam Generator Level = 0% Wide Range Level.

CRITICAL TASK BOPO PERFORM the following to isolate Steam Generator RC-2A: [Step 1]

• ENSURE all the following valves are CLOSED: [Step 1.a]

• VERIFY HCV-1041A, RC-2A MSIV CLOSED.

• VERIFY HCV-1041C, RC-2A MSIV Bypass Valve CLOSED.

• VERIFY MS-291, Air Assisted Main Steam Safety Valve CLOSED.

• VERIFY FCV-1101, RC-2A Feed Regulating Valve CLOSED.

• VERIFY HCV-1105, Feed Regulating Bypass Valve CLOSED.

BOPO

• VERIFY HCV-1386, RC-2A Feed Header Isolation Valve CLOSED.

• VERIFY HCV-1103, Feed Regulating Block Valve CLOSED.

• VERIFY HCV-1388A, Blowdown Isolation Valve CLOSED.

• VERIFY HCV-1388B, Blowdown Isolation Valve CLOSED.

• CLOSE HCV-1107A, AFW Isolation Valve.

• CLOSE HCV-1107B, AFW Isolation Valve.

• CONTACT Auxiliary Operator to CLOSE MS-298, Steam Valves HCV-1041A & 1042A Packing Leakoff Line Isolation Valve in Room 81.

[Step 1.b]

• If sampling is NOT in progress, CLOSE both Sample Valves: [Step 1.c]

• HCV-2506A, RC-2A Blowdown Sample Isolation Valve

• HCV-2506B, RC-2A Blowdown Sample Isolation Valve NRC Simulator Scenario 4 Outline Rev. Final As Run

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 5, 6, & 7 Page 37 of 37 Event

Description:

RCP Trip / Automatic Reactor Trip Failure / Instrument Air Compressors Trip / Bearing Cooling Water Pump Trip / Steam Line Break inside Containment Time Position Applicants Actions or Behavior

• PERFORM the following to CLOSE YCV-1045A: [Step 1.d]

• PLACE ISOLATION VALVE YCV-1045A OVERRIDE SW in BOPO OVERRIDE. [Step 1.d.1)]

• PLACE control switch for S/G RC-2A STM TO FW-10 HDR A BOPO ISOLATION VALVE YCV-1045A in CLOSE. [Step 1.d.2)]

NOTE Air accumulators will maintain the valve in a closed position for 30 minutes after a loss of Instrument Air.

• CONTACT Auxiliary Operator to HANDJACK YCV-1045A, MAIN STEAM LINE "A" TO AUX FEEDWATER PUMP FW-10 SUPPLY VALVE to CLOSE in Room 81. [Step 1.d.3)]

VERIFY RC-2A is most affected SG per Attachment HR-18, Most Affected CRS Steam Generator Determination. [Step 2]

Examiner Note: If a Reactor Trip was performed due to high CCW or RCP bearing temperatures and RC-2A is isolated, INITIATE Event 4 (Alternate Path Event 8), Loss of Instrument Bus AI-40A, and REFER to Page 21 of 37.

When Steam Generator RC-2A is isolated, TERMINATE the scenario.

NRC Simulator Scenario 4 Outline Rev. Final As Run

ES-301 Transient and Event Checklist Form ES-301-5 Facility: Fort Calhoun Station Date of Exam: 12/07/15 Operating Test No.: NRC A E SCENARIOS P V P E FCS #1 FCS #3 FCS #4 L N T MINIMUM(*)

I T CREW CREW CREW CREW O C

POSITION POSITION POSITION POSITION T A T A

N Y S A B S A B S A B S A B L

T P R T O R T O R T O R T O R I U E O C P O C P O C P O C P RX - - 0 1 1 0 NOR - 1 1 1 1 1 SRO-U1 I/C 2,3,4, 3,6 6 4 4 2 5

MAJ 6 5,7 3 2 2 1 TS 2,4 - 2 0 2 2 RX - - 0 1 1 0 NOR - 1 1 1 1 1 SRO-U2 I/C 2,3,4, 2,4 6 4 4 2 5

MAJ 6 7 2 2 2 1 TS 2,4 2,4 4 0 2 2 RX - 6 - 1 1 1 0 NOR - - 1 1 1 1 1 SRO-I1 I/C 2,3,4, 2,4,5 6 8 4 4 2 5

MAJ 6 7 7 3 2 2 1 TS 2,4 - - 2 0 2 2 RX - - - 0 1 1 0 NOR 1 6 1 3 1 1 1 SRO-I2 I/C 2,4,8, 3,4,8 2,4 9 4 4 2 9

MAJ 6 7 7 3 2 2 1 TS - - 2,4 2 0 2 2 RX - - 1 1 1 1 0 NOR - 6 - 1 1 1 1 SRO-I3 I/C 3,5,7 2,3,4, 2,4 9 4 4 2 5

MAJ 6 7 7 3 2 2 1 TS - 4,5 - 2 0 2 2 RX - - 0 1 1 0 NOR 1 1 2 1 1 1 SRO-I4 I/C 2,4,8, 2,3,4 7 4 4 2 9

MAJ 6 5,7 3 2 2 1 TS - 2,3,4 3 0 2 2 FCS 2015 NRC ES-301-5 Transient and Event Checklist Final As Run

ES-301 Transient and Event Checklist Form ES-301-5 Facility: Fort Calhoun Station Date of Exam: 12/07/15 Operating Test No.: NRC A E SCENARIOS P V P E FCS #1 FCS #3 FCS #4 L N T MINIMUM(*)

I T CREW CREW CREW CREW O C

POSITION POSITION POSITION POSITION T A T A

N Y S A B S A B S A B S A B R T O R T O R T O R T O L T P R I U E O C P O C P O C P O C P RX - 1 1 1 1 0 NOR - - 0 1 1 1 RO-1 I/C 3,5,7 2,4 5 4 4 2 MAJ 6 7 2 2 2 1 TS - - 0 0 2 2 RX - 1 1 1 1 0 NOR - - 0 1 1 1 RO-2 I/C 3,5,7 2,4 5 4 4 2 MAJ 6 5,7 3 2 2 1 TS - - 0 0 2 2 RX - - 0 1 1 0 NOR 1 1 2 1 1 1 RO-3 I/C 2,4,8, 6 5 4 4 2 9

MAJ 6 7 2 2 2 1 TS - - 0 0 2 2 Instructions:

1. Check the applicant level and enter the operating test number and Form ES-D-1 event numbers for each event type; TS are not applicable for RO applicants. ROs must serve in both the at-the-controls (ATC) and balance-of-plant (BOP) positions; Instant SROs (SRO-I) must serve in both the SRO and the ATC positions, including at least two instrument or component (I/C) malfunctions and one major transient, in the ATC position. If an SRO-1 additionally serves in the BOP position, one I/C malfunction can be credited toward the two I/C malfunctions required for the ATC position.

2. Reactivity manipulations may be conducted under normal or controlled abnormal conditions (refer to Section D.5.d) but must be significant per Section C.2.a of Appendix D. (*) Reactivity and normal evolutions may be replaced with additional instrument or component malfunctions on a 1-for-1 basis.

3. Whenever practical, both instrument and component malfunctions should be included; only those that require verifiable actions that provide insight to the applicants competence count toward the minimum requirements specified for the applicants license level in the right-hand columns.

4. For licensees that use the ATC operator primarily for monitoring plant parameters, the chief examiner may place SRO-I applicants in either the ATC or BOP position to best evaluate the SRO-I in manipulating plant controls.

FCS 2015 NRC ES-301-5 Transient and Event Checklist Final As Run

ES-301 Competencies Checklist Form ES-301-6 Facility: FCS Date of Examination: 12/07/15 Operating Test No. NRC 1/3/4 Applicants SROU-1 SROU-2 SROI-1 Competencies SCENARIO SCENARIO SCENARIO 1 3 4 1 3 4 1 3 4 Interpret/Diag-2,3,4,5, 3,5,6, 2,3,4,5, 2,3,4,5, 2,4,5, nose Events 6 7 6

- 2,4 6 7 6,7 and Conditions Comply With 2,3,4,5, 1,3,5, 2,3,4,5, 1,2,4, 2,3,4,5, 2,4,5, and Use 6 6,7 6 7 6 6,7 1,6,7 Procedures (1)

Operate 1,3,5, 2,4,5, Control Boards N/A -

6,7 N/A - N/A N/A 6,7 1,6,7 (2)

Communicate 1,2,3,4, 1,3,5, 1,2,3,4, 1,2,4, 1,2,3,4, 2,4,5, and 5,6 6,7 5,6 7 5,6 6,7 1,6,7 Interact Demonstrate 2,3,4,5, 2,3,4,5, 1,2,4, 2,3,4,5, Supervisory 6

- N/A 6

7 6 N/A N/A Ability (3)

Comply With and Use Tech. 2,4 - N/A 2,4 - 2,4 2,4 N/A N/A Specs. (3)

Notes:

(1) Includes Technical Specification compliance for an RO.

(2) Optional for an SRO-U.

(3) Only applicable to SROs.

FCS 2015 NRC ES-301-6 Competencies Checklist Final As Run

ES-301 Competencies Checklist Form ES-301-6 Facility: FCS Date of Examination: 12/07/15 Operating Test No. NRC 1/3/4 Applicants SROI-2 SROI-3 SROI-4 Competencies SCENARIO SCENARIO SCENARIO 1 3 4 1 3 4 1 3 4 Interpret/Diag-2,4,5,6, 3,5,7, 2,3,4, 2,4,5,6, nose Events 8,9 8 2,4 3,5,6,7 5,7 2,4,7 8,9

- 2,3,4,5 and Conditions Comply With 1,2,4,5, 1,3,5, 1,2,4, 2,3,4, 1,2,4, 1,2,4,5, 1,2,3,4, and Use 6,8,9 6,7,8 7 3,5,6,7 5,6,7 7 6,8,9 5,7 Procedures (1)

Operate 1,2,4,5, 1,3,5, 1,2,4, 1,2,4,5, Control Boards 6,8,9 6,7,8 N/A 3,5,6,7 N/A 7 6,8,9

- N/A (2)

Communicate 1,2,3,4, 1,3,5, 1,2,4, 1,2,3,4, 1,2,4, 1,2,3,4, 1,2,3,4, and 5,6,8,9 6,7,8 7 3,5,6,7 5,6,7 7 5,6,8,9 5,7 Interact Demonstrate 1,2,4, 1,2,3,4, 1,2,3,4, Supervisory N/A N/A 7

N/A 5,6,7 N/A N/A -

5,7 Ability (3)

Comply With and Use Tech. N/A N/A 2,4 N/A 4,5 N/A N/A - 2,3,4 Specs. (3)

Notes:

(1) Includes Technical Specification compliance for an RO.

(2) Optional for an SRO-U.

(3) Only applicable to SROs.

FCS 2015 NRC ES-301-6 Competencies Checklist Final As Run

ES-301 Competencies Checklist Form ES-301-6 Facility: FCS Date of Examination: 12/07/15 Operating Test No. NRC 1/3/4 Applicants RO-1 RO-2 RO-3 Competencies SCENARIO SCENARIO SCENARIO 1 3 4 1 3 4 1 3 4 Interpret/Diag-2,4,5,6, nose Events 3,5,6,7 - 2,4,7 3,5,6,7 - 2,4,5,7 8,9

- 6,7 and Conditions Comply With 1,2,4, 1,2,4, 1,2,4,5, and Use 3,5,6,7 -

7 3,5,6,7 -

5,7 6,8,9

- 1,6,7 Procedures (1)

Operate 1,2,4, 1,2,4, 1,2,4,5, Control Boards 3,5,6,7 -

7 3,5,6,7 -

5,7 6,8,9

- 1,6,7 (2)

Communicate 1,2,4, 1,2,4, 1,2,3,4, and 3,5,6,7 -

7 3,5,6,7 -

5,7 5,6,8,9

- 1,6,7 Interact Demonstrate Supervisory N/A - N/A N/A - N/A N/A - N/A Ability (3)

Comply With and Use Tech. N/A - N/A N/A - N/A N/A - N/A Specs. (3)

Notes:

(1) Includes Technical Specification compliance for an RO.

(2) Optional for an SRO-U.

(3) Only applicable to SROs.

FCS 2015 NRC ES-301-6 Competencies Checklist Final As Run

Appendix C JPM WORKSHEET Form ES-C-1 Facility: FCS JPM # NRC RA1 Task # 1361 K/A # 2.1.25 3.9 / 4.2

Title:

Perform a Time to Boil Determination Examinee (Print):

Testing Method:

Simulated Performance: Classroom: X Actual Performance: X Simulator:

Alternate Path: Plant:

Time Critical:

READ TO THE EXAMINEE I will explain the Initial Conditions, which steps to simulate or discuss, and provide an Initiating Cue.

When you complete the task successfully, the objective for this JPM will be satisfied.

Initial Conditions: Given the following conditions:

• The plant was shut down 5 days ago for a Reactor Coolant Pump seal repair.

• AOP-19, Loss of Shutdown Cooling, was just entered at 0800.

• Refer to the Attached ERFCS printout for page 195, Shutdown Status Board, for current plant conditions.

Initiating Cue: The Control Room Supervisor directs you to PERFORM the following:

• PERFORM a Time to Boil Determination per AOP-19, Loss of Shutdown Cooling, Attachment B, Time to Boil Determination Worksheet.

• RECORD required information on Attachment B, Time to Boil Determination Worksheet wherever appears.

• Instrument numbers will be looked up by another operator.

Task Standard: Utilizing AOP-19, located RCS at Mid Loop graph, recorded appropriate Time to Boil data, and determined Time to Boil at 18 +/- 1 minutes.

Required Materials: AOP-19, Loss of Shutdown Cooling, Rev. 18.

Validation Time: 7 minutes Completion Time: ________ minutes Comments:

Result: SAT UNSAT Examiner (Print / Sign): Date:

Page 1 of 5 NRC Admin JPM RA1 Rev. Final

Appendix C JPM WORKSHEET Form ES-C-1 CLASSROOM SETUP EXAMINER:

PROVIDE the examinee with a copy of:

• AOP-19, Loss of Shutdown Cooling.

• Attachment B, Time to Boil Determination Worksheet.

• Attachment C, Time to Boil Determination.

Page 2 of 5 NRC Admin JPM RA1 Rev. Final

Appendix C JPM STEPS Form ES-C-1

- Check Mark Denotes Critical Step START TIME:

Examiner Note: The following steps are from AOP-19, Attachment B.

Perform Step: 1 Time Shutdown Cooling was lost: _____

1 Standard: RECORDED time Shutdown Cooling was lost as 0800 on Attachment B.

Comment: SAT UNSAT Perform Step: 2 Last known RCS/SDCS temperature: _____ °F from instrument number:

2 _____

Standard: DETERMINED representative RCS temperature should be recorded for Core Exit Thermocouples, and RECORDED last known and HIGHEST RCS/SDCS temperature of 110°F from CETs on Attachment B.

Comment: SAT UNSAT Perform Step: 3 Record the following information and inform the Shift Manager on 10 3 minute intervals.

Standard: DETERMINED from ERFCS printout that the RCS is at MID LOOP, and referred to Mid Loop graph from Attachment B, and RECORDED 18 +/- 1 minutes on Attachment B.

Terminating Cue: This JPM is complete.

Comment: SAT UNSAT STOP TIME:

Page 3 of 5 NRC Admin JPM RA1 Rev. Final

Appendix C JPM CUE SHEET Form ES-C-1 INITIAL CONDITIONS: Given the following conditions:

• The plant was shut down 5 days ago for a Reactor Coolant Pump seal repair.

• AOP-19, Loss of Shutdown Cooling, was just entered at 0800.

• Refer to the Attached ERFCS printout for page 195, Shutdown Status Board, for current plant conditions.

INITIATING CUE: The Control Room Supervisor directs you to PERFORM the following:

• PERFORM a Time to Boil Determination per AOP-19, Loss of Shutdown Cooling, Attachment B, Time to Boil Determination Worksheet.

• RECORD required information on Attachment B, Time to Boil Determination Worksheet wherever appears.

• Instrument numbers will be looked up by another operator.

Page 4 of 5 NRC Admin JPM RA1 Rev. Final

Appendix C JPM CUE SHEET Form ES-C-1 Attachment A - ERFCS Page 195, Shutdown Status Board Page 5 of 5 NRC Admin JPM RA1 Rev. Final

Appendix C JPM WORKSHEET Form ES-C-1 Facility: FCS JPM # NRC RA2 Task # 1528 K/A # 2.1.43 4.1 / 4.3

Title:

Calculate an Estimated Critical Boron Concentration Examinee (Print):

Testing Method:

Simulated Performance: Classroom: X Actual Performance: X Simulator:

Alternate Path: Plant:

Time Critical:

READ TO THE EXAMINEE I will explain the Initial Conditions, which steps to simulate or discuss, and provide an Initiating Cue.

When you complete the task successfully, the objective for this JPM will be satisfied.

Initial Conditions: Given the following conditions:

• A Reactor Trip from 100% power occurred on 12/03/15 at 0400 following 6 months of full power operation.

• All Control Rods were fully withdrawn at the time of the trip.

• Boron concentration prior to the trip was 400 ppm.

• Average Core Burnup is 6 GWD/MTU.

• Current boron concentration is 500 ppm.

• Criticality is scheduled to occur with Regulating Group 4 at 78 inches.

• Reactor Engineering reports no correction is needed for boron depletion.

• Reactor Startup is scheduled for 12/10/15 at 0600.

Initiating Cue: The Control Room Supervisor directs you to PERFORM the following:

• CALCULATE estimated critical boron concentration per TDB-V-1-B, Estimated Critical Conditions Worksheet.

• COMPLETE data entry through TDB-V.1.B, Step E.3.d, Estimated Critical Boron Concentration.

Task Standard: Utilizing TDB-V.1.B and TDB-II, calculated Estimated Critical Boron Concentration.

Required Materials: TDB-V.1.B, Estimated Critical Conditions Worksheet, Rev. 26.

TDB-II, Technical Data Book Reactivity Curves, Rev. 35.

Validation Time: 40 minutes Completion Time: ________ minutes Comments:

Result: SAT UNSAT Examiner (Print / Sign): Date:

Page 1 of 5 NRC Admin JPM RA2 Rev. Final

Appendix C JPM WORKSHEET Form ES-C-1 CLASSROOM SETUP EXAMINER:

PROVIDE the examinee with a copy of:

• TDB-V.1.B, Estimated Critical Conditions Worksheet.

• TDB-II, Technical Data Book Reactivity Curves.

• Calculator

• Straight Edge Page 2 of 5 NRC Admin JPM RA2 Rev. Final

Appendix C JPM STEPS Form ES-C-1

- Check Mark Denotes Critical Step START TIME:

Examiner Note: The following steps are from TDB-V.1.B Perform Step: 1 Conditions at Time of Shutdown.

A Standard: ENTERED Conditions at Time of Shutdown in TDB-V-1-B Steps A.1 to A.5.

Examiner Note: Information found on Answer Key.

Comment: SAT UNSAT Perform Step: 2 Conditions at Time of Startup.

B Standard: ENTERED Conditions at Time of Startup in TDB-V.1.B Steps B.1 to B.4.

Examiner Note: Information found on Answer Key.

Comment: SAT UNSAT Perform Step: 3 ECC Applicability.

C Standard: DETERMINED early and late date/time limits for ECC Applicability and entered data in TDB-V.1.B Steps C.1 to C.4.

Examiner Note: Information found on Answer Key.

Comment: SAT UNSAT Perform Step: 4 Reactivity Changes Due To Shutdown.

D Standard: CALCULATED and ENTERED Reactivity Changes Due To Shutdown in TDB-V.1.B Steps D.1 to D.5.

Examiner Note: Information found on Answer Key.

Comment: SAT UNSAT Page 3 of 5 NRC Admin JPM RA2 Rev. Final

Appendix C JPM STEPS Form ES-C-1 Perform Step: 5 Estimated Critical Boron Concentration.

E Standard: CALCULATED and ENTERED Estimated Critical Boron Concentration in TDB-V.1.B Steps E.1 to E.3.

Step E.3.d: Calculated 884 +/- 25 ppm (Critical)

Examiner Note: Information found on Answer Key.

Terminating Cue: This JPM is complete.

Comment: SAT UNSAT STOP TIME:

Page 4 of 5 NRC Admin JPM RA2 Rev. Final

Appendix C JPM CUE SHEET Form ES-C-1 INITIAL CONDITIONS: Given the following conditions:

• A Reactor Trip from 100% power occurred on 12/03/15 at 0400 following 6 months of full power operation.

• All Control Rods were fully withdrawn at the time of the trip.

• Boron concentration prior to the trip was 400 ppm.

• Average Core Burnup is 6 GMWD/MTU.

• Current boron concentration is 500 ppm.

• Criticality is scheduled to occur with Regulating Group 4 at 78 inches.

• Reactor Engineering reports no correction is needed for boron depletion.

• Reactor Startup is scheduled for 12/10/15 at 0600.

INITIATING CUE: The Control Room Supervisor directs you to PERFORM the following:

• CALCULATE estimated critical boron concentration per TDB-V.1.B, Estimated Critical Conditions Worksheet.

• COMPLETE data entry through TDB-V.1.B, Step E.3.d, Estimated Critical Boron Concentration.

Page 5 of 5 NRC Admin JPM RA2 Rev. Final

Appendix C JPM WORKSHEET Form ES-C-1 Facility: FCS JPM # NRC RA3 Task # 0066 K/A # 2.2.35 3.6 / 4.5

Title:

Determine Technical Specification MODE of Operation Examinee (Print):

Testing Method:

Simulated Performance: Classroom: X Actual Performance: X Simulator:

Alternate Path: Plant:

Time Critical:

READ TO THE EXAMINEE I will explain the Initial Conditions, which steps to simulate or discuss, and provide an Initiating Cue.

When you complete the task successfully, the objective for this JPM will be satisfied.

Initial Conditions: Given the following conditions:

• Reactor Coolant System temperature is 200°F.

• Reactor Coolant System pressure is 150 psia.

• Reactor Coolant System boron concentration is 2180 ppm.

• Core Burnup is 1500 MWD/MTU.

• A Shutdown Cooling Pump is in operation.

Initiating Cue: The Control Room Supervisor directs you to PERFORM the following:

• DETERMINE Fort Calhoun Station Technical Specification Reactor Operating Condition.

• Refueling Boron Concentration _____ ppm.

• Operating Mode _____.

Task Standard: Utilizing Technical Specifications and Core Operating Limits Report, determined Fort Calhoun Station is in Operating Mode 5, Refueling Shutdown Condition.

Required Materials: Fort Calhoun Station Technical Specifications, Amendment #283.

TDB-VI, Core Operating Limits Report, Rev. 42.

Validation Time: 5 minutes Completion Time: ________ minutes Comments:

Result: SAT UNSAT Examiner (Print / Sign): Date:

Page 1 of 4 NRC Admin JPM RA3 Rev. Final

Appendix C JPM WORKSHEET Form ES-C-1 CLASSROOM SETUP EXAMINER:

PROVIDE the examinee with a copy of:

• Fort Calhoun Station Technical Specifications.

• TDB-VI, Core Operating Limits Report.

Page 2 of 4 NRC Admin JPM RA3 Rev. Final

Appendix C JPM STEPS Form ES-C-1

- Check Mark Denotes Critical Step START TIME:

Examiner Note: The following steps are from Technical Specifications and the Core Operating Limits Report.

Perform Step: 1 Refer to Technical Specifications for MODE definition.

Standard: REFERRED to Technical Specification Definitions, Page 2 and DETERMINED that plant is either in MODE 4 or 5 depending on boron concentration.

Comment: SAT UNSAT Perform Step: 2 Refer to Core Operating Limits Report to determine REFUELING BORON CONCENTRATION.

Standard: REFERRED to Core Operating Limits Report and DETERMINED REFUELING BORON CONCENTRATION at 1500 MWD/MTU is 2160 ppm.

Comment: SAT UNSAT Perform Step: 3 Determine Plant Operational Mode based on Reactor Coolant System Boron Concentration.

Standard: REFERRED to Technical Specification Definitions, Page 2 and DETERMINED that Plant is in Operating Mode 5, Refueling Shutdown Condition based on Reactor Coolant System Boron Concentration greater than REFUELING BORON CONCENTRATION.

Terminating Cue: This JPM is complete.

Comment: SAT UNSAT STOP TIME:

Page 3 of 4 NRC Admin JPM RA3 Rev. Final

Appendix C JPM CUE SHEET Form ES-C-1 INITIAL CONDITIONS: Given the following conditions:

• Reactor Coolant System temperature is 200°F.

• Reactor Coolant System pressure is 150 psia.

• Reactor Coolant System boron concentration is 2180 ppm.

• Core Burnup is 1500 MWD/MTU.

• A Shutdown Cooling Pump is in operation.

INITIATING CUE: The Control Room Supervisor directs you to PERFORM the following:

• DETERMINE Fort Calhoun Station Technical Specification Reactor Operating Condition.

• Refueling Boron Concentration _____ ppm.

• Operating Mode _____.

Page 4 of 4 NRC Admin JPM RA3 Rev. Final

Appendix C JPM WORKSHEET Form ES-C-1 Facility: FCS JPM # NRC RA4 Task # 1269 K/A # 2.3.11 3.8 / 4.3

Title:

Respond to Voids in the Reactor Coolant System Examinee (Print):

Testing Method:

Simulated Performance: Classroom: X Actual Performance: X Simulator:

Alternate Path: Plant:

Time Critical:

READ TO THE EXAMINEE I will explain the Initial Conditions, which steps to simulate or discuss, and provide an Initiating Cue.

When you complete the task successfully, the objective for this JPM will be satisfied.

Initial Conditions: Given the following conditions:

• A Small Break Loss of Coolant Accident (LOCA) has occurred.

• EOP-03, Loss of Coolant Accident, has been implemented.

• Reactor Coolant System (RCS) conditions:

• RCS Pressure = 450 psia.

• RCS TCOLD = 402°F.

• Pressurizer (PZR) conditions:

• PZR Level [actual] = 60% and stable.

• PZR Temperature = 456°F and stable.

• Reactor Vessel Level Monitoring System (RVLMS) is 83% and stable.

• Containment conditions:

• Containment Safety Function is satisfied.

• All Containment Ventilation Fans are operating.

• Containment Pressure = 1.2 psig.

• Containment Temperature = 118°F.

• Containment Hydrogen concentration = 1.2%.

• RC-5, Pressurizer Quench Tank (PZR QT), conditions:

• PZR QT Level = 70%.

• PZR QT Pressure = 5 psig.

• HPSI Stop and Throttle has been performed for a LOCA.

• RCS and Pressurizer sample results are normal.

• The Technical Support Center (TSC) is activated.

• Use of EOP/AOP Attachment IC-14, RCS Void Elimination, has been unsuccessful in eliminating the RCS voids.

Initiating Cue: The Control Room Supervisor directs you to PERFORM the following:

• IMPLEMENT OI-RC-12, Post Accident Venting of Noncondensable Gases from the Reactor Coolant System.

• VENT from (CIRCLE one): Reactor Vessel Head PZR

• VENT to (CIRCLE one): Containment PZR Quench Tank Page 1 of 9 NRC Admin JPM RA4 Rev. Final

Appendix C JPM WORKSHEET Form ES-C-1 Task Standard: Utilizing OI-RC-12, determined vent path from the Reactor Vessel Head to the Pressurizer Quench Tank is required.

Required Materials: OI-RC-12, Post Accident Venting of Noncondensable Gases from the Reactor Coolant System, Rev. 11.

Validation Time: 12 minutes Completion Time: ________ minutes Comments:

Result: SAT UNSAT Examiner (Print / Sign): Date:

Page 2 of 9 NRC Admin JPM RA4 Rev. Final

Appendix C JPM WORKSHEET Form ES-C-1 CLASSROOM SETUP EXAMINER:

PROVIDE the examinee with a copy of:

• OI-RC-12, Post Accident Venting of Noncondensable Gases from the Reactor Coolant System.

• Steam Tables Page 3 of 9 NRC Admin JPM RA4 Rev. Final

Appendix C JPM CUE SHEET Form ES-C-1

- Check Mark Denotes Critical Step START TIME:

Examiner Note: The following step is from OI-RC-12, Attachment 1, Prerequisites.

Perform Step: 1 PREREQUISITES:

1, 2, & 3

• Procedure Revision Verification Revision No. _____ Date: _____

• The Reactor is subcritical with a Tave less than 515°F (Ref.

Technical Specification 2.1.8).

• The RCS is being maintained in a stable condition with the following:

• Pressurizer (PZR) Level is between 49% and 93%

• Charging flow is in operation

• RCS subcooling is between 20°F and 200°F Standard: DETERMINED the following per the Initial Conditions:

• Procedure Revision is as provided.

• Reactor Coolant System TAVE is ~402°F.

• Pressurizer Level is 60%.

• Charging flow is in operation (based on HPSI Stop and Throttle has been performed).

• RCS subcooling is ~54°F.

Comment: SAT UNSAT Page 4 of 9 NRC Admin JPM RA4 Rev. Final

Appendix C JPM CUE SHEET Form ES-C-1 Examiner Note: The following steps are from OI-RC-12, Attachment 1, Procedure.

Perform Step: 2 IF one or more of the following conditions are present in the PZR, THEN 1 & all bullets determine the venting path per Attachment 2:

• Figure 1 indicates the presence of non-condensable gases

• Departure from saturation

• Sluggish pressure control

• Sampling results indicate non-condensable gases Standard: IDENTIFIED that the Pressurizer does not indicate a departure from saturated conditions and no other conditions warrant Pressurizer venting.

Comment: SAT UNSAT Perform Step: 3 Determine if bubble exists in the RV Head by monitoring RV level less 2 than 100% via the Reactor Vessel Level Monitoring System (RVLMS),

THEN determine the venting path per Attachment 2.

Standard: IDENTIFIED that a bubble exists in the Reactor Vessel Head and CIRCLED Vent from: Reactor Vessel Head Comment: SAT UNSAT Page 5 of 9 NRC Admin JPM RA4 Rev. Final

Appendix C JPM CUE SHEET Form ES-C-1 Examiner Note: The following step is from OI-RC-12, Attachment 2, Prerequisites.

Perform Step: 4 PREREQUISITES:

1, 2, 3, 4, 5, & 6

• Procedure Revision Verification Revision No. _____ Date: _____

• The reactor is subcritical with a Tave less than 515°F (Ref.

Technical Specification 2.1.8).

• Containment Isolation has been verified per EOP Safety Function Status Check.

• All available Containment Ventilation Units are in operation:

• VA-3A, Cntmt Vent Fan

• VA-3B, Cntmt Vent Fan

• VA-7C, Cntmt Vent Fan

• VA-7D, Cntmt Vent Fan

• The RCS is being maintained in a stable condition with the following:

• Pressurizer (PZR) Level is between 49% and 93%

• Charging flow is in operation

• RCS subcooling is between 20°F and 200°F

• TSC has been activated.

Standard: DETERMINED the following per the Initial Conditions:

• Procedure Revision is as provided.

• Reactor Coolant System TAVE is ~402°F.

• Containment Isolation has been verified.

• All Containment Ventilation Units are in operation.

• Pressurizer Level is 60%.

• Charging flow is in operation (based on HPSI Stop and Throttle has been performed).

• RCS subcooling is ~54°F.

• Technical Support Center has been activated.

Comment: SAT UNSAT Page 6 of 9 NRC Admin JPM RA4 Rev. Final

Appendix C JPM CUE SHEET Form ES-C-1 Examiner Note: The following steps are from OI-RC-12, Attachment 2, Procedure.

Perform Step: 5 IF one or more of the following conditions exist, THEN the Containment 1, 1.a, & all bullets vent path should be used per the following:

• Verify:

• There is no water in the RC-5, PQT AND DC Bus 1 electrical power source is available to RCGVS Valves

• Large quantities of gas need to be vented

• Rapid venting is required

• The potential for loss of core cooling exists

• There is serious interference with the ability to maintain pressure control Standard: IDENTIFIED that the Containment vent path is NOT preferred.

Comment: SAT UNSAT Page 7 of 9 NRC Admin JPM RA4 Rev. Final

Appendix C JPM CUE SHEET Form ES-C-1 Perform Step: 6 IF the following conditions exist, THEN RC-5, Pressurizer Quench Tank 2, 2.a, & all bullets (PQT) vent path should be used per the following:

• Verify:

• There is water in the PQT AND DC Bus 2 electrical power source is available to RCGVS Valves

• Small quantities of gas need to be vented

• Rapid venting is not required Standard: IDENTIFIED the following:

• A bubble exists in the Reactor Vessel Head. (NOT critical)

• Control power exists to the valves. (NOT critical)

• Quench Tank is the preferred venting path. (critical)

• Attachment 4, Venting RV Head to the Pressurizer Quench Tank. (NOT critical)

CIRCLED VENT to: PZR Quench Tank (critical)

Terminating Cue: This JPM is complete.

Comment: SAT UNSAT STOP TIME:

Page 8 of 9 NRC Admin JPM RA4 Rev. Final

Appendix C JPM CUE SHEET Form ES-C-1 INITIAL CONDITIONS: Given the following conditions:

• A Small Break Loss of Coolant Accident (LOCA) has occurred.

• EOP-03, Loss of Coolant Accident, has been implemented.

• Reactor Coolant System (RCS) conditions:

• RCS Pressure = 450 psia.

• RCS TCOLD = 402°F.

• Pressurizer (PZR) conditions:

• PZR Level [actual] = 60% and stable.

• PZR Temperature = 456°F and stable.

• Reactor Vessel Level Monitoring System (RVLMS) is 83%

and stable.

• Containment conditions:

• Containment Safety Function is satisfied.

• All Containment Ventilation Fans are operating.

• Containment Pressure = 1.2 psig.

• Containment Temperature = 118°F.

• Containment Hydrogen concentration = 1.2%.

• RC-5, Pressurizer Quench Tank (PZR QT), conditions:

• PZR QT Level = 70%.

• PZR QT Pressure = 5 psig.

• HPSI Stop and Throttle has been performed for a LOCA.

• RCS and Pressurizer sample results are normal.

• The Technical Support Center (TSC) is activated.

• Use of EOP/AOP Attachment IC-14, RCS Void Elimination, has been unsuccessful in eliminating the RCS voids.

INITIATING CUE: The Control Room Supervisor directs you to PERFORM the following:

• IMPLEMENT OI-RC-12, Post Accident Venting of Noncondensable Gases from the Reactor Coolant System.

• VENT from (CIRCLE one):

Reactor Vessel Head PZR

• VENT to (CIRCLE one):

Containment PZR Quench Tank Page 9 of 9 NRC Admin JPM RA4 Rev. Final

Appendix C JPM WORKSHEET Form ES-C-1 Facility: FCS JPM # NRC SA1 Task # 1363 K/A # 2.1.25 3.9 / 4.2

Title:

Perform an Alternate Decay Heat Removal Method Determination Examinee (Print):

Testing Method:

Simulated Performance: Classroom: X Actual Performance: X Simulator:

Alternate Path: Plant:

Time Critical:

READ TO THE EXAMINEE I will explain the Initial Conditions, which steps to simulate or discuss, and provide an Initiating Cue.

When you complete the task successfully, the objective for this JPM will be satisfied.

Initial Conditions: Given the following conditions:

• Plant was operating for 6 weeks at 100% power when a Reactor Coolant Pump seal failed.

• AOP-19, Loss of Shutdown Cooling, has been entered 7 days after shutdown.

• The Pressurizer manway has been removed.

• HCV-347, Shutdown Cooling Loop 2 Isolation Valve, is closed and cannot be reopened.

• High Pressure Safety Injection (HPSI) Pump SI-2A is available with flow of 300 gpm at 250 psia discharge pressure.

Initiating Cue: The Shift Manager directs you to PERFORM the following:

• PERFORM AOP-19, Loss of Shutdown Cooling, Attachment D, Alternate Decay Heat Removal Method Determination.

• CIRCLE the appropriate Alternate Decay Heat Removal Attachment on Attachment D (indicate your decision path on Attachment D).

Task Standard: Utilizing AOP-19, determined Reactor Coolant System pressure boundary was not intact, Reactor Vessel Head was installed, Shutdown Cooling flow was available, HPSI flow was available but insufficient, and identified Attachment E as the Alternate Decay Heat Removal Method.

Required Materials: AOP-19, Loss of Shutdown Cooling, Rev. 18.

Validation Time: 11 minutes Completion Time: ________ minutes Comments:

Result: SAT UNSAT Examiner (Print / Sign): Date:

Page 1 of 5 NRC Admin JPM SA1 post exam update.docx

Appendix C JPM WORKSHEET Form ES-C-1 CLASSROOM SETUP EXAMINER:

PROVIDE the examinee with a copy of:

• AOP-19, Loss of Shutdown Cooling.

• Attachment D, Alternate Decay Heat Removal Method Determination.

Page 2 of 5 NRC Admin JPM SA1 post exam update.docx

Appendix C JPM STEPS Form ES-C-1

- Check Mark Denotes Critical Step START TIME:

Examiner Note: The following steps are from AOP-19, Attachment D.

Examiner Note: REFER to Answer Key to follow Attachment D flowpath.

Perform Step: 1 IS RCS Pressure Boundary Intact?

1 Standard: DETERMINED answer was NO on Attachment D and PROCEEDED to next box.

Comment: SAT UNSAT Perform Step: 2 IS Reactor Vessel Head on?

2 Standard: DETERMINED answer was YES on Attachment D and PROCEEDED to next box.

Comment: SAT UNSAT Perform Step: 3 Is SDC Discharge Available?

3 Standard: DETERMINED answer was Yes on Attachment D and PROCEEDED to next box.

Comment: SAT UNSAT Perform Step: 4 Is HPSI Discharge Available?

4 Standard: DETERMINED answer was YES on Attachment D and PROCEEDED to next box.

Comment: SAT UNSAT Page 3 of 5 NRC Admin JPM SA1 post exam update.docx

Appendix C JPM STEPS Form ES-C-1 Perform Step: 5 Is Sufficient Injection Available?

5 Standard: PERFORMED the following:

• REFERRED to note (*) and DETERMINED plant operation at 100% power for greater than 30 days.

• DETERMINED Time after Shutdown was 7 days ago.

• DETERMINED Required (gpm) is >310 gpm but < 385 gpm.

• DETERMINED Sufficient Injection Flow NOT available and PROCEEDED to the next box.

Comment: SAT UNSAT Perform Step: 6 GO TO Attachment E Alternate Decay Heat Removal by Boiling.

6 Standard: DETERMINED Attachment E, Alternate Decay Heat Removal by Boiling is the appropriate Attachment.

Terminating Cue: This JPM is complete.

Comment: SAT UNSAT STOP TIME:

Page 4 of 5 NRC Admin JPM SA1 post exam update.docx

Appendix C JPM CUE SHEET Form ES-C-1 INITIAL CONDITIONS: Given the following conditions:

• The plant was operating for 6 weeks at 100% power when a Reactor Coolant Pump seal failed.

• AOP-19, Loss of Shutdown Cooling, has been entered 7 days after shutdown.

• The Pressurizer manway has been removed.

• HCV-347, Shutdown Cooling Loop 2 Isolation Valve, is closed and cannot be reopened.

• High Pressure Safety Injection (HPSI) Pump SI-2A is available with flow of 300 gpm at 250 psia discharge pressure.

INITIATING CUE: The Shift Manager directs you to PERFORM the following:

• PERFORM AOP-19, Loss of Shutdown Cooling, Attachment D, Alternate Decay Heat Removal Method Determination.

• CIRCLE the appropriate Alternate Decay Heat Removal Attachment on Attachment D (indicate your decision path on Attachment D).

Page 5 of 5 NRC Admin JPM SA1 post exam update.docx

Attachment D Alternate Decay Heat Removal Method Determination Determine the method of Alternate Decay Heat Removal from the Flow Chart.

• Appropriate flow required to remove heat by Time after shutdown Required injection, based on 100% full power (days) (gpm) operation for greater than 30 days. 1 575 5 385 10 310 30 230 End of Attachment D

Attachment D Alternate Decay Heat Removal Method Determination Determine the method of Alternate Decay Heat Removal from the Flow Chart.

• Appropriate flow required to remove heat by Time after shutdown Required injection, based on 100% full power (days) (gpm) operation for greater than 30 days. 1 575 5 385 10 310 30 230 End of Attachment D

Appendix C JPM WORKSHEET Form ES-C-1 Facility: FCS JPM # NRC SA2 Task # 1528 K/A # 2.1.43 4.1 / 4.3

Title:

Calculate an Estimated Critical Boron Concentration Examinee (Print):

Testing Method:

Simulated Performance: Classroom: X Actual Performance: X Simulator:

Alternate Path: Plant:

Time Critical:

READ TO THE EXAMINEE I will explain the Initial Conditions, which steps to simulate or discuss, and provide an Initiating Cue.

When you complete the task successfully, the objective for this JPM will be satisfied.

Initial Conditions: Given the following conditions:

• A Reactor Trip from 100% power occurred on 12/03/15 at 0400 following 6 months of full power operation.

• All Control Rods were fully withdrawn at the time of the trip.

• Boron concentration prior to the trip was 400 ppm.

• Average Core Burnup is 6 GWD/MTU.

• Current boron concentration is 500 ppm.

• Criticality is scheduled to occur with Regulating Group 4 at 78 inches.

• Reactor Engineering reports no correction is needed for boron depletion.

• Reactor Startup is scheduled for 12/10/15 at 0600.

Initiating Cue: The Shift Manager directs you to PERFORM the following:

• CALCULATE estimated critical boron concentration per TDB-V.1.B, Estimated Critical Conditions Worksheet.

• COMPLETE data entry through TDB-V.1.B, Step F.4, Estimated Critical Condition Summary.

Task Standard: Utilizing TDB-V.1.B and TDB-II, calculated Estimated Critical Boron Concentration and Minimum and Maximum Critical Rod Position, completed the Estimated Critical Condition Summary.

Required Materials: TDB-V.1.B, Estimated Critical Conditions Worksheet, Rev. 26.

TDB-II, Technical Data Book Reactivity Curves, Rev. 35.

TDB-VI, Core Operating Limits Report, Rev. 42 Validation Time: 37 minutes Completion Time: ________ minutes Comments:

Result: SAT UNSAT Examiner (Print / Sign): Date:

Page 1 of 5 NRC Admin JPM SA2 Rev. Final

Appendix C JPM WORKSHEET Form ES-C-1 CLASSROOM SETUP EXAMINER:

PROVIDE the examinee with a copy of:

• TDB-V.1.B, Estimated Critical Conditions Worksheet

• TDB-II, Technical Data Book Reactivity Curves

• TDB-VI, Core Operating Limits Report

• Calculator

• Straight Edge Page 2 of 5 NRC Admin JPM SA2 Rev. Final

Appendix C JPM STEPS Form ES-C-1

- Check Mark Denotes Critical Step START TIME:

Examiner Note: The following steps are from TDB-V.1.B.

Perform Step: 1 Conditions at Time of Shutdown.

A Standard: ENTERED Conditions at Time of Shutdown in TDB-V.1.B Steps A.1 to A.5.

Examiner Note: Information found on Answer Key.

Comment: SAT UNSAT Perform Step: 2 Conditions at Time of Startup.

B Standard: ENTERED Conditions at Time of Startup in TDB-V.1.B Steps B.1 to B.4.

Examiner Note: Information found on Answer Key.

Comment: SAT UNSAT Perform Step: 3 ECC Applicability.

C Standard: DETERMINED early and late date/time limits for ECC Applicability and entered data in TDB-V.1.B Steps C.1 to C.4.

Examiner Note: Information found on Answer Key.

Comment: SAT UNSAT Perform Step: 4 Reactivity Changes Due To Shutdown.

D Standard: CALCULATED and ENTERED Reactivity Changes Due To Shutdown in TDB-V.1.B Steps D.1 to D.5.

Examiner Note: Information found on Answer Key.

Comment: SAT UNSAT Page 3 of 5 NRC Admin JPM SA2 Rev. Final

Appendix C JPM STEPS Form ES-C-1 Perform Step: 5 Estimated Critical Boron Concentration.

E Standard: CALCULATED and ENTERED Estimated Critical Boron Concentration in TDB-V.1.B Steps E.1 to E.3.

Step E.3.d: Calculated 884 +/- 25 ppm (Critical)

Examiner Note: Information found on Answer Key.

Comment: SAT UNSAT Perform Step: 6 Minimum and Maximum Critical Rod Position.

F Standard: CALCULATED and ENTERED Minimum and Maximum Critical Rod Position in TDB-V.1.B Steps F.1 to F.4.

Step F.4: Estimated Critical Condition Summary:

• Minimum critical position: Group 3 at 75 (+/- 5) inches (Critical)

• Estimated critical position: Group 4 at 78 (+/- 5) inches (Critical)

• Maximum critical position: Group 4 at 126 (- 5) inches (Critical)

Examiner Note: Information found on Answer Key.

Terminating Cue: This JPM is complete.

Comment: SAT UNSAT STOP TIME:

Page 4 of 5 NRC Admin JPM SA2 Rev. Final

Appendix C JPM CUE SHEET Form ES-C-1 INITIAL CONDITIONS: Given the following conditions:

• A Reactor Trip from 100% power occurred on 12/03/15 at 0400 following 6 months of full power operation.

• All Control Rods were fully withdrawn at the time of the trip.

• Boron concentration prior to the trip was 400 ppm.

• Average Core Burnup is 6 GWD/MTU.

• Current boron concentration is 500 ppm.

• Criticality is scheduled to occur with Regulating Group 4 at 78 inches.

• Reactor Engineering reports no correction is needed for boron depletion.

• Reactor Startup is scheduled for 12/10/15 at 0600.

INITIATING CUE: The Shift Manager directs you to PERFORM the following:

• CALCULATE estimated critical boron concentration per TDB-V.1.B, Estimated Critical Conditions Worksheet.

• COMPLETE data entry through TDB-V.1.B, Step F.4, Estimated Critical Condition Summary.

Page 5 of 5 NRC Admin JPM SA2 Rev. Final

Appendix C JPM WORKSHEET Form ES-C-1 Facility: FCS JPM # NRC SA3 Task # 1260 K/A # 2.2.40 3.4 / 4.7

Title:

Determine In-Core Instrumentation Operability Examinee (Print):

Testing Method:

Simulated Performance: Classroom: X Actual Performance: X Simulator:

Alternate Path: Plant:

Time Critical:

READ TO THE EXAMINEE I will explain the Initial Conditions, which steps to simulate or discuss, and provide an Initiating Cue.

When you complete the task successfully, the objective for this JPM will be satisfied.

Initial Conditions: Given the following conditions:

• Plant is at 100% power.

• An In-Core Detector Status Map was just completed for Cycle 28.

Initiating Cue: The Shift Manager directs you to PERFORM the following:

• EVALUATE the In-Core detector system indications and detector status per OI-NI-2, In-Core Instrumentation Operability Requirements.

• In-Core Instrumentation System OPERABILITY per OI-NI-2 (CIRCLE): YES / NO

• IDENTIFY required actions, if any:

Task Standard: Utilizing OI-NI-2, evaluated In-Core Instrumentation Map and determined that the Incore Instrument Instrument system is Operable, and Technical Specification LCO 2.10.4(1)(a)(i) & (ii) actions are required to: Apply an increase of 1% to the total uncertainties for maximum radial peaking factor (FRT) and the total peaking factor (FQT); and Increase the frequency of performing RE-ST-RX-0001 to a minimum of once every 15 days.

Required Materials: OI-NI-2, In-Core Instrumentation Operability Requirements, Rev. 9.

TDB-I.A.7.C, Core Exit Thermocouple Status, Rev. 89.

Fort Calhoun Station Technical Specifications, Amendment #283.

Validation Time: 22 minutes Completion Time: ________ minutes Comments:

Result: SAT UNSAT Examiner (Print / Sign): Date:

Page 1 of 5 NRC Admin JPM SA3 Rev. Final As Run As Run

Appendix C JPM WORKSHEET Form ES-C-1 CLASSROOM SETUP EXAMINER:

PROVIDE the examinee with a copy of:

• OI-NI-2, In-Core Instrumentation Operability Requirements.

• TDB-I.A.7.C, Core Exit Thermocouple Status.

Page 2 of 5 NRC Admin JPM SA3 Rev. Final As Run As Run

Appendix C JPM STEPS Form ES-C-1

- Check Mark Denotes Critical Step START TIME:

Examiner Note: The following steps are from OI-NI-2, Attachment 1.

Perform Step: 1 WHEN either of the following conditions are met, THEN the In-core 1, 1.a, 1.b, & 1.b.1) Detector System is considered operable:

• At least 75% of all In-core Detector Strings are operable and at least two In-core Detector Strings are operable per full Axial Quadrant.

• Between 28% and 75% of all In-core Detector Strings are operable and:

• At least two In-core Detector Strings are operable per full Axial Quadrant Standard: REVIEWED TDB-I.A.&.C, Core Exit Thermocouple Status map and PERFORMED the following:

• DETERMINED 8 of 28 In-Core Detector Strings are inoperable (71.4%), which is between 28% and 75%

• DETERMINED at least two In-core Detector Strings are OPERABLE per full Axial Quadrant.

• CIRCLED YES.

Comment: SAT UNSAT Page 3 of 5 NRC Admin JPM SA3 Rev. Final As Run As Run

Appendix C JPM STEPS Form ES-C-1 Perform Step: 2 WHEN either of the following conditions are met, THEN the In-core 1, 1.b, 1.b.2), & 1.b.3) Detector System is considered operable:

• Between 28% and 75% of all In-core Detector Strings are operable and:

• An increase of 1% to the total uncertainties shall be applied to the maximum radial peaking factor (FRT) and the total peaking factor (FqT) and

• The frequency of performing RE-ST-RX-0001 is changed to a minimum of once every 15 days.

Standard: IDENTIFIED Required Actions and RECORDED the following:

• Apply an increase of 1% to the total uncertainties for maximum radial peaking factor (FRT) and the total peaking factor (FQT), and

• Increase the frequency of performing RE-ST-RX-0001 to a minimum of once every 15 days.

Terminating Cue: This JPM is complete.

Comment: SAT UNSAT STOP TIME:

Page 4 of 5 NRC Admin JPM SA3 Rev. Final As Run As Run

Appendix C JPM CUE SHEET Form ES-C-1 INITIAL CONDITIONS: Given the following conditions:

• Plant is at 100% power.

• An In-Core Detector Status Map, was just completed for Cycle 28.

INITIATING CUE: The Shift Manager directs you to PERFORM the following:

• EVALUATE the In-Core detector system indications and detector status per OI-NI-2, In-Core Instrumentation Operability Requirements.

• In-Core Instrumentation System OPERABILITY per OI-NI-2 (CIRCLE): YES / NO

• IDENTIFY required actions, if any:

Page 5 of 5 NRC Admin JPM SA3 Rev. Final As Run As Run

Appendix C JPM WORKSHEET Form ES-C-1 Facility: FCS JPM # NRC SA4 Task # 0741 K/A # 2.3.7 3.5 / 3.8

Title:

Authorize a Liquid Waste Release Examinee (Print):

Testing Method:

Simulated Performance: Classroom: X Actual Performance: X Simulator:

Alternate Path: Plant:

Time Critical:

READ TO THE EXAMINEE I will explain the Initial Conditions, which steps to simulate or discuss, and provide an Initiating Cue.

When you complete the task successfully, the objective for this JPM will be satisfied.

Initial Conditions: Given the following conditions:

• Raw Water Pumps AC-10A, AC-10B, and AC-10C are operating.

• Monitor Tank A is to be released and was placed on recirculation four hours ago.

• The permit has just been received in the Control Room to release Monitor Tank A.

Initiating Cue: The Shift Manager directs you to PERFORM the following:

• REVIEW the Liquid Release Permit and Plant Conditions and CIRCLE the results:

• Correct Tank is being DISCHARGED? YES / NO

• Maximum Allowable Flow DETERMINED? YES / NO

• Unloader Flow Rate SATISFACTORY? YES / NO

• Dilution Pump alignment SATISFACTORY? YES / NO Task Standard: Utilizing FC-211, determined Monitor Tank A is being discharged, Maximum Allowable Flow Rate identified as 140 gpm, Unloader Flow Rate is unsatisfactory and improper Pump alignment is in service.

Required Materials: FC-211, Waste Liquid Tank Release Permit, Rev. 25.

Validation Time: 13 minutes Completion Time: ________ minutes Comments:

Result: SAT UNSAT Examiner (Print / Sign): Date:

Page 1 of 4 NRC Admin JPM SA4 Rev. Final

Appendix C JPM WORKSHEET Form ES-C-1 CLASSROOM SETUP EXAMINER:

PROVIDE the examinee with a copy of:

• FC-211, Waste Liquid Tank Release Permit.

Page 2 of 4 NRC Admin JPM SA4 Rev. Final

Appendix C JPM STEPS Form ES-C-1

- Check Mark Denotes Critical Step START TIME:

Examiner Note: The following steps are from FC-211.

Perform Step: 1 Is the correct Tank being discharged?

Standard: DETERMINED Monitor Tank A is being released and CIRCLED YES.

Comment: SAT UNSAT Examiner Note: FC-211, Step IV, Maximum Release Rate Calculations - Set Unloader to a flow rate of 130 gpm which is less than or equal to 90% of the maximum release rate listed in Part IV.

Perform Step: 2 Is the maximum allowable flow rate determined?

Standard: DETERMINED Maximum Allowable Flow Rate set at 140 gpm and CIRCLED YES.

Comment: SAT UNSAT Examiner Note: FC-211, Step VII, Special Instructions, Item B. - Set Unloader to a flow rate of 130 gpm which is less than or equal to 90% of the maximum release rate listed in Part IV.

Perform Step: 3 Is the Unloader Flow Rate satisfactory?

Standard: DETERMINED Unloader Flow Rate is NOT satisfactory and CIRCLED NO. (Unloader Flow Rate should be 126 gpm (90% of 140 gpm).)

Comment: SAT UNSAT Examiner Note: FC-211, Step VII, Special Instructions, Item C. - Maintain 2 Circulating Water Pumps in operation.

Perform Step: 4 Is the Pump alignment satisfactory?

Standard: DETERMINED Pump alignment is NOT satisfactory and CIRCLED NO.

(3 Raw Water Pumps are operating and Step VII, Item C calls for 2 Circulating Water Pumps.)

Terminating Cue: This JPM is complete.

Comment: SAT UNSAT STOP TIME:

Page 3 of 4 NRC Admin JPM SA4 Rev. Final

Appendix C JPM CUE SHEET Form ES-C-1 INITIAL CONDITIONS: Given the following conditions:

• Raw Water Pumps AC-10A, AC-10B, and AC-10C are operating.

• Monitor Tank A is to be released and was placed on recirculation four hours ago.

• The permit has just been received in the Control Room to release Monitor Tank A.

INITIATING CUE: The Shift Manager directs you to PERFORM the following:

• REVIEW the Liquid Release Permit and Plant Conditions and CIRCLE the results:

• Correct Tank is being DISCHARGED? YES / NO

• Maximum Allowable Flow DETERMINED? YES / NO

• Unloader Flow Rate SATISFACTORY? YES / NO

• Dilution Pump alignment SATISFACTORY? YES / NO Page 4 of 4 NRC Admin JPM SA4 Rev. Final

Appendix C JPM WORKSHEET Form ES-C-1 Facility: FCS JPM # NRC SA5 Task # 1453 K/A # 2.4.41 2.9 / 4.6

Title:

Classify an Emergency Plan Event Examinee (Print):

Testing Method:

Simulated Performance: Classroom: X Actual Performance: X Simulator:

Alternate Path: Plant:

Time Critical:

READ TO THE EXAMINEE I will explain the Initial Conditions, which steps to simulate or discuss, and provide an Initiating Cue.

When you complete the task successfully, the objective for this JPM will be satisfied.

Initial Conditions: Given the following conditions:

• The Site has received an aircraft attack notification from the Nuclear Regulatory Commission Headquarters Operations Officer that was confirmed by Security at 1200.

• Actions of AOP-37, Security Events, are being implemented.

• Aircraft arrival time is estimated at 60 minutes.

Initiating Cue: The Shift Manager directs you to PERFORM the following:

• DETERMINE the Recognition Category and Event Classification per EPIP-OSC-1, Emergency Plan.

• Emergency Classification Level __________

• IC/EAL Classification __________

Task Standard: Utilizing EPIP-OSC-1 and TDB-EPIP-OSC-1H, determined Recognition Category and classified the event as a Notification of Unusual Event Category HU4.

Required Materials: EPIP-OSC-1, Emergency Plan, Rev. 48b.

TDB-EPIP-OSC-1H, Recognition Category H - Hazards and Other Conditions Affecting Plant Safety, Rev. 3.

Validation Time: 5 minutes Completion Time: ________ minutes Comments:

Result: SAT UNSAT Examiner (Print / Sign): Date:

Page 1 of 4 NRC Admin JPM SA5 Rev. Final

Appendix C JPM WORKSHEET Form ES-C-1 CLASSROOM SETUP EXAMINER:

PROVIDE the examinee with a copy of:

• EPIP-OSC-1, Emergency Plan.

• EPIP-OSC-1A/1C/1E/1F/1H/1S, EPIP Recognition Category Basis Documents.

• FC-EPF-42, Emergency Planning Form NOTE:

PROVIDE the entire EPIP-OSC-1A/1C/1E/1F/1H/1S, EPIP Recognition Category Basis Documents.

Page 2 of 4 NRC Admin JPM SA5 Rev. Final

Appendix C JPM STEPS Form ES-C-1

- Check Mark Denotes Critical Step START TIME:

Examiner Note: The following steps are from Fort Calhoun Station Emergency Action Levels.

Examiner Note: The Applicant may reference TDB-EPIP-OSC-1H which is the EPIP Bases document for HAZARDS.

Perform Step: 1 DETERMINE the Event Category.

Standard: REFERRED to FCS Emergency Action Levels:

• Figure 8.1, Recognition Categories That Apply to Operating Modes Greater Than OR Equal to 210°F.

• Figure 8.1, Recognition Categories That Apply to Operating Modes Less Than to 210°F.

Comment: SAT UNSAT Perform Step: 2 MATCH plant conditions in the Recognition Category.

Standard: IDENTIFIED EAL Recognition Category H - Hazards and Other Conditions Affecting Plant Safety.

Comment: SAT UNSAT Perform Step: 3 Declare the event emergency level.

Standard: IDENTIFIED Emergency level - NOUE (Notification of Unusual Event)

Comment: SAT UNSAT Perform Step: 4 Classify the event.

Standard: CLASSIFIED the event as a NOTIFICATION OF UNUSUAL EVENT (HU4), EAL 3. Confirmed SECURITY CONDITION or threat which indicates a potential degradation in the level of safety of the plant, EAL

1. 3: A validated notification from NRC providing information of an aircraft threat.

Terminating Cue: This JPM is complete.

Comment: SAT UNSAT STOP TIME:

Page 3 of 4 NRC Admin JPM SA5 Rev. Final

Appendix C JPM CUE SHEET Form ES-C-1 INITIAL CONDITIONS: Given the following conditions:

• The Site has received an aircraft attack notification from the Nuclear Regulatory Commission Headquarters Operations Officer that was confirmed by Security at 1200.

• Actions of AOP-37, Security Events, are being implemented.

• Aircraft arrival time is estimated at 60 minutes.

INITIATING CUE: The Shift Manager directs you to PERFORM the following:

• DETERMINE the Recognition Category and Event Classification per EPIP-OSC-1, Emergency Plan.

• Emergency Classification Level __________

• IC/EAL Classification __________

Page 4 of 4 NRC Admin JPM SA5 Rev. Final

Appendix C JPM WORKSHEET Form ES-C-1 Facility: FCS JPM # NRC S-1 Task # 0675 K/A # 001.A2.11 4.4 / 4.7 SF-1

Title:

Perform Control Element Assembly Exercises Examinee (Print):

Testing Method:

Simulated Performance: Classroom:

Actual Performance: X Simulator: X Alternate Path: X Plant:

Time Critical:

READ TO THE EXAMINEE I will explain the Initial Conditions, which steps to simulate or discuss, and provide an Initiating Cue.

When you complete the task successfully, the objective for this JPM will be satisfied.

Initial Conditions: Given the following conditions:

• Maintenance on Shutdown Group A was just completed.

• A partial movement check of Shutdown Group A is required.

• CEAs are in an All-Rods-Out configuration.

Initiating Cue: The Control Room Supervisor directs you to PERFORM the following:

• COMPLETE Control Element Assembly exercise on Shutdown Group A per OP-ST-CEA-0003, Control Element Assembly Partial Movement Check.

Task Standard: Utilizing OP-ST-CEA-0003, exercised Shutdown Group A CEAs then tripped the Reactor when 2 CEAs dropped by opening CEDM Clutch Power Supply Breakers.

Required Materials: OP-ST-CEA-0003, Control Element Assembly Partial Movement Check, Rev. 14.

EOP-00, Standard Post Trip Actions, Rev. 33.

Validation Time: 20 minutes Completion Time: ________ minutes Comments:

Result: SAT UNSAT Examiner (Print / Sign): Date:

Page 1 of 7 NRC JPM S-1 Rev. Final

Appendix C JPM WORKSHEET Form ES-C-1 SIMULATOR SETUP BOOTH OPERATOR:

INITIALIZE to IC-112:

• ENSURE DCS Computer Screen set at CEA ALL.

Type Item Value Condition Event ATWAS_PLUS MALF/CRD ROD_PWR_A30_1 (Rod 30 DE-ENERGIZED When second rod motion is clutch failure) performed MALF/CRD ROD_PWR_A33_1 (Rod 33 DE-ENERGIZED When second rod motion is clutch failure) performed BOOTH OPERATOR NOTE:

• After each JPM, VERIFY all control switches and reactor trip pushbutton cover is restored to normal condition prior to performance by the next examinee.

EXAMINER:

PROVIDE the examinee with a copy of:

• OP-ST-CEA-0003, Control Element Assembly Partial Movement Check.

• INITIALED through Prerequisites.

• N/A all CEAs from 14 to 1 on Attachment 1.

Page 2 of 7 NRC JPM S-1 Rev. Final

Appendix C JPM STEPS Form ES-C-1

- Check Mark Denotes Critical Step START TIME:

Examiner Note: The following steps are from OP-ST-CEA-0003.

NOTE Step 7.1 can be performed at anytime and repeated as necessary.

Perform Step: 1 IF this Surveillance Test is turned over, a prejob briefing must be 7.1 conducted prior to the continuation of this test.

Standard: ACKNOWLEDGED a pre-job brief is required prior to continuing.

Comment: SAT UNSAT CAUTION When the Reactor is critical, this Surveillance Test must be performed within the specified Technical Specification time interval regardless of rod configuration or use.

Perform Step: 2 IF not in an All-Rods-Out configuration, THEN contact the Reactor 7.2 Engineer prior to commencing this test for guidance to ensure the requirements of Technical Specification 3.2, Table 3-5, Item 2 are met.

Standard: DETERMINED CEAs in an All-Rods-Out configuration per Initial Conditions.

Comment: SAT UNSAT Perform Step: 3 Record Initial Position of all CEAs on Attachment 1.

7.3 Standard

RECORDED Initial Position of Shutdown Group A CEAs #30, #31, #32,

1. 33, #34, #35, #36, and #37 on Attachment 1.

Comment: SAT UNSAT Perform Step: 4 Rotate the Mode Selector Switch (M/M) to the Manual Individual (M/I) 7.4 position.

Standard: ROTATED Mode Selector Switch (M/M) to Manual Individual (M/I) position.

Comment: SAT UNSAT Page 3 of 7 NRC JPM S-1 Rev. Final

Appendix C JPM STEPS Form ES-C-1 Perform Step: 5 Rotate the Group Selector Switch (M/G) to the Group containing the 7.5 CEA to be moved.

Standard: ROTATED Group Selector Switch (M/G) to Shutdown Group A.

Comment: SAT UNSAT Perform Step: 6 If available, verify on SCEAPIS (DCS) display CEA_ALL that the group 7.6 button is DARK GREY for the group selected.

Standard: VERIFIED on Secondary Control Element Assembly Position Indicating System (SCEAPIS) Digital Control System (DCS) display CEA_ALL that Shutdown Group A button is DARK GREY.

Comment: SAT UNSAT Examiner Cue: If questioned, REPORT the CRS directs you to start with CEA #30.

Perform Step: 7 Rotate the Rod Selector Switch to the CEA to be moved.

7.7 Standard

ROTATED Rod Selector Switch to any Shutdown Group A CEA.

Comment: SAT UNSAT NOTE If Group 4 CEAs are being used for ASI control, movement of 6 inches in a single direction may be credited. The returned to position may not necessarily be the initial position. Note the time of Group 4 movement for ASI control on the Comment Sheet if applicable.

Examiner Note: When the 2nd CEA is exercised, two CEAs will drop into the Core.

Perform Step: 8 Insert or withdraw the CEA, as applicable, a minimum of six (6) inches, 7.8 THEN return the CEA to its Initial Position.

Standard: INSERTED CEA a minimum of six (6) inches, MONITORED Nuclear Instrumentation and TAVE then WITHDREW CEA to its Initial Position.

Examiner Cue: If the ROD DRIVE POWER INTERRUPT alarm is received (in the event the CEA is moved 8 inches), REPORT as CRS that permission is granted to use the Rod Block Bypass Switch to move the CEA back to its original position.

Examiner Note: Candidate may reposition the Manual Mode Selector switch to off to respond to alarms. If so, the candidate must return the switch to MI to move the selected rod.

Comment: SAT UNSAT Page 4 of 7 NRC JPM S-1 Rev. Final

Appendix C JPM STEPS Form ES-C-1 NOTE Step 7.9 may be completed after all CEAs within a Group have been exercised, after all CEAs have been exercised, OR after exercising each CEA.

Examiner Note: When the 2nd CEA is exercised, two CEAs will drop into the Core.

Perform Step: 9 Record Inserted/Withdrawn To AND Return To information on Test Data Sheet, THEN initial Attachment 1.

Standard: Record Inserted/Withdrawn To AND Return To information on Test Data Sheet, THEN initial Attachment 1.

Comment: SAT UNSAT Examiner Note: The following steps represent the Alternate Path of this JPM.

Perform Step: 10 Determine that 2 CEAs have dropped into the core.

Standard: OBSERVED Annunciator Alarms and DETERMINED two CEAs have dropped.

Comment: SAT UNSAT Examiner Note: The following steps are from EOP-00, Standard Post Trip Actions.

Perform Step: 11 Verify Reactivity Control is established by performing steps a and b:

1

• Verify ALL of the following:

• No more than one Regulating or Shutdown CEA is NOT inserted

• Reactor power is lowering

• Startup rate is negative Standard: DETERMINED Reactor did NOT trip when both CEAs dropped and REFERRED to CONTINGENCY ACTIONS (CA).

Comment: SAT UNSAT Perform Step: 12a IF the reactor did NOT trip, THEN establish Reactivity Control by 1.1 & 1.1.a CA performing step a, b, c or d:

• Manually trip the Reactor (CB-4).

Standard: DEPRESSED REACTOR TRIP pushbutton on CB-4 and DETERMINED Reactor did NOT trip.

Comment: SAT UNSAT Page 5 of 7 NRC JPM S-1 Rev. Final

Appendix C JPM STEPS Form ES-C-1 Perform Step: 12b IF the reactor did NOT trip, THEN establish Reactivity Control by 1.1 & 1.1.b CA performing step a, b, c or d:

• Manually trip the Reactor (AI-31).

Standard: DEPRESSED REACTOR TRIP pushbutton on AI-31 and DETERMINED Reactor did NOT trip.

Comment: SAT UNSAT Perform Step: 12c IF the reactor did NOT trip, THEN establish Reactivity Control by 1.1 & 1.1.c CA performing step a, b, c or d:

• Place the DSS Manual Trip Switches in "TRIP" (AI-66A/B).

Standard: PERFORMED the following:

• PLACED DSS Manual Trip Switch in TRIP position on AI-66A and DETERMINED Reactor did NOT trip.

• PLACED DSS Manual Trip Switch in TRIP position on AI-66B and DETERMINED Reactor did NOT trip.

Comment: SAT UNSAT Perform Step: 12d IF the reactor did NOT trip, THEN establish Reactivity Control by 1.1 & 1.1.d CA performing step a, b, c or d:

• Manually open the CEDM Clutch Power Supply Breakers (AI-57).

Standard: PERFORMED the following:

• OPENED both CLUTCH POWER SUPPLY BREAKER RPS/

CB-A/B on AI-57 (critical).

• OPENED both CLUTCH POWER SUPPLY BREAKER RPS/

CB-C/D on AI-57 (critical)

• OBSERVED all Rod Bottom lights LIT on SCEAPIS (NOT critical).

Terminating Cue: This JPM is complete.

Comment: SAT UNSAT STOP TIME:

Page 6 of 7 NRC JPM S-1 Rev. Final

Appendix C JPM CUE SHEET Form ES-C-1 INITIAL CONDITIONS: Given the following conditions:

• Maintenance on Shutdown Group A was just completed.

• A partial movement check of Shutdown Group A is required.

• CEAs are in an All-Rods-Out configuration.

INITIATING CUE: The Control Room Supervisor directs you to PERFORM the following:

• COMPLETE Control Element Assembly exercise on Shutdown Group A per OP-ST-CEA-0003, Control Element Assembly Partial Movement Check.

Page 7 of 7 NRC JPM S-1 Rev. Final

Appendix C JPM WORKSHEET Form ES-C-1 Facility: FCS JPM # NRC S-2 Task # 1391 K/A # 004.A4.08 3.8 / 3.4 SF-2

Title:

Align Charging Flow Via the HPSI Header Examinee (Print):

Testing Method:

Simulated Performance: Classroom:

Actual Performance: X Simulator: X Alternate Path: Plant:

Time Critical:

READ TO THE EXAMINEE I will explain the Initial Conditions, which steps to simulate or discuss, and provide an Initiating Cue.

When you complete the task successfully, the objective for this JPM will be satisfied.

Initial Conditions: Given the following conditions:

• AOP-33, CVCS Leak, is in progress.

• Leak isolation has restored Charging Pump CH-1C.

• AOP-33, Step 13.d, directs use of Attachment C.

Initiating Cue: The Control Room Supervisor directs you to PERFORM the following:

• ALIGN Charging flow via the High Pressure Safety Injection (HPSI) header per AOP-33, CVCS Leak, Attachment C, Charging Via the HPSI Header Using Only CH-1C.

Task Standard: Utilizing AOP-33, opened HCV-308, opened HCV-312, and started Charging Pump CH-1C to restore Pressurizer level.

Required Materials: AOP-33, CVCS Leak, Rev. 9.

Validation Time: 13 minutes Completion Time: ________ minutes Comments:

Result: SAT UNSAT Examiner (Print / Sign): Date:

Page 1 of 6 NRC JPM S-2 Rev. Final

Appendix C JPM WORKSHEET Form ES-C-1 SIMULATOR SETUP BOOTH OPERATOR:

INITIALIZE to IC-112:

• VERIFY Pressurizer level is lowered to 55%.

• ENSURE all Charging Loop Isolation Valves CLOSED per AOP-33.

• ENSURE all Auxiliary Spray Valves are CLOSED per AOP-33.

• ENSURE all Charging Pumps in PULL-TO-LOCK per AOP-33.

Type Item Value Condition Remote/CVC REM:CVC_CH172 0 Remote/CVC REM:CVC_CH173 0 Remote/CVC REM:CVC_CH191 0 Remote/CVC REM:CVC_CH192 0 Remote/CVC REM:CVC_CH193 0 Remote/CVC REM:CVC_CH194 0 EXAMINER:

PROVIDE the examinee with a copy of:

• AOP-33, CVCS Leak.

• Attachment C, Charging Via the HPSI Header Using Only CH-1C.

Page 2 of 6 NRC JPM S-2 Rev. Final

Appendix C JPM STEPS Form ES-C-1

- Check Mark Denotes Critical Step START TIME:

Examiner Note: The following steps are from AOP-33, Attachment C.

NOTE Charging flow can be verified on the associated HPSI flow indicator(s) for the HPSI Loop Valve(s) in use, or on ERF (Page 323).

Perform Step: 1 Ensure all Charging Pumps are in "PULL-TO-LOCK".

1 Standard: DETERMINED all Charging Pumps in PULL-TO-LOCK.

Comment: SAT UNSAT Perform Step: 2 Unlock and close BOTH of the following valves:

2 & all bullets

• CH-194, "CHARGING PUMPS CH-1A/B/C DISCHARGE HEADER CONTAINMENT OUTBOARD ISOLATION VALVE" (Room 13)

• CH-191, "CHARGING PUMPS CH-1A & B DISCHARGE HEADER TO SAFETY INJECTION ISOLATION VLV." (Charging Pump Valve Room)

Standard: CONTACTED Auxiliary Operator to UNLOCK and CLOSE CH-194 in Room 13 and CH-191 in Charging Pump Valve Room.

Booth Operator: When contacted, UNLOCK and CLOSE CH-194 and CH-191.

REPORT CH-194 and CH-191 UNLOCKED and CLOSED.

Comment: SAT UNSAT Perform Step: 3 Close ALL of the following valves:

3 & all bullets

• CH-192, "CHARGING PUMP CH-1B DISCHARGE VALVE" (Charging Pump Valve Room)

• CH-173, "CHARGING PUMP CH-1B SUCTION VALVE" (Charging Pump Valve Room)

• CH-193, "CHARGING PUMP CH-1A DISCHARGE VALVE" (Charging Pump Valve Room)

• CH-172, "CHARGING PUMP CH-1A SUCTION VALVE" (Charging Pump Valve Room)

Standard: CONTACTED Auxiliary Operator to UNLOCK and CLOSE CH-192, CH-173, CH-193, and CH-172 in Charging Pump Valve Room.

Booth Operator: When contacted, CLOSE CH-192, CH-173, CH-193, and CH-172.

REPORT CH-192, CH-173, CH-193, and CH-172 CLOSED.

Comment: SAT UNSAT Page 3 of 6 NRC JPM S-2 Rev. Final

Appendix C JPM STEPS Form ES-C-1 Perform Step: 4 Open HCV-308, Charging Pump HPSI Header Isolation Valve.

4 Standard: PERFORMED the following:

• PLACED HCV-308, CHARGING PUMP/HPSI HDR ISOLATION VALVE in OPEN (critical).

• OBSERVED red OPEN light lit (NOT critical).

Comment: SAT UNSAT Perform Step: 5 Ensure ALL of the following valves are open:

5 & all bullets

• HCV-2987, HPSI Header Isolation Valve

• HCV-307, HPSI Header Isolation Valve

• HCV-305, SI-2A and SI-2C Discharge Cross-Connect Valve

• HCV-304, SI-2B and SI-2C Discharge Cross-Connect Valve

• HCV-306, HPSI Header Isolation Valve Standard: VERIFIED all of the following valves OPEN and red OPEN lights lit:

• HCV-2987, HPSI HEADER NUMBER 2 DISCHARGE VALVE

• HCV-307, HPSI HEADER NUMBER 2 DISCHARGE VALVE

• HCV-305, HPSI PUMPS SI-2A/SI-2C DISCH CROSSCONNECT VLV

• HCV-304, HPSI PUMPS SI-2B/SI-2C DISCH CROSSCONNECT VLV

• HCV-306, HPSI HEADER NUMBER 1 DISCHARGE VALVE Comment: SAT UNSAT Examiner Note: HCV-312 was selected for consistency of Applicants.

Examiner Cue: The CRS directs you to open HCV-312, HPSI Loop Injection Valve.

Perform Step: 6a Open at least ONE of the following HPSI Loop Injection Valves:

6, 6.a, & 6.a.1)

• Open HCV-312 (Loop 1B) by performing the following:

• Rotate thumbwheel for PCV-2909, "LEAKAGE CLR SI-4A DISCH VLV CNTRLR" fully clockwise to close "C".

Standard: PERFORMED the following:

• VERIFIED thumbwheel for PCV-2909, LEAKAGE CLR SI-4A DISCHARGE VALVE CONTROLLER fully CLOCKWISE in CLOSE (C position).

• OBSERVED needle in C position.

Comment: SAT UNSAT Page 4 of 6 NRC JPM S-2 Rev. Final

Appendix C JPM STEPS Form ES-C-1 Perform Step: 6b Open at least ONE of the following HPSI Loop Injection Valves:

6, 6.a, & 6.a.2)

• Open HCV-312 (Loop 1B) by performing the following:

• Place PCV-2909, "LEAKAGE CLR SI-4A DISCHARGE VALVE" in "MANUAL".

Standard: PERFORMED the following:

• PLACED PCV-2909, LEAKAGE CLR SI-4A DISCHARGE VALVE to MANUAL position (critical).

• OBSERVED switch in MANUAL (NOT critical).

• OBSERVED amber light off and green light lit (NOT critical).

Comment: SAT UNSAT Perform Step: 6c Open at least ONE of the following HPSI Loop Injection Valves:

6, 6.a, & 6.a.3)

• Open HCV-312 (Loop 1B) by performing the following:

• Open HCV-312, "LOOP 1B HPSI INJECTION VALVE".

Standard: PERFORMED the following:

• TURNED and HELD HCV-312, LOOP 1B HPSI INJECTION VALVE to OPEN position (critical).

• OBSERVED red OPEN light lit (NOT critical).

Comment: SAT UNSAT NOTE Charging flow can be verified on the associated HPSI flow indicator(s) for the HPSI Loop Valve(s) in use, or on ERF (Page 323).

Perform Step: 7 Operate CH-1C as necessary to maintain PZR level within 4% of 7 programmed level.

Standard: PERFORMED the following:

• PLACED CH-1C, CHRG PUMP in START (critical).

• OBSERVED red START light lit (NOT critical).

• OBSERVED flow on HCV-312, HPSI Loop Injection Valve or on ERF Computer Page 323 (NOT critical).

• OBSERVED ~75 amps on CH-1C ammeter (NOT critical).

Terminating Cue: This JPM is complete.

Comment: SAT UNSAT STOP TIME:

Page 5 of 6 NRC JPM S-2 Rev. Final

Appendix C JPM CUE SHEET Form ES-C-1 INITIAL CONDITIONS: Given the following conditions:

• AOP-33, CVCS Leak, is in progress.

• Leak isolation has restored Charging Pump CH-1C.

• AOP-33, Step 13.d, directs use of Attachment C.

INITIATING CUE: The Control Room Supervisor directs you to PERFORM the following:

• ALIGN Charging flow via the High Pressure Safety Injection (HPSI) header per AOP-33, CVCS Leak, Attachment C, Charging Via the HPSI Header Using Only CH-1C.

Page 6 of 6 NRC JPM S-2 Rev. Final

Appendix C JPM WORKSHEET Form ES-C-1 Facility: FCS JPM # NRC S-3 Task # 1129 K/A # 009.EA2.34 3.6 / 4.2 SF-3

Title:

Perform HPSI Stop and Throttle Examinee (Print):

Testing Method:

Simulated Performance: Classroom:

Actual Performance: X Simulator: X Alternate Path: X Plant:

Time Critical:

READ TO THE EXAMINEE I will explain the Initial Conditions, which steps to simulate or discuss, and provide an Initiating Cue.

When you complete the task successfully, the objective for this JPM will be satisfied.

Initial Conditions: Given the following conditions:

• A Small Break Loss of Coolant Accident is in progress.

• EOP-03, Loss of Coolant Accident, has been entered.

Initiating Cue: The Control Room Supervisor directs you to PERFORM the following:

• EVALUATE then EXECUTE actions for HPSI Stop and Throttle per EOP/AOP Floating Step F, HPSI Stop and Throttle Criteria.

Task Standard: Utilizing Floating Step F, stopped all but one HPSI Pump and throttled Loop Injection Valves. Upon leak increase, restarted HPSI Pumps and opened Loop Injection Valves as required.

Required Materials: EOP/AOP Floating Steps, Rev. 7.

Validation Time: 10 minutes Completion Time: ________ minutes Comments:

Result: SAT UNSAT Examiner (Print / Sign): Date:

Page 1 of 6 NRC JPM S-3 Rev. Final As Run As Run

Appendix C JPM WORKSHEET Form ES-C-1 SIMULATOR SETUP BOOTH OPERATOR:

INITIALIZE to IC-118:

Type Item Value Condition MALF/RCS RCS01B (RCS-01B - RCS Loop 0.43 Recall/modify after flow Leak - Loop 1B Cold Leg ** Leak is present and level are balanced, Medium) when you restore increase leak rate to value this IC 1.5 EXAMINER:

PROVIDE the examinee with a copy of:

• EOP/AOP Floating Step F, HPSI Stop and Throttle Criteria.

Page 2 of 6 NRC JPM S-3 Rev. Final As Run As Run

Appendix C JPM STEPS Form ES-C-1

- Check Mark Denotes Critical Step START TIME:

Examiner Note: The following steps are from Floating Step FS-F.

CAUTIONS

1. If emergency boration is required then at least one charging pump must remain running.

2. As natural circulation develops, the expected rise in TH will reduce subcooling. This may jeopardize HPSI Stop and Throttle Criteria.

3. Reducing SI flow should be approached cautiously.

4. The purpose of HPSI stop and throttle is to prevent an over pressurization of the RCS and a solid PZR, however, maintaining RCS inventory is more important than pressure control.

Perform Step: 1 Verify ALL of the following stop and throttle criteria are satisfied:

1 & all bullets

• RCS subcooling is greater than or equal to 20°F

• PZR level is greater than or equal to 10% and not lowering

• At least one S/G is available for RCS heat removal

• RVLMS indicates level is at or above the top of the Hot Leg (43%, ERF "I" display)

Standard: OBSERVED the following:

• RCS subcooling greater than 20°F.

• Pressurizer level greater than 10% and not lowering.

• Both Steam Generators are available for RCS heat removal.

• Reactor Vessel Level Monitoring System is greater than 43%.

Comment: SAT UNSAT Examiner Note: Applicant must place any 2 of 3 Charging Pumps in PULL-TO-LOCK otherwise they will AUTO START.

Perform Step: 2 Ensure only ONE Charging Pump is operating.

2 Standard: STOPPED 2 of 3 Charging Pumps by PERFORMING the following (any 2):

• PLACED CH-1A, CHARGING PUMP in PULL-TO-LOCK (critical).

• PLACED CH-1B, CHARGING PUMP in PULL-TO-LOCK (critical).

• PLACED CH-1C, CHARGING PUMP in PULL-TO-LOCK (critical).

• OBSERVED pump breaker lights off (NOT critical).

Comment: SAT UNSAT Page 3 of 6 NRC JPM S-3 Rev. Final As Run As Run

Appendix C JPM STEPS Form ES-C-1 CAUTIONS

1. During a UHE HPSI stop and throttle should be performed before the expansion of the relatively cold SI water overfills the pressurizer.

2. Operators should closely monitor RCS pressure-temperature limits. Pressurizer spray may be required to prevent exceeding the maximum subcooling limit.

3. Allowing the RCS to repressurize to 1300 psia will effectively stop HPSI flow.

Perform Step: 3 IF a UHE is in progress, THEN maintain RCS pressure control by 3 performing the following:

Standard: DETERMINED Uncontrolled Heat Extraction is NOT in progress.

Comment: SAT UNSAT Examiner Note: It is acceptable to stop one or more HPSI Pumps and throttle HPSI Loop Injection Valves to achieve control over Pressurizer level, resulting in stable or slowly rising Pressurizer level. Applicant may stop one HPSI Pump in this step and/or throttle HPSI Loop Injection Valves to achieve this condition.

CAUTIONS

1. LOCAs pose a significant threat to RCS subcooling. Therefore, full SI Flow should be maintained until subcooled margin is stable and natural circulation has developed.

2. During a SGTR, the depressurization of the RCS to less than 1000 psia should be stopped when HPSI flow is initially being stopped and throttled.

Perform Step: 4a IF a LOCA or SGTR is in progress, THEN maintain RCS pressure 4 & 4.a control by performing the following:

• Ensure at least one HPSI Pump is operating.

Standard: DETERMINED SI-2A and SI-2B, HPSI Pumps are running. May stop one HPSI pump or leave both running.

Comment: SAT UNSAT Perform Step: 4b IF a LOCA or SGTR is in progress, THEN maintain RCS pressure 4 & 4.b control by performing the following:

• Throttle HPSI Loop Injection Valve(s).

Standard: THROTTLED CLOSE any or all the following:

• HCV-314, LOOP 1A HPSI INJECTION VALVE.

• HCV-317, LOOP 2A HPSI INJECTION VALVE.

• HCV-311, LOOP 1B HPSI INJECTION VALVE.

• HCV-321, LOOP 2B HPSI INJECTION VALVE.

Comment: SAT UNSAT Page 4 of 6 NRC JPM S-3 Rev. Final As Run As Run

Appendix C JPM STEPS Form ES-C-1 Examiner Note: The following steps represent the Alternate Path of this JPM.

Examiner Note: Once HPSI flow has been throttled, the break size will increase and Stop and Throttle criteria will no longer be met. HPSI Pumps must be restarted and/or Loop Injection Valves reopened. Applicant must recognize this and begin to take action before Reactor Vessel Level Monitoring System (RVLMS) indicates less than 100%.

Perform Step: 5a IF HPSI stop and throttle criteria can NOT be maintained, THEN raise 5 & 5.a HPSI flow by performing the following:

• Start either HPSI Pumps, SI-2A/B or SI-2B/C, as necessary.

Standard: DETERMINED SI-2A and SI-2B, HPSI Pumps are running. If HPSI pumps were secured in step 4, restart HPSI pumps. (critical if one of more HPSI pumps were secured)

Comment: SAT UNSAT Examiner Note: Applicant should throttle open valve(s) closed at Perform Step 4b.

Perform Step: 5b IF HPSI stop and throttle criteria can NOT be maintained, THEN raise HPSI flow by performing the following:

• Open HPSI Loop Injection Valves, as necessary.

Standard: OPEN any or all the following:

• HCV-314, LOOP 1A HPSI INJECTION VALVE.

• HCV-317, LOOP 2A HPSI INJECTION VALVE.

• HCV-311, LOOP 1B HPSI INJECTION VALVE.

• HCV-321, LOOP 2B HPSI INJECTION VALVE.

Terminating Cue: This JPM is complete.

Comment: SAT UNSAT STOP TIME:

Page 5 of 6 NRC JPM S-3 Rev. Final As Run As Run

Appendix C JPM CUE SHEET Form ES-C-1 INITIAL CONDITIONS: Given the following conditions:

• A Small Break Loss of Coolant Accident is in progress.

• EOP-03, Loss of Coolant Accident, has been entered.

INITIATING CUE: The Control Room Supervisor directs you to PERFORM the following:

• EVALUATE then EXECUTE actions for HPSI Stop and Throttle per EOP/AOP Floating Step F, HPSI Stop and Throttle Criteria.

Page 6 of 6 NRC JPM S-3 Rev. Final As Run As Run

Appendix C JPM WORKSHEET Form ES-C-1 Facility: FCS JPM # NRC S-4 Task # 0612 K/A # 003.A2.02 3.7 / 3.9 SF-4P

Title:

Start a Reactor Coolant Pump Examinee (Print):

Testing Method:

Simulated Performance: Classroom:

Actual Performance: X Simulator: X Alternate Path: Plant:

Time Critical:

READ TO THE EXAMINEE I will explain the Initial Conditions, which steps to simulate or discuss, and provide an Initiating Cue.

When you complete the task successfully, the objective for this JPM will be satisfied.

Initial Conditions: Given the following conditions:

• Plant Startup is in progress.

• Reactor Coolant Pumps RC-3A, RC-3B, and RC-3C are running.

Initiating Cue: The Control Room Supervisor directs you to PERFORM the following:

• START Reactor Coolant Pump RC-3D per OI-RC-9, Reactor Coolant Pump Operation, Attachment 1, Starting Reactor Coolant Pumps (Coupled).

• START at Step 11.

Task Standard: Utilizing OI-RC-9, started RC-3D-1 Oil Lift Pump and RCP RC-3D.

Required Materials: OI-RC-9, Reactor Coolant Pump Operation, Rev. 78.

Validation Time: 7 minutes Completion Time: ________ minutes Comments:

Result: SAT UNSAT Examiner (Print / Sign): Date:

Page 1 of 6 NRC JPM S-4 Rev. Final

Appendix C JPM WORKSHEET Form ES-C-1 SIMULATOR SETUP BOOTH OPERATOR:

INITIALIZE to IC-116:

• ENSURE Reactor Coolant Pump RC-3-D is STOPPED.

• ENSURE ERF Computer Page 342 or DCS "RCP Summary" on display.

EXAMINER:

PROVIDE the examinee with a copy of:

• OI-RC-9, Reactor Coolant Pump Operation.

• Attachment 1, Starting Reactor Coolant Pumps (Coupled), INITIALED through Step 10.

Page 2 of 6 NRC JPM S-4 Rev. Final

Appendix C JPM STEPS Form ES-C-1

- Check Mark Denotes Critical Step START TIME:

Examiner Note: The following steps are from OI-RC-9, Attachment 1.

NOTE The Oil Lift Pump shall not be run for longer than 10 minutes before starting the Reactor Coolant Pump.

Perform Step: 1 Announce the Reactor Coolant Pump start on the Gaitronics.

11 Standard: ANNOUNCED start of Reactor Coolant Pump RC-3D on Gaitronics.

Comment: SAT UNSAT Perform Step: 2 Place the Oil Lift Pump for the selected RCP to AFTER START:

12

• RC-3D-1, Oil Lift Pump Standard: PERFORMED the following:

• PLACED RC-3D-1, OIL LIFT PUMP handswitch in AFTER START (critical).

• OBSERVED red indicating light lit (NOT critical).

Comment: SAT UNSAT Examiner Note: ARRD is the Anti-Reverse Rotation Device on the RCP.

Perform Step: 3 Verify adequate ARRD Lube Oil Flow for the selected RCP. ERF/DCS 13 indication shall read NORMAL: (ERF page 342 or DCS "RCP Summary")

• RC-3D F3190 Standard: OBSERVED ERF Computer Page 342 or Digital Computer System RCP Summary and VERIFIED ARRD Lube Oil Flow for RC-3D is NORMAL.

Comment: SAT UNSAT Perform Step: 4 Prior to starting RCP, inform the Radiation Protection Department so it 14 can monitor changing radiological conditions.

Standard: CONTACTED Radiation Protection Department about RCP start.

Examiner Cue: Radiation Protection acknowledges start of RCP.

Comment: SAT UNSAT Page 3 of 6 NRC JPM S-4 Rev. Final

Appendix C JPM STEPS Form ES-C-1 Perform Step: 5a Startup sequence for selected RCP:

15 & 15.1

• Run the oil lift pump for the selected RCP a minimum of 2 minutes.

• RC-3D-1, Oil Lift Pump Standard: DETERMINED RC-3D-1, OIL LIFT PUMP already running.

Examiner Cue: If Applicant begins timing, REPORT two minutes have passed.

Comment: SAT UNSAT Perform Step: 5b Startup sequence for selected RCP:

15 & 15.2

• Place the selected RCP control switch in AFTER START:

• RC-3D, RC Pump Standard: PERFORMED the following:

• PLACED RC-3D, RC PUMP handswitch in AFTER START (critical).

• OBSERVED red START light lit (NOT critical).

Comment: SAT UNSAT Perform Step: 5c Startup sequence for selected RCP:

15 & 15.3

• IF the Reactor Coolant Pump motor amps fail to drop below 425 amps within the time listed below, THEN place the control switch in AFTER STOP:

• RC-3D - seventeen (17) seconds Standard: DETERMINED RC-3D ammeter reads less than 425 amps in less than 17 seconds.

Comment: SAT UNSAT Perform Step: 6a Verify the following for the selected RCP:

16 & 16.1

• Oil Lift Pump stops (Green indicating light ON):

Standard: OBSERVED RC-3D-1, OIL LIFT PUMP green indicating light lit.

Comment: SAT UNSAT Page 4 of 6 NRC JPM S-4 Rev. Final

Appendix C JPM STEPS Form ES-C-1 Perform Step: 6b Verify the following for the selected RCP:

16 & 16.2

• For the selected RCP verify the following Annunciator is clear:

• RC-3D REACTOR COOLANT PUMP RC-3D REVERSE ROTATION (CB-1/2/3, A6, D5)

Standard: OBSERVED CB-1/2/3/A6, Window D REACTOR COOLANT PUMP RC-3D REVERSE ROTATION is CLEAR.

Comment: SAT UNSAT Perform Step: 6c Verify the following for the selected RCP:

16 & 16.3

• For the selected RCP verify the following Annunciator is clear:

• RC-3D REACTOR COOLANT PUMP RC-3D VIBRATION HI (CB-1/2/3, A6, D4)

Standard: OBSERVED CB-1/2/3/A6, Window D REACTOR COOLANT PUMP RC-3D VIBRATION HI is CLEAR.

Comment: SAT UNSAT NOTE At low RCS Pressure, verification of positive Controlled Bleedoff Flow may NOT be possible.

Perform Step: 7 Verify positive Controlled Bleedoff flow for the selected RCP:

17 (ERF page 342 or DCS "RCP Summary")

• RC-3D F3175 Standard: OBSERVED ERF Computer Page 342 or Digital Computer System RCP Summary and VERIFIED Control Bleedoff Flow for RC-3D is POSITIVE.

Comment: SAT UNSAT Perform Step: 8 Monitor the ERF Computer or DCS and verify all parameters are normal 18 for the selected RCP:

• RC-3D Standard: MONITORED ERF Computer or Digital Computer System and VERIFIED RCP RC-3D parameters NORMAL.

Terminating Cue: This JPM is complete.

Comment: SAT UNSAT STOP TIME:

Page 5 of 6 NRC JPM S-4 Rev. Final

Appendix C JPM CUE SHEET Form ES-C-1 INITIAL CONDITIONS: Given the following conditions:

• Plant Startup is in progress.

• Reactor Coolant Pumps RC-3A, RC-3B, and RC-3C are running.

INITIATING CUE: The Control Room Supervisor directs you to PERFORM the following:

• START Reactor Coolant Pump RC-3D per OI-RC-9, Reactor Coolant Pump Operation, Attachment 1, Starting Reactor Coolant Pumps (Coupled).

• START at Step 11.

Page 6 of 6 NRC JPM S-4 Rev. Final

Appendix C JPM WORKSHEET Form ES-C-1 Facility: FCS JPM # NRC S-5 Task # 0033 K/A # G 2.1.19 3.9 / 3.8 SF-4S

Title:

Perform Control Room Evacuation Immediate Actions Examinee (Print):

Testing Method:

Simulated Performance: Classroom:

Actual Performance: X Simulator: X Alternate Path: X Plant:

Time Critical:

READ TO THE EXAMINEE I will explain the Initial Conditions, which steps to simulate or discuss, and provide an Initiating Cue.

When you complete the task successfully, the objective for this JPM will be satisfied.

Initial Conditions: Given the following conditions:

• A toxic gas leak is requiring evacuation of the Control Room.

Initiating Cue: The Control Room Supervisor directs you to PERFORM the following:

• EXECUTE required actions prior to a Control Room Evacuation per AOP-07, Evacuation of Control Room,Section I, Plant to Hot Shutdown.

Task Standard: Utilizing AOP-07, trip the Reactor and Turbine, secured one Main Feedwater, one Condensate, and two Heater Drain Pumps, and started Turbine Lube Oil Pumps.

Required Materials: AOP-07, Evacuation of Control Room, Rev. 17.

Validation Time: 6 minutes Completion Time: ________ minutes Comments:

Result: SAT UNSAT Examiner (Print / Sign): Date:

Page 1 of 6 NRC JPM S-5 Rev. Final

Appendix C JPM WORKSHEET Form ES-C-1 SIMULATOR SETUP BOOTH OPERATOR:

INITIALIZE to IC-111:

• RUN event No Turbine Trip from CR Edit.evt Type Item Value Condition Other THATFS_050A18O_1FREEZE 1 (FREEZE FLAG)

Other THATFS_050A28O_1FREEZE 1 (FREEZE FLAG)

Other THATFS_050A18O_2FREEZE 1 (FREEZE FLAG)

Other THATFS_050A28O_2FREEZE 1 (FREEZE FLAG)

Remote/GE REM:86-1/G1-TRP (86-1/G1 Trip Tripped P10_235SD_1 eq 1 (B EHC N Signal) pump to PTL)

Override P10_102S1_1 0 (FALSE)

Override P10_102S1_1 1 (TRUE) P10_235SD_1 eq 1 (B EHC pump to PTL)

EXAMINER:

PROVIDE the examinee with a copy of:

• AOP-07, Evacuation of Control Room.

• Section I, Plant to Hot Shutdown.

Page 2 of 6 NRC JPM S-5 Rev. Final

Appendix C JPM STEPS Form ES-C-1

- Check Mark Denotes Critical Step START TIME:

Examiner Note: The following steps are from AOP-07,Section I.

Examiner Note: When the Reactor is tripped, an uncontrolled cooldown will begin because the Turbine has NOT tripped.

Perform Step: 1 Perform the following steps prior to evacuating the Control Room:

1 & 1.a

• Manually trip the Reactor.

Standard: PERFORMED the following:

• DEPRESSED REACTOR TRIP pushbutton on CB-4 (critical).

• OBSERVED all Rods inserted, Reactor Power lowering, and Negative Startup Rate. (NOT critical).

Examiner Cue: If cooldown is addressed by applicant, REPORT as Control Room Supervisor that the ATCO will perform any required Emergency Boration. Continue with AOP-07 actions.

Comment: SAT UNSAT Perform Step: 2 Perform the following steps prior to evacuating the Control Room:

1 & 1.b

• Verify the Turbine is tripped as indicated by Stop and Intercept Valves indicating closed.

Standard: OBSERVED Stop and Intercept Valves, DETERMINED Turbine was NOT tripped, and REFERRED to CONTINGENCY ACTIONS (CA).

Comment: SAT UNSAT Examiner Note: The following steps represent the Alternate Path of this JPM.

Perform Step: 2a Trip the Turbine (CB-10, 11) b.1 & b.1.1) CA Standard: PERFORMED the following:

• DEPRESSED TURBINE ST-1 MASTER TRIP PUSHBUTTON A.

• DEPRESSED TURBINE ST-1 MASTER TRIP PUSHBUTTON B.

• OBSERVED all Stop and Intercept Valves OPEN.

Comment: SAT UNSAT Page 3 of 6 NRC JPM S-5 Rev. Final

Appendix C JPM STEPS Form ES-C-1 Perform Step: 2b Trip the Turbine by performing:

b.1 & b.1.2) CA

• Stop the EHC pumps by placing BOTH of the following control switches in "PULL-TO-LOCK":

• EHC-3A

• EHC-3B Standard: PERFORMED the following:

• PLACED EHC-3A, EHC PUMP handswitch in PULL-TO-LOCK (critical).

• PLACED EHC-3B, EHC PUMP handswitch in PULL-TO-LOCK (critical).

• OBSERVED all Stop and Intercept Valves CLOSED and DETERMINED Turbine is tripped (NOT critical).

Comment: SAT UNSAT Perform Step: 3 Perform the following steps prior to evacuating the Control Room:

1 & 1.c

• Place the "43/FW" Switch in "OFF".

Standard: PLACED 43/FW Switch in OFF.

Comment: SAT UNSAT Perform Step: 4 Perform the following steps prior to evacuating the Control Room:

1 & 1.d

• Ensure no more than one Feed Pump, FW-4A/B/C is operating.

Standard: DETERMINED only FW-4C, MFW Pump is running.

Comment: SAT UNSAT Examiner Note: Applicant can stop either Condensate Pump. Operations expectation is to stop FW-2A.

Perform Step: 5 Perform the following steps prior to evacuating the Control Room:

1 & 1.e

• Ensure no more than one Condensate Pump, FW-2A/B/C is operating.

Standard: PERFORMED one of the following:

• PLACED FW-2A, COND PUMP handswitch in STOP (critical).

OR

• PLACED FW-2C, COND PUMP handswitch in STOP (critical).

• OBSERVED green STOP light lit (NOT critical).

Comment: SAT UNSAT Page 4 of 6 NRC JPM S-5 Rev. Final

Appendix C JPM STEPS Form ES-C-1 Perform Step: 6 Perform the following steps prior to evacuating the Control Room:

1 & 1.f

• Stop ALL operating Heater Drain Pumps, FW-5A/B/C.

Standard: PERFORMED the following:

• PLACED FW-5C, HTR DRN PUMP handswitch in STOP (critical).

• OBSERVED green STOP light lit (NOT critical).

Comment: SAT UNSAT Perform Step: 7 Perform the following steps prior to evacuating the Control Room:

1 & 1.g

• Ensure ALL of the following Turbine Lube Oil equipment is running:

• LO-3, Turning Gear Oil Pump

• LO-8, Motor Suction Oil Pump

• LO-4, DC Oil Pump,

• Turbine Lift Pumps, LO-14A/B/C Standard: PERFORMED the following:

• PLACED LO-3, TURNING GEAR OIL PUMP handswitch in START (critical).

• PLACED LO-8, MOTOR SUCTION OIL PUMP handswitch in START (critical).

• PLACED LO-4, EMGY BRG OIL PUMP handswitch in START (critical).

• PLACED LO-14A, TURBINE BEARING LUBE OIL LIFT OIL PUMP handswitch in START (critical).

• PLACED LO-14B, TURBINE BEARING LUBE OIL LIFT OIL PUMP handswitch in START (critical).

• PLACED LO-14C, TURBINE BEARING LUBE OIL LIFT OIL PUMP handswitch in START (critical).

• OBSERVED red START lights lit (NOT critical).

• OBSERVED six white DISCH PRESS lights lit (NOT critical).

Terminating Cue: This JPM is complete.

Comment: SAT UNSAT STOP TIME:

Page 5 of 6 NRC JPM S-5 Rev. Final

Appendix C JPM CUE SHEET Form ES-C-1 INITIAL CONDITIONS: Given the following conditions:

• A toxic gas leak is requiring evacuation of the Control Room.

INITIATING CUE: The Control Room Supervisor directs you to PERFORM the following:

• EXECUTE required actions prior to a Control Room Evacuation per AOP-07, Evacuation of Control Room,Section I, Plant to Hot Shutdown.

Page 6 of 6 NRC JPM S-5 Rev. Final

Appendix C JPM WORKSHEET Form ES-C-1 Facility: FCS JPM # NRC S-6 Task # 0369 K/A # 026.A4.05 3.5 / 3.5 SF-5

Title:

Reset Containment Spray Actuation Signal Examinee (Print):

Testing Method:

Simulated Performance: Classroom:

Actual Performance: X Simulator: X Alternate Path: Plant:

Time Critical:

READ TO THE EXAMINEE I will explain the Initial Conditions, which steps to simulate or discuss, and provide an Initiating Cue.

When you complete the task successfully, the objective for this JPM will be satisfied.

Initial Conditions: Given the following conditions:

• A Main Steam Line Break has occurred inside Containment.

• EOP- 05, Uncontrolled Heat Extraction, is in progress.

• Containment Spray Pumps SI-3A and SI-3B are running.

• Containment Pressure is less than 3 psig.

• All Containment Cooling and Filtering Units are in service.

Initiating Cue: The Control Room Supervisor directs you to PERFORM the following:

• RESET the Containment Spray Actuation Signal and return Containment Spray Pumps to Standby per EOP/AOP Floating Step A, Containment Spray Termination.

Task Standard: Utilizing Floating Step A, reset CPHS, CSAS, and SGLS lockout relays and secured Containment Spray Pumps.

Required Materials: EOP/AOP Floating Steps, Rev. 7.

Validation Time: 15 minutes Completion Time: ________ minutes Comments:

Result: SAT UNSAT Examiner (Print / Sign): Date:

Page 1 of 10 NRC JPM S-6 Rev. Final As Run

Appendix C JPM WORKSHEET Form ES-C-1 SIMULATOR SETUP BOOTH OPERATOR:

INITIALIZE to IC-120:

Type Item Value Condition MALF/SGN SGN01B (Main Steam Line B 0.25 **Simulator is frozen >30 Leak Inside Containment) minutes into a steam header rupture in containment

• PLACE Simulator in RUN then DELETE all CPHS overrides.

BOOTH OPERATOR NOTE:

• VERIFY CPHS overrides are deleted after Simulator is in RUN.

EXAMINER:

PROVIDE the examinee with a copy of:

• EOP/AOP Floating Step A, Containment Spray Termination.

Page 2 of 10 NRC JPM S-6 Rev. Final As Run

Appendix C JPM STEPS Form ES-C-1

- Check Mark Denotes Critical Step START TIME:

Examiner Note: The following steps are from EOP/AOP Floating Steps, FS-A.

NOTE Stopping SI-3A or SI-3B will result in closure of one spray valve, HCV-344 or HCV-345 by interlock which will extend the time to RAS.

CAUTION Containment Spray may affect proper operation of RCPs, non-qualified equipment, Containment Sump, and instrumentation inside the Containment.

When the termination criterion is satisfied, Containment Spray should be promptly secured Perform Step: 1 IF Containment Spray has been initiated AND ALL of the following 1 conditions are satisfied:

• Two CS pumps are operating

• Containment pressure is less than 60 psig and NOT rising

• At least one VA-3A/B in service

• At least one VA-7C/D in service THEN perform the following:

Standard: DETERMINED all conditions are met per Initial Conditions.

Comment: SAT UNSAT Examiner Note: Applicant may go directly to Procedure step 2 (JPM Perform Step

5) and secure BOTH Containment Spray Pumps. If this is done, this critical action is satisfied by the action performed in Procedure step 2.

Applicant may notice a slight rise in Containment temperature when the Containment Spray Pump is secured.

Examiner Cue: If applicant hesitates, REPORT as Control Room Supervisor that the slight temperature rise is expected and they should continue.

Perform Step: 2 Ensure only ONE CS pump is operating.

1.a Standard: PERFORMED the following:

• PLACED SI-3A or SI-3B, CNTMT SPRAY PUMP in PULL-TO-LOCK (critical).

• OBSERVED pump indicating lights off (NOT critical).

Comment: SAT UNSAT Page 3 of 10 NRC JPM S-6 Rev. Final As Run

Appendix C JPM STEPS Form ES-C-1 Examiner Note: Applicant may place HCV-344 or HCV-345, CNTMT SPRAY VLV CONTROL SWITCH in OVERRIDE, but is not required.

Perform Step: 3 Ensure only ONE of the following valves is open:

1.b

• HCV-344

• HCV-345 Standard: PERFORMED the following:

• OBSERVED that red OPEN light lit for one valve and green CLOSED light lit for the other valve.

• HCV-344 is open when SI-3B is running

• HCV-345 is open when SI-3A is running Comment: SAT UNSAT Perform Step: 4 Ensure total CS flow is at least 1800 gpm.

1.c Standard: OBSERVED approximately 2400 gpm of combined flow on FI-343 and FI-342 SPRAY FLOW meters.

Comment: SAT UNSAT NOTE Terminating Containment Spray prior to resetting actuation relays will require increased monitoring of containment parameters.

Examiner Note: Applicant may go directly to Procedure step 2 (JPM Perform Step

5) and secure BOTH Containment Spray Pumps. If this is done, the applicant must secure BOTH Containment Spray Pumps in this step.

Perform Step: 5 IF CS pump(s) are operating, AND ALL of the following conditions are 2 satisfied:

• Containment pressure is less than 30 psig and stable or lowering

• Containment Spray is NOT required for Containment cooling

• At least one VA-3A/B in service

• At least one VA-7C/D in service THEN terminate Containment Spray by performing the following:

Standard: DETERMINED all conditions are met per Initial Conditions.

Comment: SAT UNSAT Page 4 of 10 NRC JPM S-6 Rev. Final As Run

Appendix C JPM STEPS Form ES-C-1 Examiner Note: Applicant will place valve left open at Perform Step 3 in OPEN.

Perform Step: 6 Place the control switches for the open valve(s) in "OPEN":

2.a

• HCV-344

• HCV-345 Standard: PERFORMED the following for the valve that is open:

• PLACED HCV-344 or HCV-345, CNTMT SPRAY VLV CONTROL SWITCH in OPEN (critical).

• OBSERVED red OPEN light lit (NOT critical).

Comment: SAT UNSAT Examiner Note: Applicant will place SI-3A and SI-3B in PULL-TO-LOCK. Due to system configuration, SI-3C is not lined up for auto start, and is NOT required to be placed in PULL-TO-LOCK.

Perform Step: 7 Place all CS pumps in "PULL-TO-LOCK":

2.b

• SI-3A

• SI-3B

• SI-3C Standard: PERFORMED one the following:

• PLACED SI-3A, CNTMT SPRAY PUMP in PULL-TO-LOCK (critical).

• PLACED SI-3B, CNTMT SPRAY PUMP in PULL-TO-LOCK (critical).

• OBSERVED all pump indicating lights off (NOT critical).

Comment: SAT UNSAT Examiner Note: Applicant will place valve left OPEN at Perform Step 6 in CLOSE.

Perform Step: 8 Close BOTH Containment Spray Valves:

2.c Standard: PERFORMED the following:

• PLACED HCV-344 or HCV-345, CNTMT SPRAY VLV CONTROL SWITCH in OVERRIDE or AUTO (critical).

• OBSERVED green CLOSE light lit (NOT critical).

Comment: SAT UNSAT Page 5 of 10 NRC JPM S-6 Rev. Final As Run

Appendix C JPM STEPS Form ES-C-1 Perform Step: 9 Place BOTH control switches in "AUTO":

2.d

• HCV-344

• HCV-345 Standard: PERFORMED the following:

• PLACED HCV-344, CNTMT SPRAY VLV CONTROL SWITCH in AUTO (critical).

• PLACED HCV-345, CNTMT SPRAY VLV CONTROL SWITCH in AUTO (critical)

• OBSERVED white AUTO light off (NOT critical).

Comment: SAT UNSAT NOTES

1. Resetting CPHS and SGLS Lockout Relays may reset SGIS. HCV-1105 and HCV-1106 may reopen.

2. Resetting PPLS, CPHS or SGLS Lockout Relays will reset Containment Spray.

Perform Step: 10 IF resetting actuation relays, THEN perform the following:

3 Standard: DETERMINED actuation relays will be RESET.

Comment: SAT UNSAT Perform Step: 11 IF Containment pressure less than or equal to 3 psig, THEN reset all of 3.a the following relays:

• 86A/CPHS

• 86B/CPHS Standard: PERFORMED the following:

• TURNED 86A/CPHS relay in CLOCKWISE direction until LATCHED (critical).

• TURNED 86B/CPHS relay in CLOCKWISE direction until LATCHED (critical)

• OBSERVED black relay flag and amber light lit (NOT critical).

Comment: SAT UNSAT Page 6 of 10 NRC JPM S-6 Rev. Final As Run

Appendix C JPM STEPS Form ES-C-1 Perform Step: 12 Reset ALL of the following relays:

3.b

• 86A1/CPHS

• 86B1/CPHS Standard: PERFORMED the following:

• TURNED 86A1/CPHS relay in CLOCKWISE direction until LATCHED (critical).

• TURNED 86B1/CPHS relay in CLOCKWISE direction until LATCHED (critical)

• OBSERVED black relay flag and amber light lit (NOT critical).

Comment: SAT UNSAT Perform Step: 13 Reset ALL of the following CSAS relays:

3.c

• 86A/CSAS

• 86B/CSAS

• 86A1/CSAS

• 86B1/CSAS Standard: PERFORMED the following:

• TURNED 86A/CSAS relay in CLOCKWISE direction until LATCHED (critical).

• TURNED 86B/CSAS relay in CLOCKWISE direction until LATCHED (critical)

• TURNED 86A1/CSAS relay in CLOCKWISE direction until LATCHED (critical).

• TURNED 86B1/CSAS relay in CLOCKWISE direction until LATCHED (critical)

• OBSERVED black relay flag and amber light lit (NOT critical).

Comment: SAT UNSAT Page 7 of 10 NRC JPM S-6 Rev. Final As Run

Appendix C JPM STEPS Form ES-C-1 Examiner Note: The following actions are performed at CB-4.

Perform Step: 14 Reset SGLS by performing the following:

3.d

• Block SGLS-A and SGLS-B by performing the following:

• Place the SGLS Block key into the SGLS Block key switch.

• Block SGLS-A and SGLS-B by turning key to "BLOCK".

• Verify at least one of the following SGLS Blocked alarms annunciates (CB-4; A8):

• "SGLS "A" BLOCKED"

• SGLS "B" BLOCKED" Time _____

Standard:

• DETERMINED Steam Generator Low Pressure Signal will be RESET. (NOT critical)

• REMOVED SGLS Block Key from Key Holder and PLACED SGLS Block Key into SGLS Block Key Switch. (critical)

• TURNED Key in SGLS Block Key Switch to BLOCK position.

(critical)

• VERIFIED Annunciator Panel A8 SGLS "A" BLOCKED (Window D-4L) or SGLS "B" BLOCKED (Window D-5U) in alarm and RECORDED time. (NOT critical)

Comment: SAT UNSAT Perform Step: 15 Reset BOTH of the following SGLS relays:

3.d.1).c).2)

• 86A/SGLS

• 86B/SGLS Standard: PERFORMED the following:

• TURNED 86A/SGLS relay in CLOCKWISE direction until LATCHED (critical).

• TURNED 86B/SGLS relay in CLOCKWISE direction until LATCHED (critical).

• OBSERVED black relay flag (NOT critical).

Comment: SAT UNSAT Page 8 of 10 NRC JPM S-6 Rev. Final As Run

Appendix C JPM STEPS Form ES-C-1 Examiner Note: Only SI-3A AND SI-3B will be placed in AFTER STOP.

Perform Step: 16 IF returning CS to standby, THEN perform the following:

3.e.1)

• Place CS Pumps SI-3A/B/C to "AFTER STOP".

Standard: PERFORMED the following:

• PLACED SI-3A, CNTMT SPRAY PUMP in AFTER STOP (critical).

• PLACED SI-3B, CNTMT SPRAY PUMP in AFTER STOP (critical).

• OBSERVED green STOP lights lit (NOT critical).

Comment: SAT UNSAT Perform Step: 17 Place BOTH Containment Spray Valves in "AUTO":

3.e.2)

• HCV-344

• HCV-345 Standard: DETERMINED HCV-344 and HCV-345 already in AUTO.

Terminating Cue: This JPM is complete.

Comment: SAT UNSAT STOP TIME:

Page 9 of 10 NRC JPM S-6 Rev. Final As Run

Appendix C JPM CUE SHEET Form ES-C-1 INITIAL CONDITIONS: Given the following conditions:

• A Main Steam Line Break has occurred inside Containment.

• EOP- 05, Uncontrolled Heat Extraction, is in progress.

• Containment Spray Pumps SI-3A and SI-3B are running.

• Containment Pressure is less than 3 psig.

• All Containment Cooling and Filtering Units are in service.

INITIATING CUE: The Control Room Supervisor directs you to PERFORM the following:

• RESET the Containment Spray Actuation Signal and return Containment Spray Pumps to Standby per EOP/AOP Floating Step A, Containment Spray Termination.

Page 10 of 10 NRC JPM S-6 Rev. Final As Run

Appendix C JPM WORKSHEET Form ES-C-1 Facility: FCS JPM # NRC S-7 Task # 0344 K/A # 064.A4.06 3.9 / 3.9 SF-6

Title:

Parallel and Load Emergency Diesel Generator Examinee (Print):

Testing Method:

Simulated Performance: Classroom:

Actual Performance: X Simulator: X Alternate Path: X Plant:

Time Critical:

READ TO THE EXAMINEE I will explain the Initial Conditions, which steps to simulate or discuss, and provide an Initiating Cue.

When you complete the task successfully, the objective for this JPM will be satisfied.

Initial Conditions: Given the following conditions:

• DG-1 has been manually started and is at idle speed.

• YCV-871 G/H/E Inlet and Exhaust Dampers have been verified OPEN.

• Jacket water temperature is 128°F.

• All Prerequisites are met.

Initiating Cue: The Control Room Supervisor directs you to PERFORM the following:

• PARALLEL and LOAD Emergency Diesel Generator DG-1 with Bus 1A3 per OI-DG-1, Diesel Generator Operation, Attachment 1, Idle Speed Start and Loading.

• START at Step 4.b.

• LOAD DG-1 to 2000 KW.

Task Standard: Utilizing OI-DG-1, raised DG-1 speed, paralleled and loaded to Bus 1A3, then tripped DG-1 Breaker when load rose uncontrollably.

Required Materials: OI-DG-1, Diesel Generator Operation, Rev. 63.

TDB-III.26, Diesel Generator Capability Curve (4160 V), Rev. 5.

TDB-III.26.A, Diesel Generator Loading Curve, Rev. 16.

Validation Time: 15 minutes Completion Time: ________ minutes Comments:

Result: SAT UNSAT Examiner (Print / Sign): Date:

Page 1 of 7 NRC JPM S-7 Rev. Final

Appendix C JPM WORKSHEET Form ES-C-1 SIMULATOR SETUP BOOTH OPERATOR:

INITIALIZE to IC-111. Load and execute scenario NRC JPM S7 Type Item Value Condition Override P20_185_3 (DG/1 Governor Sel 1 Condition: H_P20_033_1 SW Raise Position) ge 800 (When DG Watts are greater than 800, override governor switch to raise)

BOOTH OPERATOR NOTE:

• After each JPM, VERIFY Synchroscope Switch is moved from the D1/BUS 1A3 Sync Switch position prior to performance by the next examinee.

EXAMINER:

PROVIDE the examinee with a copy of:

• OI-DG-1, Diesel Generator Operation.

• Attachment 1, Idle Speed Start and Loading, INITIALED through Step 4.a.

• TDB-III.26, Diesel Generator Capability Curve (4160 V).

• TDB-III.26.A, Diesel Generator Loading Curve.

Page 2 of 7 NRC JPM S-7 Rev. Final

Appendix C JPM STEPS Form ES-C-1

- Check Mark Denotes Critical Step START TIME:

Examiner Note: The following steps are from OI-DG-1, Attachment 1.

Perform Step: 1 Place CS-65/D1, Diesel Generator D1 Governor, to Raise until the 4.b Diesel Speed is 900 rpm.

Standard: PERFORMED the following:

• PLACED CS-65/D1, DIESEL GENERATOR D1 GOVERNOR, to RAISE position until Diesel Speed is 900 rpm. (critical).

• OBSERVED Diesel speed rising to 900 rpm on Diesel Generator DG-1 Engine Tachometer (NOT critical).

Comment: SAT UNSAT Perform Step: 2 Verify the Generator Field flashed by performing one of the following:

4.c

• Ready to Load light is ON (AI-30A)

OR

• Generator frequency is responding Standard: OBSERVED the following:

• Ready to Load red light is ON at Panel AI-30A.

• Generator frequency is 60 Hz at 900 rpm.

Examiner Note: Automatic field flashing occurs at approximately 700 rpm.

Comment: SAT UNSAT Perform Step: 3 (LOCAL) Inspect field flash circuitry by performing the following:

4.d

• Verify that Control Relay 2CR in Panel AI-133A is not energized.

• Verify that Field Flash Current Limiting Resistors (1R4, 1R5, 1R6, and 1R7) in Panel AI-133A are not damaged due to overheating.

Contact System Engineer if damage is suspected.

Standard: CONTACTED Auxiliary Operator at DG-1 to verify Field Flash Circuitry.

Examiner Cue: Auxiliary Operator reports Field Flash Circuitry is satisfactory.

Comment: SAT UNSAT Page 3 of 7 NRC JPM S-7 Rev. Final

Appendix C JPM STEPS Form ES-C-1 Perform Step: 4 (LOCAL) Place AI-133A-S4, Diesel Generator DG-1 Electronic Droop 4.e Control Switch, in ENABLED. (AI-133A)

Standard: CONTACTED Auxiliary Operator at DG-1 to PLACE AI-133A-S4, Diesel Generator DG-1 Electronic Droop Control Switch, in ENABLED position.

Examiner Cue: Auxiliary Operator reports Electronic Droop Control Switch in ENABLED position.

Comment: SAT UNSAT Examiner Note: The panel holding the Synchroscope and Running and Incoming Volts indications is hinged and can be rotated for better viewing.

Perform Step: 5 Place D1/BUS 1A3 Sync Switch to ON.

4.f Standard: PERFORMED the following:

• LOCATED and INSERTED Synchroscope Switch into D1/BUS 1A3 SYNC SWITCH position and TURNED to ON (critical).

• OBSERVED Synchroscope rotation (NOT critical).

Comment: SAT UNSAT Perform Step: 6 Adjust CS-90/D1, Diesel Generator D1 Voltage Regulator, until the 4.g RUNNING VOLTS is approximately matched to the INCOMING VOLTS on the Synchroscope or the ERF DGD Display.

Standard: PERFORMED the following:

• ADJUSTED CS-90/D1, DIESEL GENERATOR D1 VOLTAGE REGULATOR switch until RUNNING VOLTS is approximately MATCHED (within ~100 volts) to INCOMING VOLTS (critical).

Comment: SAT UNSAT NOTE Recommended synchroscope speed is less than 1 revolution per 10 seconds.

Perform Step: 7 Adjust CS-65/D1 until the Synchroscope is rotating slowly in the FAST 4.h direction.

Standard: ADJUSTED CS-65/D1, DIESEL GENERATOR D1 GOVERNOR until Synchroscope is ROTATING SLOWLY in FAST direction and less than 1 revolution per 10 seconds.

Comment: SAT UNSAT Page 4 of 7 NRC JPM S-7 Rev. Final

Appendix C JPM STEPS Form ES-C-1 NOTE Steps 4.i and 4.j may be performed without the procedure in hand.

Sign-offs may be completed after these steps are performed.

CAUTIONS

1. Load must be immediately picked up following closure of 1AD1 to prevent motorizing the Diesel Generator.

2. Governor controls are extremely sensitive.

Perform Step: 8 WHEN the Synchroscope is between 11 and 12 O'CLOCK, THEN close 4.i 1AD1 BREAKER.

Standard: PERFORMED the following:

• When Synchroscope was between 11 and 12 oclock, PLACED 1AD1 BREAKER in CLOSE position (critical).

• OBSERVED red CLOSE light lit (NOT critical).

Comment: SAT UNSAT Perform Step: 9 Place CS-65/D1 to Raise to pick up 250-350 KW.

4.j

• Time _____

Standard: PERFORMED the following:

• PLACED CS-65/D1, DIESEL GENERATOR D1 GOVERNOR in RAISE and PICKED UP 250 KW to 350 KW load (critical).

• OBSERVED load rising on DG-1 WATT METER (NOT critical).

• RECORDED Time at Step 4.j (NOT critical).

Comment: SAT UNSAT Perform Step: 10 Place D1/BUS 1A3 Sync Switch to OFF.

4.k Standard: PLACED D1/BUS 1A3 SYNC SWITCH in OFF.

Comment: SAT UNSAT Perform Step: 11 IF the Diesel is loaded AND Y3287A, ERF 1A3 Bus Voltage, is greater 4.l than 4375 VAC, THEN immediately notify the System Engineer.

Standard: DETERMINED Bus 1A3 voltage is normal.

Comment: SAT UNSAT Page 5 of 7 NRC JPM S-7 Rev. Final

Appendix C JPM STEPS Form ES-C-1 Examiner Note: When DG-1 load is > 800 KW and CS-65/D1 is placed in RAISE, DG-1 load will continue to rise to ~3200 KW.

Examiner Note: The following steps represent the Alternate Path of this JPM.

NOTE

1. Load should be maintained below the 2000 hr Rating vs Ambient Temp curve per TDB-III.26A Figure 1, DG-1 Output Power Rate.

2. Power factor may be determined by using TDB-III.26, Diesel Generator Capability Curve.

3. Current is normally limited to 400 amps at 2500 KW.

4. Diesel Generator manual loading and unloading rates should be maintained at less than 500 KW per minute.

5. Steps 4.m and 4.n may be repeated as necessary while the diesel is loaded. Sign-offs may be completed after these steps are performed.

Examiner Note: Applicant must OPEN the Diesel Generator Output Breaker within 2 minutes of exceeding a loading rate of 500 KW per minute.

Perform Step: 12 Place CS-65/D1 to RAISE picking up the required DG-1 Load.

4.m Standard: PERFORMED the following:

• PLACED CS-65/D1, DIESEL GENERATOR D1 GOVERNOR in RAISE and PICKED UP load (critical).

• DETERMINED DG-1 load rising out of control and PLACED 1AD1 BREAKER in TRIP position (critical).

• OBSERVED green TRIP light lit (NOT critical).

Comment: SAT UNSAT Perform Step: 13 Inform Control Room Supervisor of problem.

Standard: INFORMED Control Room Supervisor DG-1 Output Breaker tripped due to excessive Diesel loading.

Terminating Cue: The CRS has been notified. This JPM is complete.

Comment: SAT UNSAT STOP TIME:

Page 6 of 7 NRC JPM S-7 Rev. Final

Appendix C JPM CUE SHEET Form ES-C-1 INITIAL CONDITIONS: Given the following conditions:

• DG-1 has been manually started and is at idle speed.

• YCV-871 G/H/E Inlet and Exhaust Dampers have been verified OPEN.

• Jacket water temperature is 128°F.

• All Prerequisites are met.

INITIATING CUE: The Control Room Supervisor directs you to PERFORM the following:

• PARALLEL and LOAD Emergency Diesel Generator DG-1 with Bus 1A3 per OI-DG-1, Diesel Generator Operation, Attachment 1, Idle Speed Start and Loading.

• START at Step 4.b.

• LOAD DG-1 to 2000 KW.

Page 7 of 7 NRC JPM S-7 Rev. Final

PAGE 1 OF 2 Fort Calhoun Station Unit 1 TDB-III.26 TECHNICAL DATA BOOK DIESEL GENERATOR CAPABILITY CURVE (4160 VOLTS)

Change No. EC 38104 Reason for Change Correct the Note associated with Diesel Generator Capability Curve.

Requestor Richard Ronning Preparer Daniel A Hochstein Issue Date 03-28-06 3:00 pm R5

FORT CALHOUN STATION TDB-III.26 TECHNICAL DATA BOOK PAGE 2 OF 2 Diesel Generator Capability Curve (4160 Volts)

NOTE: Safe operating area is between the 0.5 and 1.0 power factor lines and less then the 450 amp line.

R5

PAGE 1 OF 9 Fort Calhoun Station Unit 1 TDB-III.26.A TECHNICAL DATA BOOK DIESEL GENERATOR LOADING CURVE Change No. EC 62189 Reason for Change The changes were made as a result of revising EA-FC-92-072.

Adding additional figures to TDB-III.26.

Requestor E. Noseir Preparer K. Bessey Issue Date 10-02-13 1628 R16

FORT CALHOUN STATION TDB-III.26.A TECHNICAL DATA BOOK PAGE 2 OF 9 Figure 1 - (LOCA) DG-1 Output Power Rating Ethylene Glycol Coolant (110.0 Deg. F Maximum Ambient Limit)

R16

FORT CALHOUN STATION TDB-III.26.A TECHNICAL DATA BOOK PAGE 3 OF 9 Figure 2 - (MSLB) DG-1 Output Power Rating Ethylene Glycol Coolant (110.0 Deg. F Maximum Ambient Limit)

R16

FORT CALHOUN STATION TDB-III.26.A TECHNICAL DATA BOOK PAGE 4 OF 9 Figure 3 - (LOCA) DG-2 Output Power Rating Ethylene Glycol Coolant (109 Deg. F Maximum Ambient Limit)

R16

FORT CALHOUN STATION TDB-III.26.A TECHNICAL DATA BOOK PAGE 5 OF 9 Figure 4 - (MSLB) DG-2 Output Power Rating Ethylene Glycol Coolant (105 Deg. F Maximum Ambient Limit)

R16

FORT CALHOUN STATION TDB-III.26.A TECHNICAL DATA BOOK PAGE 6 OF 9 Figure 5 - (LOCA) DG-1 Output Power Rating Water Coolant (110 Deg. F Maximum Ambient Limit)

R16

FORT CALHOUN STATION TDB-III.26.A TECHNICAL DATA BOOK PAGE 7 OF 9 Figure 6 - (MSLB) DG-1 Output Power Rating Water Coolant (110 Deg. F Maximum Ambient Limit)

R16

FORT CALHOUN STATION TDB-III.26.A TECHNICAL DATA BOOK PAGE 8 OF 9 Figure 7 - (LOCA) DG-2 Output Power Rating Water Coolant (114 Deg. F Maximum Ambient Limit)

R16

FORT CALHOUN STATION TDB-III.26.A TECHNICAL DATA BOOK PAGE 9 OF 9 Figure 8 - (MSLB) DG-2 Output Power Rating Water Coolant (105 Deg. F Maximum Ambient Limit)

R16

Appendix C JPM WORKSHEET Form ES-C-1 Facility: FCS JPM # NRC S-8 Task # 0778 K/A # 012.A4.02 3.3 / 3.4 SF-7

Title:

Adjust Reactor Protection System TCOLD Calibration Examinee (Print):

Testing Method:

Simulated Performance: Classroom:

Actual Performance: X Simulator: X Alternate Path: Plant:

Time Critical:

READ TO THE EXAMINEE I will explain the Initial Conditions, which steps to simulate or discuss, and provide an Initiating Cue.

When you complete the task successfully, the objective for this JPM will be satisfied.

Initial Conditions: Given the following conditions:

• Plant is operating at 100% power.

• Channel D TCOLD calibration is indicating high.

Initiating Cue: The Shift Manager directs you to PERFORM the following:

• ADJUST Reactor Protection System (RPS) TCOLD calibration on Channel D per OI-RPS-2, Reactor Protective System-TM/LP TCOLD CAL Calibration.

Task Standard: Utilizing OI-RPS-2, bypassed Channel D TM/LP Trip Unit, adjusted TCOLD CAL Calibration, then returned Channel D TM/LP Trip Unit to service.

Required Materials: OI-RPS-2, Reactor Protective System-TM/LP TCOLD CAL Calibration, Rev. 10.

TM/LP Trip Unit # 9 Bypass Key Validation Time: 19 minutes Completion Time: ________ minutes Comments:

Result: SAT UNSAT Examiner (Print / Sign): Date:

Page 1 of 7 NRC JPM S-8 Rev. Final

Appendix C JPM WORKSHEET Form ES-C-1 SIMULATOR SETUP BOOTH OPERATOR:

INITIALIZE to IC-111 or any at power Initial Condition:

• ENSURE TCOLD Calibrate Potentiometer on Channel D is set to greater than 5.20.

BOOTH OPERATOR NOTE:

• After each JPM, VERIFY Channel D TM/LP Trip Unit # 9 Bypass Key is removed from AI-31D prior to performance by the next examinee.

EXAMINER:

PROVIDE the examinee with a copy of:

• OI-RPS-2, Reactor Protective System-TM/LP TCOLD CAL Calibration.

• INITIALED through Prerequisites.

• PROVIDE the TM/LP Trip Unit # 9 Bypass Key.

EXAMINER NOTE: Only SROs can check out keys from the Key Locker at FCS.

Page 2 of 7 NRC JPM S-8 Rev. Final

Appendix C JPM STEPS Form ES-C-1

- Check Mark Denotes Critical Step START TIME:

Examiner Note: The following steps are from OI-RPS-2.

Perform Step: 1 Record Tcold DVM readings on all four RPS channels.

1.a

• AI-31A °F

• AI-31B °F

• AI-31C °F

• AI-31D °F Standard: SELECTED Channel A/B/C/D TCOLD on Digital Voltmeters (DVM) and RECORDED temperatures at Step 1.a.

Comment: SAT UNSAT Perform Step: 2 Record Tcold cal DVM readings on all four RPS channels.

1.b

• AI-31A °F

• AI-31B °F

• AI-31C °F

• AI-31D °F Standard: SELECTED Channel A/B/C/D TCOLD CAL on Digital Voltmeters (DVM) and RECORDED temperatures at Step 1.b.

Comment: SAT UNSAT Perform Step: 3 Record Tcold cal POT settings.

1.c

• AI-31A _____

• AI-31B _____

• AI-31C _____

• AI-31D _____

Standard: RECORDED Channel A/B/C/D TCOLD CAL POT settings at Step 1.c.

Comment: SAT UNSAT Examiner Note: PROVIDE RPS TM/LP Trip Unit # 9 Bypass Key.

Perform Step: 4 Obtain the RPS TM/LP Trip Unit # 9 Bypass Key.

1.d Standard: OBTAINED RPS TM/LP Trip Unit # 9 Bypass Key from Key Locker.

Comment: SAT UNSAT Page 3 of 7 NRC JPM S-8 Rev. Final

Appendix C JPM STEPS Form ES-C-1 CAUTION Only ONE channel shall be adjusted at a time.

Perform Step: 5 Log into Technical Specification 2.15.1(1) 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> LCO for selected 1.e channel:

• AI-31D Standard: INFORMED CRS of entry into Technical Specification LCO 2.15.1 for Reactor Protection System Channel D.

Examiner Cue: CRS logs entry into Technical Specification LCO.

Comment: SAT UNSAT Perform Step: 6 Bypass TM/LP trip unit # 9 on the selected channel using Bypass Key:

1.f

• AI-31D Standard: PERFORMED the following:

• INSERTED key into Channel D TM/LP Trip Unit # 9 and TURNED to BYPASS Channel D (critical).

• OBSERVED Channel D Trip Unit amber light lit (NOT critical).

Comment: SAT UNSAT Perform Step: 7 Adjust Tcold cal POT on the selected channel until the Tcold cal DVM 1.g reading equals highest RPS channel Tcold recorded in Step a.

• AI-31D Standard: ADJUSTED T COLD CAL POT on Channel D until T COLD CAL DVM reading equals highest RPS channel T COLD recorded in Step a.

Comment: SAT UNSAT Perform Step: 8 Ensure selected TM/LP Trip Unit #9 is RESET by depressing T/U Alarm 1.h Reset.

• AI-31D Standard: PERFORMED the following:

• DEPRESSED T/U Alarm Reset on Channel D TM/LP Trip Unit #9.

• OBSERVED Channel D Trip Unit #9 alarm RESET light off.

Comment: SAT UNSAT Page 4 of 7 NRC JPM S-8 Rev. Final

Appendix C JPM STEPS Form ES-C-1 Perform Step: 9 Verify TM/LP Trip Unit #9 Lights are reset.

1.i Standard: VERIFIED Channel D Trip Unit #9 alarm RESET light off.

Comment: SAT UNSAT Perform Step: 10 Remove Bypass Key for selected TM/LP Trip Unit.

1.j

• AI-31D Standard: PERFORMED the following:

• REMOVED Channel D Trip Unit #9 Bypass Key (critical).

• OBSERVED Channel D Trip Unit amber light off (NOT critical).

Comment: SAT UNSAT Perform Step: 11 Exit Technical Specification 2.15.1(1) for the selected channel.

1.k

• AI-31D Standard: INFORMED CRS of exit from Technical Specification LCO 2.15.1 for Reactor Protection System Channel D.

Examiner Cue: CRS logs exit from Technical Specification LCO.

Comment: SAT UNSAT Perform Step: 12 Repeat Steps e through k for any remaining channels out of 1.l specification.

Standard: DETERMINED there are NO remaining Channels out of specification.

Comment: SAT UNSAT Perform Step: 13 Record Tcold cal DVM readings.

1.m

• AI-31A °F

• AI-31B °F

• AI-31C °F

• AI-31D °F Standard: SELECTED Channel A/B/C/D TCOLD CAL on Digital Voltmeters (DVM) and RECORDED temperatures at Step 1.m.

Comment: SAT UNSAT Page 5 of 7 NRC JPM S-8 Rev. Final

Appendix C JPM STEPS Form ES-C-1 Perform Step: 14 Record Tcold cal POT settings.

1.n

• AI-31A _____

• AI-31B _____

• AI-31C _____

• AI-31D _____

Standard: RECORDED Channel A/B/C/D TCOLD CAL POT settings at Step 1.n.

Terminating Cue: This JPM is complete.

Comment: SAT UNSAT STOP TIME:

Page 6 of 7 NRC JPM S-8 Rev. Final

Appendix C JPM CUE SHEET Form ES-C-1 INITIAL CONDITIONS: Given the following conditions:

• Plant is operating at 100% power.

• Channel D TCOLD calibration is indicating high.

INITIATING CUE: The Shift Manager directs you to PERFORM the following:

• ADJUST Reactor Protection System (RPS) TCOLD calibration on Channel D per OI-RPS-2, Reactor Protective System-TM/LP TCOLD CAL Calibration.

Page 7 of 7 NRC JPM S-8 Rev. Final

Appendix C JPM WORKSHEET Form ES-C-1 Facility: FCS JPM # NRC P-1 Task # 1398 K/A # 033.A2.02 2.7 / 3.0 SF-8

Title:

Spent Fuel Pool Cooling Restoration with SIAS Examinee (Print):

Testing Method:

Simulated Performance: X Classroom:

Actual Performance: Simulator:

Alternate Path: Plant: X Time Critical:

READ TO THE EXAMINEE I will explain the Initial Conditions, which steps to simulate or discuss, and provide an Initiating Cue.

When you complete the task successfully, the objective for this JPM will be satisfied.

Initial Conditions: Given the following conditions:

• AOP-36, Loss of Spent Fuel Pool Cooling, has been entered following a Safety Injection Actuation Signal.

• SIAS and CIAS have been RESET.

• 480 Volt Bus 1B4C is energized and available for Spent Fuel Pool Cooling Pump AC-5B.

• HC-478, STORAGE POOL HX AC-8 AC OUTL HCV-478 at AI-45 is OPEN.

Initiating Cue: The Control Room Supervisor directs you to PERFORM the following:

• RESTORE Spent Fuel Pool Cooling using Spent Fuel Pool Cooling Pump AC-5B per AOP-36, Loss of Spent Fuel Pool Cooling, Attachment H, Spent Fuel Pool Cooling Restoration with SIAS.

• START at Step 6.

Task Standard: Utilizing AOP-36, Attachment H, aligned valves, started Spent Fuel Pool Cooling Pump AC-5B, and restored SFP cooling flow.

Required Materials: AOP-36, Loss of Spent Fuel Pool Cooling, Rev. 11.

Validation Time: 15 minutes Completion Time: ________ minutes Comments:

Result: SAT UNSAT Examiner (Print / Sign): Date:

Page 1 of 6 NRC JPM P-1 Rev. Final

Appendix C JPM WORKSHEET Form ES-C-1 PLANT SETUP EXAMINER:

PROVIDE the examinee with a copy of:

• AOP-36, Loss of Spent Fuel Pool Cooling.

• Attachment H, Spent Fuel Pool Cooling Restoration with SAIS.

• INITIALED/PLACE KEEPING through Step 5.

Page 2 of 6 NRC JPM P-1 Rev. Final

Appendix C JPM STEPS Form ES-C-1

- Check Mark Denotes Critical Step START TIME:

Examiner Note: The following steps are from AOP-36, Attachment H, Step 6.

Examiner Note: MCC 4C2 is located in Corridor 4, 989' elev. of Auxiliary Building.

Examiner Note: MCC-4C2-F05 is a bucket style breaker. Turn to the left to RESET, then turn to right to ON.

Perform Step: 1 Verify breaker MCC-4C2-F05, "AC-5B FUEL STORAGE POOL CIRC 6 PUMP" is reset and closed (Corridor 4)

Standard: PERFORMED the following:

• OBSERVED breaker in the ON position with Green light LIT

• Candidate *may* (not required) reset breaker by: TURN handle on MCC-4C2-F05, AC-5B FUEL STORAGE POOL CIRC PUMP to LEFT to RESET breaker, then TURN handle on MCC-4C2-F05, AC-5B FUEL STORAGE POOL CIRC PUMP to RIGHT to ON.

Examiner Cue: Breaker handle pointing to ON. Green light is lit.

Comment: SAT UNSAT Examiner Note: Room 5 is located in Corridor 4, 989' elev. of Auxiliary Building.

Examiner Note: Radiation levels are somewhat elevated in Room 5. By selecting SFP AC-5B, the Applicant can remain in a LOW DOSE WAITING AREA while describing manipulations to be performed.

Perform Step: 2 Contact Shift RP for Room 5 entry.

7 Standard: CONTACTED Shift Radiation Protection for Room 5 entry.

Examiner Cue: Shift Radiation Protection has been contacted and approves entry.

Comment: SAT UNSAT Page 3 of 6 NRC JPM P-1 Rev. Final

Appendix C JPM STEPS Form ES-C-1 Examiner Note: AC-191 is a 90-degree butterfly valve. The operating handle has notches to keep the valve in position. The specific throttle position in this step is not critical, but throttling the valve is critical to prevent pump runout on start. Applicant may determine appropriate throttle position, or the Examiner may provide a cue.

Examiner Cue: If requested: The CRS directs you to place AC-191 at approximately 50% open.

Perform Step: 3 Throttle open the selected SFP pump discharge valve (Room 5):

8

• AC-191, "SPENT FUEL POOL CIRC PUMP AC-5B DISCHARGE VALVE" Standard: PERFORMED the following:

• SQUEEZED handle on AC-191, SPENT FUEL POOL CIRC PUMP AC-5B DISCHARGE VALVE and PLACED in a throttled position less than full open.

• OBSERVED valve handle at 45° from piping.

Examiner Cue: Valve handle is 45° offset from piping (or throttled appropriately).

Comment: SAT UNSAT CAUTION If the Discharge Valve for the non-running Spent Fuel Pool Circ Pump is Open, the pump will windmill backwards. The discharge valve must be Closed while starting the pump to prevent excessive starting current.

Perform Step: 4 Close the non-selected SFP pump discharge valve:

9

• AC-192, AC-5A Standard: PERFORMED the following:

• SQUEEZED handle on AC-192, SPENT FUEL POOL CIRC PUMP AC-5A DISCHARGE VALVE and PLACED in CLOSED position then RELEASED (critical).

• OBSERVED valve handle perpendicular to piping (NOT critical).

Examiner Cue: Valve handle is perpendicular to piping.

Comment: SAT UNSAT Page 4 of 6 NRC JPM P-1 Rev. Final

Appendix C JPM STEPS Form ES-C-1 Examiner Note: START pushbutton is located on wall behind pump.

Perform Step: 5 Start the selected SFP pump (Room 5):

10

• AC-5B, "SPENT FUEL POOL COOLING PUMP" Standard: PERFORMED the following:

• DEPRESSED black START pushbutton for AC-5B, SPENT FUEL POOL COOLING PUMP (critical).

• OBSERVED red START light lit and GREEN stop light off (NOT critical).

Examiner Cue: RED light is lit and GREEN light is off. Noise emanating from pump.

Comment: SAT UNSAT Perform Step: 6 Open the selected SFP pump discharge valve:

11

• AC-191 Standard: PERFORMED the following:

• SQUEEZED handle on AC-191, SPENT FUEL POOL CIRC PUMP AC-5B DISCHARGE VALVE and PLACED in OPEN position then RELEASED (critical).

• OBSERVED valve handle parallel with piping (NOT critical).

Examiner Cue: Valve handwheel is parallel with piping.

Comment: SAT UNSAT Examiner Note: Pressure gauge ranges from 0 to 300 psig.

Perform Step: 7 Verify SFP pump discharge pressure 40-60 psig.

12 Standard: OBSERVED pressure gauge on discharge of AC-5B between 40 psig and 60 psig.

Terminating Cue: Pressure gauge needle positioned 1/5th upscale. Another operator will throttle CCW flow to minimize dose rate. This JPM is complete.

Comment: SAT UNSAT STOP TIME:

Page 5 of 6 NRC JPM P-1 Rev. Final

Appendix C JPM CUE SHEET Form ES-C-1 INITIAL CONDITIONS: Given the following conditions:

• AOP-36, Loss of Spent Fuel Pool Cooling, has been entered following a Safety Injection Actuation Signal.

• SIAS and CIAS have been RESET.

• 480 Volt Bus 1B4C is energized and available for Spent Fuel Pool Cooling Pump AC-5B.

• HC-478, STORAGE POOL HX AC-8 AC OUTL HCV-478 at AI-45 is OPEN.

INITIATING CUE: The Control Room Supervisor directs you to PERFORM the following:

• RESTORE Spent Fuel Pool Cooling using Spent Fuel Pool Cooling Pump AC-5B per AOP-36, Loss of Spent Fuel Pool Cooling, Attachment H, Spent Fuel Pool Cooling Restoration with SIAS.

• START at Step 6.

Page 6 of 6 NRC JPM P-1 Rev. Final

Appendix C JPM WORKSHEET Form ES-C-1 Facility: FCS JPM # NRC P-2 Task # 0809 K/A #061.A2.05 3.1 / 3.4 SF-4S

Title:

Locally Start FW-54, Diesel Driven AFW Pump Examinee (Print):

Testing Method:

Simulated Performance: X Classroom:

Actual Performance: Simulator:

Alternate Path: X Plant: X Time Critical:

READ TO THE EXAMINEE I will explain the Initial Conditions, which steps to simulate or discuss, and provide an Initiating Cue.

When you complete the task successfully, the objective for this JPM will be satisfied.

Initial Conditions: Given the following conditions:

• EOP-06, Loss of All Feedwater, is in progress.

• FW-54, Diesel Driven Auxiliary Feedwater Pump, failed to start from the Control Room.

• Feedwater flow CANNOT be aligned to the Feed Ring.

• Control Room has provided AI-114, FW-54 Control Panel keys.

Initiating Cue: The Control Room Supervisor directs you to PERFORM the following:

• Locally START FW-54, Diesel Driven AFW Pump per EOP/AOP Attachments-HR Heat Removal, HR-16, FW-54 Operation.

• ESTABLISH Auxiliary Feedwater flow.

Task Standard: Utilizing HR-16, started FW-54 then deenergized and opened HCV-1384.

Required Materials: EOP/AOP Attachments-HR Heat Removal, Rev. 1.

Validation Time: 15 minutes Completion Time: ________ minutes Comments:

Result: SAT UNSAT Examiner (Print / Sign): Date:

Page 1 of 6 NRC JPM P-2 Rev. Final

Appendix C JPM WORKSHEET Form ES-C-1 PLANT SETUP EXAMINER:

PROVIDE the examinee with a copy of:

• EOP/AOP Attachments-HR Heat Removal.

• HR-16, FW-54 Operation.

Page 2 of 6 NRC JPM P-2 Rev. Final

Appendix C JPM STEPS Form ES-C-1

- Check Mark Denotes Critical Step START TIME:

Examiner Note: The following steps are from EOP/AOP Attachments, HR-16.

Examiner Note: FW-54 Room is located in Southeast corner of Turbine Building basement.

Perform Step: 1 Start FW-54, Diesel AFW Pump, by performing step a or b.

1, 1.a, & 1.b

• Start FW-54 from the Control Room by placing HC/FW-54, "AFW PUMP FW-54" in "START".

• (LOCAL) Start FW-54 by performing the following:

Standard: DETERMINED from Initial Conditions that Local Start of FW-54 is required.

Comment: SAT UNSAT Perform Step: 2 Obtain AI-114, FW-54 Control Panel keys, from one of the following:

b.1)

• EONT key ring

• Control Room Standard: DETERMINED from Initial Conditions that Control Room has provided AI-114, FW-54 Control Panel keys.

Comment: SAT UNSAT Perform Step: 3 Using key, place HC/FW-54-1, "LOCAL CONTROL SWITCH" in b.2) "STOP".

Standard: INSERTED key into 2 position switch and TURNED HC/FW-54-1, FW-54 LOCAL CONTROL SWITCH to STOP (left) position.

Examiner Cue: Key is in STOP position.

Comment: SAT UNSAT Perform Step: 4 Using key, place 43/FW-54, "CONTROL TRANSFER SWITCH" in b.3) "RESET".

Standard: INSERTED key into 3 position switch and TURNED 43/FW-54, FW-54 CONTROL TRANSFER SWITCH to RESET (center) position.

Examiner Cue: Key is in RESET position.

Comment: SAT UNSAT Page 3 of 6 NRC JPM P-2 Rev. Final

Appendix C JPM STEPS Form ES-C-1 Perform Step: 5 Using key, place "43/FW-54" in "LOCAL".

b.4)

Standard: INSERTED key into 3 position switch and TURNED 43/FW-54, FW-54 CONTROL TRANSFER SWITCH to LOCAL (right) position.

Examiner Cue: Key is in LOCAL position.

Comment: SAT UNSAT Perform Step: 6 Using key, place "HC/FW-54-1" in "RUN".

b.5)

Standard: PERFORMED the following:

• INSERTED key into 2 position switch and TURNED HC/FW-54-1, FW-54 LOCAL CONTROL SWITCH to RUN (right) position.

(critical).

• OBSERVED engine speed rising (NOT critical).

• OBSERVED engine noise rising (NOT critical).

• OBSERVED Diesel Room area for leaks (NOT critical).

Examiner Cue: Key is in RUN position. Engine noise and speed are rising.

Comment: SAT UNSAT Perform Step: 7 Feed through the Feed Ring by performing the following:

2 & 2.1 CA

• IF the Feed Ring is NOT available, THEN GO TO Step 0.

Standard: DETERMINED Feed Ring is NOT available per Initial Conditions and TRANSITIONED to Step 4.

Comment: SAT UNSAT Examiner Note: The following steps represent the Alternate Path of this JPM.

Perform Step: 8 Feed through the AFW Nozzles by performing the following:

4 & 4.a

• Open HCV-1384, FW/AFW Header Cross-Connect Valve.

Standard: CONTACTED Control Room to open HCV-1384, FW/AFW Header Cross-Connect Valve.

Examiner Cue: Control Room reports HCV-1384, FW/AFW Header Cross-Connect Valve will NOT open.

Comment: SAT UNSAT Page 4 of 6 NRC JPM P-2 Rev. Final

Appendix C JPM STEPS Form ES-C-1 Examiner Note: West Upper Electrical Penetration Room is located in the Auxiliary Building directly below Control Room.

Perform Step: 9 (LOCAL) IF valve will NOT open, THEN perform the following:

4.a.1 CA & 4.a.1.1) CA

• Place Breaker MCC-4C1-E03, "FW AND AUX FEED WATER CROSS CONNECTION VALVE" in "OFF" (West Upper Electrical Penetration Room).

Standard: ROTATED switch for MCC-4C1-E03, HCV-1384, FW AND AUX FEEDWATER CROSSCONNECTION VALVE breaker COUNTERCLOCKWISE to OFF position.

Examiner Cue: Breaker switch rotated then stopped.

Comment: SAT UNSAT Examiner Note: Room 81 is located on 1036' elev. of the Auxiliary Building and is accessed from the Turbine Deck.

Perform Step: 10 Manually open HCV-1384 (Room 81).

4.a.1.2) CA Standard: DEPRESSED clutch arm and ROTATED handwheel for HCV-1384, FW-AFW MAIN AND AUXILIARY FEEDWATER CROSSCONNECT VALVE in COUNTERCLOCKWISE direction until stopped.

Terminating Cue: Valve handwheel rotated then stopped. This JPM is complete.

Comment: SAT UNSAT STOP TIME:

Page 5 of 6 NRC JPM P-2 Rev. Final

Appendix C JPM CUE SHEET Form ES-C-1 INITIAL CONDITIONS: Given the following conditions:

• EOP-06, Loss of All Feedwater, is in progress.

• FW-54, Diesel Driven Auxiliary Feedwater Pump, failed to start from the Control Room.

• Feedwater flow CANNOT be aligned to the Feed Ring.

• Control Room has provided AI-114, FW-54 Control Panel keys.

INITIATING CUE: The Control Room Supervisor directs you to PERFORM the following:

• Locally START FW-54, Diesel Driven AFW Pump per EOP/AOP Attachments-HR Heat Removal, HR-16, FW-54 Operation.

• ESTABLISH Auxiliary Feedwater flow.

Page 6 of 6 NRC JPM P-2 Rev. Final

Appendix C JPM WORKSHEET Form ES-C-1 Facility: FCS JPM # NRC P-3 Task # 0735 K/A # 071 G2.1.30 4.4 / 4.0 SF-9

Title:

Terminate Release of Radioactive Gas Examinee (Print):

Testing Method:

Simulated Performance: X Classroom:

Actual Performance: Simulator:

Alternate Path: Plant: X Time Critical:

READ TO THE EXAMINEE I will explain the Initial Conditions, which steps to simulate or discuss, and provide an Initiating Cue.

When you complete the task successfully, the objective for this JPM will be satisfied.

Initial Conditions: Given the following conditions:

• A release from WD-29A, Waste Gas Decay Tank is in progress.

• RM-052, Auxiliary Building Ventilation Stack Radiation Monitor has gone into HIGH alarm.

• AOP-09, High Radioactivity is in progress.

• When HC-532, Waste Gas Release Control Switch was placed in CLOSE, FCV-532A and FCV-532C on AI-100 did NOT close.

• Independent Verification has been waived by the Shift Manager due to AOP-09 entry and high radiation in the Auxiliary Building.

Initiating Cue: The Control Room Supervisor directs you to PERFORM the following:

• TERMINATE release from WD-29A per OI-WDG-2, Waste Gas Disposal System Release, Attachment 3, Manual Waste Gas Release with FE-532 Unavailable.

• START at Step 2.17.

Task Standard: Utilizing OI-WDG-2, closed WD-158, isolated WD-29A via WD-132 terminating release from Waste Gas Decay Tank WG-29A.

Required Materials: AOP-09, High Radioactivity, Rev. 11.

OI-WDG-2, Waste Gas Disposal System Release, Rev. 30.

Validation Time: 15 minutes Completion Time: ________ minutes Comments:

Result: SAT UNSAT Examiner (Print / Sign): Date:

Page 1 of 6 NRC JPM P-3 Rev. Final

Appendix C JPM WORKSHEET Form ES-C-1 PLANT SETUP EXAMINER:

PROVIDE the examinee with a copy of:

• OI-WDG-2, Waste Gas Disposal System Release.

• Attachment 3, Manual Waste Gas Release with FE-532 Unavailable.

• Attachment 3 is INITIALED through Step 2.16.

Page 2 of 6 NRC JPM P-3 Rev. Final

Appendix C JPM STEPS Form ES-C-1

- Check Mark Denotes Critical Step START TIME:

Examiner Note: The following steps are from OI-WDG-2, Attachment 3.

Examiner Note: Panel AI-100 and Room 16 are adjacent to each other and located in Corridor 4, 989' elev. of Auxiliary Building.

Perform Step: 1 Verify the following Gas Release Control Valves closed:

2.17

• FCV-532A (AI-100)

• FCV-532C (AI-100)

• FCV-532B (Room 16)

Standard: NOTED the following:

• DETERMINED FCV-532A and FCV-532C did NOT close from Initial Conditions, or

• OBSERVED red OPEN lights lit and green CLOSE lights off at AI-100, and

• OBSERVED FCV-532B open in Room 16.

Examiner Cue: Red lights are lit on AI-100 for FCV-532A & FCV-532C. In Room 16, FCV-532B indicates mid position (between open and close position discs).

Comment: SAT UNSAT Examiner Note: All valves are located on the East wall of Room 16.

Perform Step: 2 Close WD-158.

2.18 Standard: ROTATED WD-158, WASTE GAS RELEASE HEADER FLOW ELEMENT FE-532 BYPASS LINE ISOLATION VALVE handwheel in CLOCKWISE direction until stopped.

Examiner Cue: Valve handwheel rotated then stopped.

Comment: SAT UNSAT Page 3 of 6 NRC JPM P-3 Rev. Final

Appendix C JPM STEPS Form ES-C-1 Examiner Note: Valve located about 6 feet above the floor.

Perform Step: 3 Close the selected WGDT Outlet to Gas Release Header Valve (Rm 16):

2.19

• WD-132, WD-29A Standard: ROTATED WD-132, GAS DECAY TANK WD-29A OUTLET VALVE handwheel in CLOCKWISE direction until stopped.

Examiner Cue: Valve handwheel rotated then stopped. If asked, REPORT as Control Room Supervisor to continue in procedure.

Comment: SAT UNSAT Perform Step: 4 Close and lock the following Gas Release Header Isolation Valves 2.20 & 1st bullet (Rm 16):

• WD-150 Standard: PERFORMED the following:

• ROTATED WD-150, WASTE GAS DECAY TANKS WD-29A, B, C

& D GAS RELEASE HEADER ISOLATION VALVE handwheel in CLOCKWISE direction until stopped (critical).

• INSTALLED chain and LOCKED valve (NOT critical).

• INFORMED Control Room WD-150 is LOCKED per SO-O-44 (NOT critical).

Examiner Cue: Valve handwheel rotated then stopped.

If Control Room is contacted, ACKNOWLEDGE locking of WD-150.

Comment: SAT UNSAT Examiner Note: Valve located > 8 feet above the floor. May require a ladder that can be obtained from the Corridor (West) just beyond AI-100, and notification to RP that they are working above 7 ft.

Perform Step: 5 Close and lock the following Gas Release Header Isolation Valves 2.20 & 2nd bullet (Rm 16):

• WD-167 Standard: PERFORMED the following:

• If needed, OBTAINED a ladder, simulated notifying RP

• ROTATED WD-167, WASTE GAS DECAY TANKS WD-29A, B, C

& D GAS RELEASE HEADER ISOLATION VALVE handwheel in CLOCKWISE direction until stopped.

Examiner Cue: Valve handwheel rotated then stopped.

Comment: SAT UNSAT Page 4 of 6 NRC JPM P-3 Rev. Final

Appendix C JPM STEPS Form ES-C-1 Examiner Note: Valve located at floor level.

Perform Step: 6 Close WD-165, Gas Release Header Bypass Valve (Rm 16).

2.21 Standard: ROTATED WD-165, GAS RELEASE HEADER BYPASS VALVE handwheel in CLOCKWISE direction until stopped.

Terminating Cue: Valve handwheel rotated then stopped. This JPM is complete.

Comment: SAT UNSAT STOP TIME:

Page 5 of 6 NRC JPM P-3 Rev. Final

Appendix C JPM CUE SHEET Form ES-C-1 INITIAL CONDITIONS: Given the following conditions:

• A release from WD-29A, Waste Gas Decay Tank is in progress.

• RM-052, Auxiliary Building Ventilation Stack Radiation Monitor has gone into HIGH alarm.

• AOP-09, High Radioactivity is in progress.

• When HC-532, Waste Gas Release Control Switch was placed in CLOSE, FCV-532A and FCV-532C on AI-100 did NOT close.

• Independent Verification has been waived by the Shift Manager due to AOP-09 entry and high radiation in the Auxiliary Building.

INITIATING CUE: The Control Room Supervisor directs you to PERFORM the following:

• TERMINATE release from WD-29A per OI-WDG-2, Waste Gas Disposal System Release, Attachment 3, Manual Waste Gas Release with FE-532 Unavailable.

• START at Step 2.17.

Page 6 of 6 NRC JPM P-3 Rev. Final

Appendix D Scenario Outline Form ES-D-1 Facility: Fort Calhoun Station Scenario No.: 1 Op Test No.: Dec 2015 NRC Examiners: Operators:

Initial Conditions: 100% power MOL - RCS Boron is 482 ppm (by sample).

Turnover: Maintain steady-state power conditions. Chemistry requests two Charging Pumps be placed in service per OI-CH-1, CVCS System Normal Operation.

Critical Tasks:

• Manually Actuate Pressurizer Pressure Low Signal (PPLS) when RCS Pressure 1600 psia and before 1350 psia to ensure SIAS / VIAS / CIAS Activation. (Event 8)

• Stop All Reactor Coolant Pumps (RCPs) when Subcooling is approaching or is less than 20°F but before 0°F due to Loss of Net Positive Suction Head (NPSH) per RCP NPSH Curve. (Event 6)

• Commence a Cooldown and Depressurization of the Reactor Coolant System before Reactor Vessel Level Monitoring System (RVLMS) is less than 83%, indicating a bubble has formed in the head, to Reestablish RCS Inventory Control while maintaining RCS Heat Removal.

(Event 6)

Event No. Malf. No. Event Type* Event Description 1 N (ATCO) Raise Charging and Letdown Flow per OI-CH-1, CVCS System

+15 min Normal Operation, Attachment 3.

2 C (ATCO, CRS) Component Cooling Water (CCW) Pump Trip.

+25 min TS (CRS) Start Either Standby CCW Pump.

3 C (BOPO, CRS) Plant Air System Leak @ 15% on 2 minute ramp.

+35 min Start Instrument Air Compressors.

4 I (ATCO, CRS) Pressurizer Pressure Control Channel PT-103X Fails to 2150 psia

+45 min TS (CRS) on 15 Minute Ramp. Transfer Pressure Control to PT-103Y.

5 R (ATCO) Condenser Evacuation Pump Trip with Auto Start Failure.

+55 min C (BOPO, CRS) Partial Loss of Condenser Vacuum. Reduce Turbine Load.

6 M (ATCO, BOPO, Inadvertent Main Turbine Trip.

+55 min CRS) Pressurizer Safety Valve Fails 50% Open on Reactor Trip.

7 C (BOPO) Total Loss of Condenser Vacuum.

+55 min Place HCV-1040, Atmospheric Dump Valve in Service.

8 I (ATCO) Pressurizer Pressure Low Signal Actuation Failure.

+65 min Manually Initiate Safety Injection.

9 C (ATCO) Low Pressure Safety Injection (LPSI) Pumps Start Failure.

+65 min Manually Start LPSI Pumps.

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Technical Specifications Actual Target Quantitative Attributes 3 Malfunctions after EOP entry (1-2) 5 Abnormal events (2-4) 1 Major transients (1-2) 1 EOPs entered/requiring substantive actions (1-2) 0 EOP contingencies requiring substantive actions (0-2) 3 Critical tasks (2-3)

NRC Simulator Scenario 1 Outline Rev. Final

Scenario Event Description NRC Scenario 1 SCENARIO

SUMMARY

NRC 1 The crew will assume the shift at 100% power per OP-4, Load Change and Normal Power Operation.

The scheduled activity is to start a second Charging Pump per OI-CH-1, CVCS System Normal Operation, Attachment 3, Raising Charging and Letdown Flows per Chemistry request.

The next event is a Component Cooling Water Pump Trip with auto start failure of the standby pumps.

The crew enters AOP-11, Loss of Component Cooling Water, and restores flow by starting either CCW Pump AC-3A or AC-3B. The SRO will refer to Technical Specification LCO 2.4(1) - Component Cooling Water Pump.

The next event is a Plant Air System leak and entry into AOP-17, Loss of Instrument Air, is required.

Crew should recognize that the Control Room Standby Instrument Air Compressor is not loading (ammeter at 0) and start a 3rd Air Compressor. Procedure exit occurs when the Plant Air System is locally isolated from the Instrument Air System.

When plant conditions are stable, Pressurizer Pressure Control Channel, PT-103X, will fail to 2150 psia over 15 minutes. Operator actions are per ARP-CB-1/2/3/A4, Windows A-4 & B-4, PRESSURIZER PRESSURE OFF NORMAL HI-LO CHANNEL X & CHANNEL Y. The crew will transfer to the standby channel PT-103Y and restore Reactor Coolant System (RCS) pressure. The SRO will refer to Technical Specification LCO 2.10.4 - DNBR Margin during Power Operation above 15% of Rated Power.

The next event is a partial Loss of Condenser Vacuum. The crew enters AOP-26, Turbine Malfunctions,Section I, Loss of Vacuum. Actions include starting a Condenser Evacuation Pump and transitioning to AOP-05, Emergency Shutdown, to lower Turbine load and restore Condenser vacuum. When power has been reduced 3% to 5%, an inadvertent Main Turbine trip will occur.

The inadvertent Main Turbine trip results in lifting of a Pressurizer Safety Valve resulting in a Small Break Loss of Coolant Accident (Vapor Space LOCA). The crew enters EOP-00, Standard Post Trip Actions, and manually actuates Safety Injection when it is determined that a Pressurizer Pressure Low Signal Actuation failure has occurred. When Diagnostic Actions are completed at the end of EOP-00, a transition will be made to EOP-03, Loss of Coolant Accident. Two Reactor Coolant Pumps are secured while in EOP-00 when pressure drops to 1350 psia. Eventually all RCPs will be secured due to a loss of subcooling (< 20°F). Upon entry into EOP-03, Containment Cooling Fans VA-7C and VA-7D will need to be started. Containment pressure remains less than 3 psig throughout the event.

The event is complicated by total Loss of Condenser Vacuum which will require placing the Atmospheric Steam Dump Valve, HCV-1040 in service and manual starting of the Low Pressure Safety Injection Pumps due to an automatic start failure.

This scenario is terminated when a cooldown and depressurization is commenced while in EOP-03 using HR-12, Secondary Heat Removal Operation, and PC-11, Pressure Control.

Risk Significance:

• Failure of risk important system prior to trip: Loss of CCW Pump

• Risk significant core damage sequence: Small Break LOCA Safety Injection Actuation Failure

• Risk significant operator actions: Manually Actuate Safety Injection Stop RCPs Upon Loss of Subcooling Cooldown and Depressurize RCS NRC Simulator Scenario 1 Outline Rev. Final

Scenario Event Description NRC Scenario 1 BOOTH OPERATOR INSTRUCTIONS for SIMULATOR SETUP RESET to IC-108 and LOAD & EXECUTE NRC 1.sce for NRC Scenario 1.

Preset Item - Event 2 - Block Autostart of Non-running CCW Pumps Type Item Value Condition Expert CCAAFU_STDBY_AC_3BCC 1 Scenario Event: AC-3B (AC-3B standby fuse failure) Stbyfuse blown CCBPFU_STDBY_AC_3ACC 1 Scenario Event: AC-3A (AC-3A standby fuse failure) Stby Fuse blown Preset Item - Event 3 - Block Autostart of CA-1B Type Item Value Condition Remote REM:CA1B_3SS (CA-1B control Off (value = 3) Scenario Event: Block start selector switch) of CA-1B Preset Item - Event 5 - Block Auto Start of Condenser Evacuation Pump FW-8C Type Item Value Condition Expert CEACWL_CLTVSP Triggered Scenario Event: block start FW-8C Preset Item - Event 8 - PPLS Fail to Actuate Type Item Value Condition Malfunction ESF07 (PPLS Actuation - Train A) Block Scenario Event: PPLS auto ESF08 (PPLS Actuation - Train B) Block fail Preset Item - Event 9 - LPSI Pumps Fail to Automatically Start Type Item Value Condition Expert ESEARL62_2_1X_SI_1BTVSP Deenergized Scenario Event: LPSI fail ESEBRL62_2_2X_SI_1BTVSP Deenergized to start ESCBRL62_1_2X_SI_1ATVSP Deenergized ESCARL62_1_1X_SI_1ATVSP Deenergized Event 2 - CCW Pump AC-3C Trips Type Item Value Condition Malfunction BUS_1B3C_4C_4_BKR_TRIP trip When directed by examiner, (CCW pump AC-3C breaker fail to trigger/activate this event.

the trip position) Scenario Event: CCW Pump AC-3C Trip Event 3 - Plant Air Leak Type Item Value Condition Malfunction CAS02C (Plant Air Leak) 15 When directed by examiner, Ramp: 120 seconds trigger/activate this event.

Scenario Event: Plant Air Leak Remote REM:CAS_CA630 0 When directed to close CA-REM:CAS_PCV1753 0 121 to isolate the instrument air leak, trigger/activate this event. Scenario Event:

When directed to close CA121 NRC Simulator Scenario 1 Outline Rev. Final

Scenario Event Description NRC Scenario 1 Event 4 - Pressurizer Pressure Transmitter PT-103X Fails High Type Item Value Condition Transmitter RCS_PT103X 2150 When directed by examiner, Ramp: 900 seconds trigger/activate this event.

Scenario Event: PT-103x fail high Event 5 - Running Condenser Evacuation Pump Trips, Degrading Condenser Vacuum Type Item Value Condition Malfunction CES06 (Condenser Evacuation FW- Trip When directed by examiner, 8B Pump trips) trigger/activate this event.

CND01 (Loss of Main Condenser 3%, ramp = 60 sec Scenario Event: Cond Vacuum) Evac trip Event 6 - Inadvertent Trip, Pressurizer Safety Valve Opens Type Item Value Condition Remote REM:86-1/G1-TRP (relay 86-1/G1 Trip When directed by examiner, fail to trip position) trigger/activate this event.

REM: 86-2/G1-TRP (relay 86-2/G1 Trip Scenario Event: Trip, fail to trip position) safety valve open Malfunction RCS_RC141 (safety valve RC-141) After reactor trip, value = 50, ramp =

15 seconds, delay =

5 seconds Event 7 - Total Loss of Condenser Vacuum Type Item Value Condition Malfunction CND01 (Loss of Main Condenser 100%, 300 second 60 seconds after reactor trip, Vacuum) ramp automatically trigger/activate event:

Complete Loss of Cond Vacuum NRC Simulator Scenario 1 Outline Rev. Final

Scenario Event Description NRC Scenario 1 Booth Operator: INITIALIZE to IC-108 and LOAD NRC 1.sce.

ENSURE all Simulator Annunciator Alarms are ACTIVE.

ENSURE all Control Board Tags are removed.

ENSURE CH-1C, Charging Pump is running.

ENSURE AC-3C, Component Cooling Water Pump running.

ENSURE Channel X Pressurizer Pressure and Level selected.

ENSURE FW-8B, Condenser Evacuation Pump running.

ENSURE Reactivity Briefing Sheet printout provided with Turnover.

ENSURE Middle-of-Life Thumb Rule Sheet provided with Turnover.

ENSURE Containment Pressure Relief (CPR) is secured.

ENSURE procedures in progress provided to crew in Briefing Room:

- COPY of OI-CH-1, Chemical and Volume Control System Normal Operation, Attachment 3, Raising Charging and Letdown Flows, INITIALED through Step 2.i.

Control Room Annunciators in Alarm:

NONE 0B Procedure List Event 1: OP-4, Load Change and Normal Power Operation Event 1: OI-CH-1, CVCS System Normal Operation, Attachment 3, Raising Charging and Letdown Flows Event 2: AOP-11, Loss of Component Cooling Water Event 3: AOP-17, Loss of Instrument Air Event 4: ARP-CB-1/2/3/A4, Windows A-4 & B-4, PRESSURIZER PRESSURE OFF NORMAL HI-LO CHANNEL X & CHANNEL Y Event 5: AOP-26, Turbine Malfunctions,Section I, Loss of Vacuum Event 6: EOP-00, Standard Post Trip Actions Event 6: EOP-03, Loss of Coolant Accident Event 6: HR-12, Secondary Heat Removal Operation Event 6: PC-11, Pressure Control NRC Simulator Scenario 1 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 1 Page 6 of 29 Event

Description:

Raise Charging and Letdown Flow Time Position Applicants Actions or Behavior Booth Operator: When directed, RESPOND to requests from Control Room. REPORT back that plant conditions requested are normal unless otherwise scripted.

Indications Available:

NONE Examiner Note: The following steps are from OI-CH-1, Chemical and Volume Control System Normal Operation, Attachment 3, Raising Charging and Letdown Flows.

+1 min ATCO START the selected Charging Pump CH-1B. [Step 3]

• PLACE CH-1B switch to START.

NOTES

1. PIC-210 Letdown Press Cntrlr should be continuously monitored while adjusting letdown flow.

2. Steps 4 and 5 may be performed concurrently without the procedure in hand. Sign-offs may be completed after these steps are performed.

RAISE bias on HIC-101-1/101-2, Letdown Throttle Valves Controller, and ATCO OBSERVE an increase in Letdown flow. [Step 4]

• ROTATE HIC-101-1/101-2 in COUNTERCLOCKWISE direction to increase Letdown flow.

Examiner Note: It is acceptable to place Letdown Pressure Control and Flow Control in MANUAL or AUTOMATIC control during rotation of Charging Pumps.

ADJUST PIC-210, Letdown Press Controller as necessary to maintain ATCO Letdown pressure approximately 300 psig. [Step 5]

Continue to ADJUST bias on HIC-101-1/101-2 until Pressurizer level is ATCO STABILIZED at the programmed setpoint. [Step 6]

When Letdown flow is stable, PROCEED to Event 2.

NRC Simulator Scenario 1 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 2 Page 7 of 29 Event

Description:

Component Cooling Water Pump Trip Time Position Applicants Actions or Behavior Booth Operator: When directed, EXECUTE Event 2.

- AC-3C, CCW Pump trip.

Indications Available:

CB-1/2/3/A2 - CCW PUMPS TRIP CB-1/2/3/A2 - CC WATER FROM DISCH HEADER FLOW LO CB-1/2/3/A2 - CCW PUMPS AC-3A/B/C STANDBY START CB-1/2/3/A2 - AUXILIARY COOLANT FROM CRDM FLOW LO CCW Pump AC-3C white TRIP and green STOP lights lit Multiple loss of CCW flow alarms Booth Operator: When contacted for pump conditions, REPORT as Auxiliary Building Operator all conditions normal. REPORT as Water Plant Operator that breaker tripped on overcurrent.

+30 sec ATCO RESPOND to Annunciator Response Procedures.

ATCO INFORM CRS of CCW Pump AC-3C trip with NO auto start of standby pump.

Examiner Note: ATCO may Operate to Mitigate per OPD 3-01 and START a CCW Pump.

CRS REFER to AOP-11, Loss of Component Cooling Water.

Examiner Note: The following steps are from AOP-11, Loss of Component Cooling Water.

ATCO VERIFY normal CCW/RW System operation: [Step 4.1]

• START CCW Pump AC-3A or AC-3B. [Step 4.1.a]

• VERIFY CCW System pressure 60 psig. [Step 4.1.b]

• DETERMINE AC-1B, Raw Water CCW Heat Exchanger in service.

[Step 4.1.c]

• DETERMINE RCP Coolers CCW Valves, HCV-438A/B/C/D all OPEN.

[Step 4.1.d]

ATCO VERIFY Raw Water Pump operating. [Step 4.2]

ATCO If CCW Surge Tank level < 42 inches, FILL the CCW Surge Tank: [Step 4.3]

NRC Simulator Scenario 1 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 2 Page 8 of 29 Event

Description:

Component Cooling Water Pump Trip Time Position Applicants Actions or Behavior

• OPEN LCV-2801, CCW Surge Tank Makeup Valve, to refill CCW Surge Tank. [Step 4.3.a]

• PLACE LCV-2801 in CLOSE or AUTO. [Step 4.3.b]

CRS EVALUATE Technical Specification LCO 2.4, Containment Cooling

• LCO 2.4.(1).a - Component Cooling Water Pump AC-3C

• CONDITION 2.4.(1).a - Component Cooling Water Pump AC-3C inoperable.

• ACTION 2.4.(1).b - RESTORE Component Cooling Water Pump AC-3C within 7 days OR PLACE Reactor in HOT SHUTDOWN condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

When Technical Specifications are addressed, PROCEED to Event 3.

NRC Simulator Scenario 1 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 3 Page 9 of 29 Event

Description:

Plant Air System Leak Time Position Applicants Actions or Behavior Booth Operator: When directed, EXECUTE Event 3.

- Plant Air System leak @ 15% on 2 minute ramp.

Indications Available:

CB-10,11/A21 - PLANT AIR PRESS LO PI-1700, Plant Air Press lowering on CB-10,11

+30 sec BOPO RESPOND to Annunciator Response Procedures.

BOPO INFORM CRS of Plant Air System pressure less than 96 psig and lowering.

Examiner Note: BOPO may Operate to Mitigate per OPD 3-01 and START an Air Compressor.

CRS REFER to AOP-17, Loss of Instrument Air.

Examiner Note: The following steps are from AOP-17, Loss of Instrument Air.

BOPO ENSURE all available Air Compressors start. [Step 4.1]

• START Air Compressor CA-1A.

• CONTACT Auxiliary Operator to START Air Compressor CA-1B.

Examiner Note: Air Compressor CA-1B does NOT Auto Start from the Control Room. Control Board indications for CA-1B show the breaker closed but compressor is NOT running or loaded (observe CA-1B amps).

Booth Operator: If contacted, REPORT standby Air Compressor CA-1B switch alignment is normal.

CONTACT Equipment Operator to ensure proper operation of Instrument Air BOPO Compressors, Dryers, and Filters. [Step 4.2]

Booth Operator: If contacted, REPORT Compressors, Dryers, and Filters appear to be operating normally.

ANNOUNCE and REPEAT message using Plant Communication System:

CREW

[Step 4.3]

NRC Simulator Scenario 1 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 3 Page 10 of 29 Event

Description:

Plant Air System Leak Time Position Applicants Actions or Behavior

• "Attention all personnel, attention all personnel; there is a plant air leak in progress. Report any large air usage to the Control Room."

CRS DIRECT available operators to search for source of air leakage. [Step 4.4]

Booth Operator: When contacted, REPORT leak is downstream of PCV-1753. When directed, EXECUTE remote function to isolate leak and report CA-121, Service Air Supply System Manual Isolation Valve is CLOSED.

DETERMINE Instrument Air pressure is < 80 psig, and CONTACT BOPO Equipment Operator to VERIFY PCV-1753, Service Air System Automatic Isolation Valve CLOSED. [Step 4.5]

DETERMINE Instrument Air pressure slowly returning to normal after service CRS air was isolated. [Step 4.6]

• VERIFY CA-121, Service Air Supply System Manual Isolation Valve is closed. [Step 4.6.a]

• GO TO Section 5.0, Exit Conditions. [Step 4.6.b]

Examiner Note: Plant Air System remains isolated for the duration of the Scenario.

When Instrument Air pressure returns to normal, PROCEED to Event 4.

NRC Simulator Scenario 1 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 4 Page 11 of 29 Event

Description:

Pressurizer Pressure Control Channel Failure Time Position Applicants Actions or Behavior Booth Operator: When directed, EXECUTE Event 4.

- Pressurizer Pressure Control Channel PT-103X fails to 2150 psia on 15 minute ramp.

Indications Available:

CB-1/2/3/A4 - PRESSURIZER PRESSURE OFF NORMAL HI-LO CHANNEL Y (1st alarm)

CB-1/2/3/A4 - PRESSURIZER PRESSURE OFF NORMAL HI-LO CHANNEL X (2nd alarm ~ 2 min later)

Examiner Note: Due to the nature of this failure, Channel Y alarm comes in 1st as it senses PZR pressure < 2080 psia (alarm setpoint) even though Channel X is the Controlling Channel. As the Channel X setpoint failure ramps in and reaches

> 2145 psia (alarm setpoint), Channel X annunciator will alarm.

+30 sec ATCO RESPOND to Annunciator Response Procedures.

Examiner Note: ATCO may Operate to Mitigate per OPD 4-09 and TRANSFER to Channel Y.

Examiner Note: The following steps are from ARP-CB-1/2/3/A4, Window A-4 for Channel X.

ATCO VERIFY RCS pressure using all available indications. [Step 1]

• MONITOR Pressurizer Pressure and operation of PC-103X. [Step 1.1]

• DETERMINE PC-103X is not controlling pressure and PLACE HC-103, Pressurizer Pressure Channel Selector Switch to CHAN Y position. [Step 1.1.1]

• MAY REFER to OI-RC-7, Reactor Coolant System Pressure Control CRS Normal Operation. [Step 1.1.2]

ATCO PERFORM the following for the low pressure condition: [Step 2]

• REFER to Technical Specification LCO 2.10.4.(5) if pressure 2075 CRS psia. [Step 2.1]

• DETERMINE Pressurizer Spray Valves PCV-103-1 and PCV-103-2 are ATCO CLOSED. [Step 2.2]

• ENSURE all Pressurizer Heater Control Switches in AUTO or ON.

ATCO

[Step 2.3]

ATCO

• ENERGIZE additional Pressurizer Heaters as required. [Step 2.4]

• DETERMINE Pressurizer level NOT lowering on LR-101X/LR-101Y.

ATCO

[Step 2.5]

NRC Simulator Scenario 1 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 4 Page 12 of 29 Event

Description:

Pressurizer Pressure Control Channel Failure Time Position Applicants Actions or Behavior ATCO

• VERIFY VCT level trend on LI-219. [Step 2.6]

CRS EVALUATE Technical Specification LCO 2.10.4, Power Distribution Limits

• LCO 2.10.4.(5) - DNBR Margin during Power Operation above 15% of Rated Power

• CONDITION 2.10.4.(5).(a).(ii) - Pressurizer Pressure < 2075 psia.

• ACTION 2.10.4.(5).(b) - RESTORE Pressurizer Pressure within its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or REDUCE power to less than 15% of rated power within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

When Technical Specifications have been addressed, PROCEED to Event 5.

NRC Simulator Scenario 1 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 5 Page 13 of 29 Event

Description:

Partial Loss of Condenser Vacuum / Condenser Evacuation Pump Trip With Auto Start Failure Time Position Applicants Actions or Behavior Booth Operator: When directed, EXECUTE Event 5

- Partial Loss of Condenser vacuum @ 5% on 3 minute ramp.

- Condenser Evacuation Pump FW-8B trip.

- Condenser Evacuation Pump FW-8C Auto Start failure.

Indications Available:

CB-10,11/A9 - VACUUM PUMP B STOPPED OR SEAL WATER TEMP HI Emergency Response Facility Computer System (ERFCS) Alarm on Low Condenser Vacuum Condenser Evacuation Pump FW-8B green STOP light lit Lowering Condenser Vacuum on PI-925A/B or P0976A/B Examiner Note: Rate of lowering Condenser vacuum may be modified at your discretion to advance or retard the pace of this and the next event.

+30 sec BOPO RESPOND to Annunciator Response Procedures.

INFORM CRS of lowering Condenser vacuum and Condenser Evacuation BOPO Pump FW-8B trip.

Examiner Note: BOPO may Operate to Mitigate per OPD 4-09 and START FW-8C.

CRS REFER to AOP-26, Turbine Malfunctions,Section I, Loss of Vacuum.

Examiner Note: The following steps are from AOP-26, Turbine Malfunctions,Section I, Loss of Vacuum.

MONITOR Condenser vacuum on ERF Computer System/PI-925A/

BOPO PI-925B/P0976A/P0976B. [Step 4.1]

BOPO ENSURE all Condenser Evacuation Pumps are running. [Step 4.2]

• START FW-8C, Condenser Evacuation Pump.

NRC Simulator Scenario 1 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 5 Page 14 of 29 Event

Description:

Partial Loss of Condenser Vacuum / Condenser Evacuation Pump Trip With Auto Start Failure Time Position Applicants Actions or Behavior CAUTION The Turbine should not be operated with a Generator load of less than 150 MW when vacuum is less than or equal to 23.85" Hg (ERF, P0976A/B) or 6.07" Hg absolute (PI-925A/B) due to possible overheating of final stage blades.

If Condenser vacuum is < 25" Hg or 4.92" Hg Absolute, COMMENCE a plant CRS shutdown to restore vacuum per AOP-05 Emergency Shutdown. [Step 4.3]

Examiner Note: The following steps are from AOP-05, Emergency Shutdown.

NOTE TDB-III-23a and the Power Ascension/Power Reduction Strategy (PAPRs) provide guidance for the shutdown.

CRS CONTACT Reactor Engineer if additional guidance is required. [Step 4.1]

NOTE Operation of more than one Charging Pump will raise the rate of the power reduction.

Examiner Note: Unless directed, boration will occur from the Safety Injection Refueling Water Tank (SIRWT) when in AOP-05 to avoid time constraints.

If borating from SIRWT, COMMENCE boration by performing the following:

CRS

[Step 4.2]

• DETERMINE Charging Pumps CH-1B & CH-1C RUNNING.

ATCO

[Step 4.2.a]

• OPEN LCV-218-3, Charging Pump Suction SIRWT Isolation Valve.

ATCO

[Step 4.2.b]

ATCO

• CLOSE LCV-218-2, VCT Outlet Valve. [Step 4.2.c]

CRS DETERMINE Boration alignment from CVCS NOT required. [Step 4.3]

CRS NOTIFY Energy Marketing of power reduction. [Step 4.4]

NRC Simulator Scenario 1 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 5 Page 15 of 29 Event

Description:

Partial Loss of Condenser Vacuum / Condenser Evacuation Pump Trip With Auto Start Failure Time Position Applicants Actions or Behavior NOTE During the power reduction, maintain TC PER TDB Figure III.1, Tave Program.

MAINTAIN RCS Temperature Control via Turbine Load per HR-12, BOPO Secondary Heat Removal Operation: [Step 4.5]

• MAINTAIN TCOLD 527°F to 547°F AND

• MAINTAIN TCOLD +0°F to -1°F of program.

MAINTAIN Pressurizer Level via Charging and Letdown per IC-11, Inventory ATCO Control: [Step 4.6]

• MAINTAIN Pressurizer Level 45% to 60% AND

• MAINTAIN Pressurizer Level within 4% of program.

When power has been lowered 3% to 5%, PROCEED to Events 6, 7, 8, and 9.

NRC Simulator Scenario 1 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 6, 7, 8, & 9 Page 16 of 29 Event

Description:

Inadvertent Main Turbine Trip / Pressurizer Safety Valve Failure / Loss of Condenser Vacuum /

Pressurizer Pressure Low Signal Actuation Failure / Low Pressure Safety Injection Pumps Start Failure Time Position Applicants Actions or Behavior Booth Operator: When directed, EXECUTE Events 6, 7, 8, and 9.

- Inadvertent Turbine Trip.

- Pressurizer Safety Valve fails 50% open on Reactor Trip.

- Loss of Condenser Vacuum @ 100%.

- Pressurizer Pressure Low Signal (PPLS) Actuation failure.

- Low Pressure Safety Injection Pumps start failure.

Indications Available:

Numerous Reactor Trip and Turbine Trip Alarms.

+10 sec ATCO RECOGNIZE Reactor Trip due to Turbine Trip.

CRS DIRECT performance of EOP-00, Standard Post Trip Actions.

Examiner Note: The following steps are from EOP-00, Standard Post Trip Actions.

ATCO VERIFY Reactivity Control: [Step 1]

• VERIFY ALL of the following: [Step 1.a]

• VERIFY no more than one Regulating or Shutdown CEA NOT inserted.

• VERIFY Reactor Power is LOWERING.

• VERIFY Startup Rate is NEGATIVE.

• MONITOR plant for an uncontrolled RCS Cooldown. [Step 1.b]

CRS DETERMINE Reactivity Control criteria SATISFIED.

BOPO VERIFY Turbine Trip: [Step 2]

• VERIFY HP & LP Stop and Intercept Valves CLOSED.

BOPO ENSURE all Generator Breakers are tripped: [Step 3]

• DETERMINE Generator Output Breaker 3451-4 tripped.

• DETERMINE Generator Output Breaker 3451-5 tripped.

• DETERMINE Generator Field Breaker 41E/G1F tripped.

BOPO VERIFY 4160 V Safeguards Buses 1A3 and 1A4 are ENERGIZED. [Step 4]

NRC Simulator Scenario 1 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 6, 7, 8, & 9 Page 17 of 29 Event

Description:

Inadvertent Main Turbine Trip / Pressurizer Safety Valve Failure / Loss of Condenser Vacuum /

Pressurizer Pressure Low Signal Actuation Failure / Low Pressure Safety Injection Pumps Start Failure Time Position Applicants Actions or Behavior CRS DETERMINE Maintenance of Vital Auxiliaries criteria SATISFIED.

Examiner Note: Diesel Generators only start after Safeguards (PPLS) actuation.

VERIFY both Diesel Generators RUNNING on Safety Injection Actuation BOPO Signal. [Step 5]

VERIFY 4160 V Non-Safeguards Buses 1A1 and 1A2 are ENERGIZED.

BOPO

[Step 6]

BOPO VERIFY 125 VDC Buses 1 and 2 are ENERGIZED. [Step 7]

BOPO VERIFY Instrument Air is AVAILABLE: [Step 8]

• DETERMINE Instrument Air pressure 90 psig.

• DETERMINE Instrument Air Compressor CA-1A RUNNING.

ATCO VERIFY Component Cooling Water System operation NORMAL: [Step 9]

• DETERMINE at least one CCW pump RUNNING. [Step 9.a]

• DETERMINE CCW Pump discharge pressure 60 psig. [Step 9.b]

• DETERMINE HCV-438A/B/C/D, CCW to RCP Coolers OPEN. [Step 9.c]

• DETERMINE at least one Raw Water Pump RUNNING. [Step 9.d]

CRS VERIFY RCS Inventory Control criteria satisfied: [Step 10]

• DETERMINE PZR level NOT between 30% and 70% and NOT ATCO TRENDING to between 45% and 60%.

• DETERMINE RCS subcooling 20°F:

• [CA] RESTORE Inventory Control by manually controlling Charging and Letdown. [Step 10.1.a]

CRS DETERMINE RCS Inventory Control criteria NOT SATISFIED.

CRS VERIFY RCS Pressure Control criteria satisfied: [Step 11]

NRC Simulator Scenario 1 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 6, 7, 8, & 9 Page 18 of 29 Event

Description:

Inadvertent Main Turbine Trip / Pressurizer Safety Valve Failure / Loss of Condenser Vacuum /

Pressurizer Pressure Low Signal Actuation Failure / Low Pressure Safety Injection Pumps Start Failure Time Position Applicants Actions or Behavior

• DETERMINE RCS pressure NOT between 1800 psia and 2300 psia and ATCO NOT trending to between 2050 psia and 2150 psia.

• [CA] DETERMINE RCS pressure < 2300 psia and PORV NOT open.

[Step 11.1]

• [CA] DETERMINE RCS pressure 1350 psia and TRIP one RCP in each loop. [Step 11.2]

• [CA] DETERMINE RCS pressure 1600 psia and ENSURE PPLS actuated. [Step 11.3]

Manually Actuate Pressurizer Pressure Low Signal (PPLS) when RCS Pressure CRITICAL TASK 1600 psia and before 1350 psia to ensure SIAS / VIAS / CIAS Activation.

STATEMENT Pressure at Time of PPLS Trip: ______ psia.

CRITICAL TASK ATCO DETERMINE PPLS relays NOT tripped and manually ACTUATE PPLS.

• [CA] INSERT and TURN keys at 86A/PPLS Test Switch & 86B/PPLS ATCO Test Switch on AI-30A & AI-30B. [Step 11.3.a]

• [CA] DETERMINE PPLS relays 86A/PPLS / 86B/PPLS / 86A1/PPLS /

86B1/PPLS have TRIPPED. [Step 11.3.b]

• [CA] DETERMINE VIAS relays 86A/VIAS / 86B1/VIAS / 86B/VIAS /

86A1/VIAS have TRIPPED. [Step 11.3.c]

• [CA] DETERMINE SIAS relays 86A/SIAS / 86AX/SIAS / 86B1/SIAS /

86B1X/SIAS / 86B/SIAS / 86BX/SIAS / 86A1/SIAS / 86A1X/SIAS have TRIPPED. [Step 11.3.d]

• [CA] DETERMINE CIAS relays 86A/CIAS / 86B1/CIAS / 86B/CIAS /

86A1/CIAS have TRIPPED. [Step 11.3.e]

• [CA] ENSURE HPSI / LPSI / Charging Pumps RUNNING. [Step 11.3.f]

• DETERMINE HPSI Pumps SI-2A & SI-2B or SI-2B & SI-2C RUNNING.

• DETERMINE LPSI Pumps NOT RUNNING and manually ATCO START SI-1A and SI-1B.

• DETERMINE Charging Pumps CH-1A, CH-1B, & CH-1C RUNNING.

• [CA] ENSURE adequate SI flow per IC-13, Safety Injection Flow vs.

Pressurizer Pressure. [Step 11.3.g]

• [CA] DETERMINE Emergency Boration in progress. [Step 11.3.h]

NRC Simulator Scenario 1 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 6, 7, 8, & 9 Page 19 of 29 Event

Description:

Inadvertent Main Turbine Trip / Pressurizer Safety Valve Failure / Loss of Condenser Vacuum /

Pressurizer Pressure Low Signal Actuation Failure / Low Pressure Safety Injection Pumps Start Failure Time Position Applicants Actions or Behavior CRS DETERMINE RCS Pressure Control criteria NOT SATISFIED.

Examiner Note: The following steps are from RC-11, Emergency Boration Verification.

ATCO ENSURE the following valves are CLOSED: [Step 1]

• FCV-269X, Demin Water Makeup Valve

• FCV-269Y, Boric Acid Makeup Valve

• HCV-264, CH-4A Recirc Valve

• HCV-257, CH-4B Recirc Valve ATCO VERIFY all the following valves OPEN: [Step 2]

• HCV-268, Boric Acid Pump Header to Charging Pumps Isolation Valve

• HCV-265, CH-11A Gravity Feed Valve

• HCV-258, CH-11B Gravity Feed Valve ATCO ENSURE all available Boric Acid Pumps RUNNING: [Step 3]

• CH-4A, Boric Acid Pump

• CH-4B, Boric Acid Pump ATCO ENSURE all available Charging Pumps RUNNING: [Step 4]

• CH-1A, Charging Pump

• CH-1B, Charging Pump

• CH-1C, Charging Pump ATCO ENSURE the following valves are CLOSED: [Step 5]

• LCV-218-2, VCT Outlet Valve

• LCV-218-3, Charging Pump Suction SIRWT Isolation Valve Examiner Note: The following steps continue from EOP-00, Standard Post Trip Actions.

CRS VERIFY Core Heat Removal criteria satisfied: [Step 12]

NRC Simulator Scenario 1 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 6, 7, 8, & 9 Page 20 of 29 Event

Description:

Inadvertent Main Turbine Trip / Pressurizer Safety Valve Failure / Loss of Condenser Vacuum /

Pressurizer Pressure Low Signal Actuation Failure / Low Pressure Safety Injection Pumps Start Failure Time Position Applicants Actions or Behavior ATCO

• DETERMINE RCP NPSH requirements met.

• DETERMINE at least one RCP operating.

• DETERMINE Core T 10°F.

Examiner Note: Depending on Crew actions, RCS subcooling will be lost in either EOP-00, SPTAs or EOP-03, LOCA.

Stop All Reactor Coolant Pumps (RCPs) when Subcooling is approaching or is CRITICAL TASK less than 20°F but before 0°F due to Loss of Net Positive Suction Head (NPSH)

STATEMENT per RCP NPSH Curve.

Subcooling at Time of RCP Trip: ______ °F.

CRITICAL TASK ATCO DETERMINE RCP subcooling < 20°F and PERFORM the following:

ATCO

• STOP all RCPs.

• [CA] PLACE TCV-909, Temperature Controller in MANUAL on DCS.

BOPO

[Step 12.2.a]

• [CA] ENSURE TCV-909, Temperature Controller OUTPUT is zero BOPO (0). [Step 12.2.b]

CRS * [CA] VERIFY Natural Circulation in at least one Loop. [Step 12.2.c]

• [CA] DETERMINE Core T 50°F.

• [CA] DETERMINE difference between CETs and RCS THOT is 10°F on ERF "CHR" display.

• [CA] DETERMINE RCS subcooling is 20°F.

• [CA] DETERMINE THOT and TCOLD are stable or lowering.

CRS DETERMINE Core Heat Removal criteria NOT SATISFIED.

NOTE If Instrument Air to valves HCV-1105, HCV-1106, HCV-1107A/B and HCV-1108A/B is not available, throttling of these valves is not possible.

Open or close operation of these valves is possible for a minimum of three cycles.

CRS VERIFY RCS Heat Removal criteria satisfied:

NRC Simulator Scenario 1 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 6, 7, 8, & 9 Page 21 of 29 Event

Description:

Inadvertent Main Turbine Trip / Pressurizer Safety Valve Failure / Loss of Condenser Vacuum /

Pressurizer Pressure Low Signal Actuation Failure / Low Pressure Safety Injection Pumps Start Failure Time Position Applicants Actions or Behavior VERIFY Main Feedwater is restoring SG levels to 35% to 80% NR and 73%

BOPO to 94% WR. [Step 13]

• DETERMINE FCV-1101 & FCV-1102 Feed Regulating Valves CLOSED.

[Step 13.a]

• DETERMINE HCV-1105 & HCV-1106 Feed Regulating Bypass Valves ramped to between 40% & 45% OPEN. [Step 13.b]

BOPO

• PLACE 43/FW switch in OFF. [Step 13.c]

• ENSURE no more than one Main Feedwater Pump RUNNING.

[Step 13.d]

• ENSURE no more than one Condensate Pump RUNNING. [Step 13.e]

• STOP running Heater Drain Pumps FW-5A, FW-5B, and/or FW-5C.

BOPO

[Step 13.f]

• ENSURE both sets of SG Blowdown Isolation Valves CLOSED.

[Step 13.g]

VERIFY Steam Dump and Bypass Valves controlling both of the following:

BOPO

[Step 14]

• VERIFY RCS TCOLD between 525°F and 535°F.

• VERIFY Steam Generator pressure between 850 psia & 925 psia.

• [CA] DETERMINE loss of Condenser vacuum and PLACE HCV-1040, BOPO Atmosphere Dump Valve in service.

• SELECT HCV-1040 on DCS Secondary Screen.

CRS DETERMINE RCS Heat Removal criteria SATISFIED.

CRS VERIFY Containment Isolation criteria satisfied:

ATCO VERIFY Normal Containment conditions exist: [Step 15]

• DETERMINE unexpected rise in Containment Sump level. [Step 15.a]

• DETERMINE Containment Area Radiation Monitors in ALARM.

ATCO

[Step 15.b]

• DETERMINE Containment Ventilation Radiation Monitors RM-050 and ATCO RM-051 in ALARM. [Step 15.c]

• [CA] ENSURE VIAS has ACTUATED and 86A/VIAS, 86A1/VIAS, 86B/VIAS, & 86B1/VIAS relays TRIPPED.

NRC Simulator Scenario 1 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 6, 7, 8, & 9 Page 22 of 29 Event

Description:

Inadvertent Main Turbine Trip / Pressurizer Safety Valve Failure / Loss of Condenser Vacuum /

Pressurizer Pressure Low Signal Actuation Failure / Low Pressure Safety Injection Pumps Start Failure Time Position Applicants Actions or Behavior

• [CA] ENSURE RM-050 & RM-051 Containment Radiation Monitor Sample Pump STOPPED.

• [CA] ENSURE RM-065, Post Accident Control Room Iodine Monitor RUNNING.

• DETERMINE SG Blowdown & Condenser Off Gas Radiation Monitors (RM-054A / RM-054B / RM-057) NOT in alarm. [Step 15.d]

• DETERMINE SG Blowdown & Condenser Off Gas Radiation Monitors (RM-054A / RM-054B / RM-057) NOT trending to alarm. [Step 15.e]

ATCO

• VERIFY Containment conditions: [Step 15.f]

• DETERMINE Containment pressure < 3 psig.

• DETERMINE Containment temperature > 120°F.

CRS DETERMINE Containment Integrity criteria NOT SATISFIED.

CRS DIAGNOSE event in progress: [Step 16]

• DETERMINE Reactivity Control requirements met.

• DETERMINE both DC buses energized.

• DETERMINE at least one Vital 4160 V Bus energized.

• DETERMINE at least one Non-Vital 4160 V Bus energized.

• VERIFY at least one RCP running.

• If not, CONSIDER EOP-02, Loss of Offsite Power/Forced Circulation.

• DETERMINE adequate Feedwater flow to at least one SG.

• VERIFY Pressurizer pressure > 1800 psia with high subcooled margin, normal SG pressure, and no indications of primary to secondary leakage.

• If not, CONSIDER EOP-03, Loss of Coolant Accident.

NOTE Certain events (i.e., LOCA, SGTR, UHE and Loss of All Feedwater) do not require offsite power in order to adequately, mitigate the effects of the accident. For this reason, the LOCA, SGTR, UHE or Loss of All Feedwater procedure may be implemented even if a Loss of Offsite Power has also occurred.

• DETERMINE all Safety Function Acceptance Criteria NOT SATISFIED.

NRC Simulator Scenario 1 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 6, 7, 8, & 9 Page 23 of 29 Event

Description:

Inadvertent Main Turbine Trip / Pressurizer Safety Valve Failure / Loss of Condenser Vacuum /

Pressurizer Pressure Low Signal Actuation Failure / Low Pressure Safety Injection Pumps Start Failure Time Position Applicants Actions or Behavior

• DETERMINE single event in progress and transition to EOP-03, Loss of Coolant in Accident.

Examiner Note: The following steps are from EOP-03, Loss of Coolant Accident.

CRS CONFIRM Standard Post Trip Actions have been performed. [Step 1]

CRS CONFIRM Loss of Coolant Accident Diagnosis: [Step 2]

• VERIFY Safety Function Status Check Acceptance Criteria being satisfied. [Step 2.a]

• VERIFY CIAS is NOT present and SAMPLE both SGs. [Step 2.b]

CRS IMPLEMENT the Emergency Plan. [Step 3]

• Time: __________

NOTE Floating Step BB, Minimizing DC Loads, requires operator action within 15 minutes of loss of either battery charger.

CREW MONITOR the Floating Steps. [Step 4]

DETERMINE RCS pressure 1600 psia and Containment pressure 5 psig CRS and CSAS NOT present. [Step 5]

DETERMINE RCS pressure 1600 psia and VERIFY Engineered ATCO Safeguards Actuation: [Step 6]

• DETERMINE PPLS relays 86A/PPLS / 86B/PPLS / 86A1/PPLS /

86B1/PPLS have TRIPPED. [Step 6.a]

• DETERMINE VIAS relays 86A/VIAS / 86B1/VIAS / 86B/VIAS /

86A1/VIAS relays TRIPPED. [Step 6.b]

• DETERMINE SIAS relays 86A/SIAS / 86AX/SIAS / 86B1/SIAS /

86B1X/SIAS / 86B/SIAS / 86BX/SIAS / 86A1/SIAS / 86A1X/SIAS have TRIPPED. [Step 6.c]

• DETERMINE CIAS relays 86A/CIAS / 86B1/CIAS / 86B/CIAS /

86A1/CIAS relays TRIPPED. [Step 6.d]

NRC Simulator Scenario 1 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 6, 7, 8, & 9 Page 24 of 29 Event

Description:

Inadvertent Main Turbine Trip / Pressurizer Safety Valve Failure / Loss of Condenser Vacuum /

Pressurizer Pressure Low Signal Actuation Failure / Low Pressure Safety Injection Pumps Start Failure Time Position Applicants Actions or Behavior CRS DETERMINE Containment pressure 5 psig. [Step 7]

ATCO DETERMINE SIAS actuated and OPTIMIZE SI flow. [Step 8]

• ENSURE HPSI / LPSI / Charging Pumps RUNNING. [Step 8.a]

• DETERMINE HPSI Pumps SI-2A & SI-2B RUNNING.

• DETERMINE LPSI Pumps SI-1A and SI-1B RUNNING.

• DETERMINE Charging Pumps CH-1A, CH-1B, & CH-1C RUNNING.

• DETERMINE Emergency Boration in progress per RC-11, Emergency Boration Verification. [Step 8.b]

• DETERMINE adequate SI flow per IC-13, Safety Injection Flow vs.

Pressurizer Pressure. [Step 8.c]

CRS VERIFY RCP operating parameters: [Step 9]

ATCO

• ENSURE at least one RCP stopped if TCOLD < 500°F. [Step 9.a]

• ENSURE one RCP stopped in each loop if RCS pressure 1350 psia.

ATCO

[Step 9.b]

• ENSURE all RCPs STOPPED if RCS pressure < NPSH requirements ATCO per PC-12, RCS Pressure-Temperature Limits. [Step 9.c]

CRS RECORD time of SIAS initiation. [Step 10]

• Time: __________

VERIFY normal Component Cooling Water (CCW) and Raw Water (RW)

ATCO System operation: [Step 11]

• ENSURE at least 2 CCW Pumps RUNNING. [Step 11.a]

• VERIFY CCW Pump discharge pressure 60 psig. [Step 11.b]

• ENSURE at least 2 RW Pumps RUNNING. [Step 11.c]

• ENSURE at least 3 CCW Heat Exchangers in service. [Step 11.d]

• ENSURE all RCP Coolers CCW Valves OPEN. [Step 11.e]

NRC Simulator Scenario 1 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 6, 7, 8, & 9 Page 25 of 29 Event

Description:

Inadvertent Main Turbine Trip / Pressurizer Safety Valve Failure / Loss of Condenser Vacuum /

Pressurizer Pressure Low Signal Actuation Failure / Low Pressure Safety Injection Pumps Start Failure Time Position Applicants Actions or Behavior NOTE Do NOT isolate a PORV if the pressurizer is water solid.

ATCO VERIFY PORVs and PZR Code Safety Valves are CLOSED. [Step 12]

• DETERMINE Quench Tank temperature, pressure, and level in ALARM.

[Step 12.a]

• DETERMINE PZR Safety Valve discharge temperature high in ALARM.

[Step 12.b]

ATCO

• NOTIFY CRS that a PZR Safety Valve is OPEN.

• DETERMINE PORV Acoustic Flow Alarms are CLEAR. [Step 12.c]

NOTE Rising Radiation Monitor RM-053 count rate, rising CCW surge tank level or rising CCW surge tank pressure may be indications of a RCS-to-CCW leak.

ATCO DETERMINE RCS to CCW leak is NOT in progress. [Step 13]

CRS DETERMINE LOCA is inside Containment. [Step 14]

ATCO PERFORM the following for a LOCA inside Containment: [Step 15]

• PLACE HC-504A, CNTMT SUMP PUMP WD-3A CONTROL SWITCH, in PULL-TO-LOCK. [Step 15.a]

• PLACE HC-504B, CNTMT SUMP PUMP WD-3B CONTROL SWITCH, in PULL-TO-LOCK. [Step 15.a]

• CLOSE HCV-506A, Containment Sump Isolation Valve. [Step 15.b]

• CLOSE HCV-506B, Containment Sump Isolation Valve. [Step 15.b]

ATCO VERIFY all the following conditions exist: [Step 16]

• DETERMINE all HPSI Pumps are operating.

• DETERMINE SI flowrate is acceptable per IC-13 SI Flow vs. PZR Pressure.

• DETERMINE Representative CET temperature less than superheat.

• DETERMINE Reactor Vessel Level Monitoring System > 43% and NOT lowering.

NRC Simulator Scenario 1 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 6, 7, 8, & 9 Page 26 of 29 Event

Description:

Inadvertent Main Turbine Trip / Pressurizer Safety Valve Failure / Loss of Condenser Vacuum /

Pressurizer Pressure Low Signal Actuation Failure / Low Pressure Safety Injection Pumps Start Failure Time Position Applicants Actions or Behavior ATCO ENSURE SI-2C, HPSI Pump Control Switch in PULL-TO-LOCK.

ATCO DETERMINE NONE of the following conditions exist: [Step 17]

• SI flowrate is less than IC-13 SI Flow vs. PZR Pressure.

• Representative CET temperature greater than superheat.

• Reactor Vessel Level Monitoring System < 43% and lowering.

CRS DETERMINE RCS leak is NOT isolated. [Step 18]

DETERMINE Steam Generator Isolation Signal (SGIS) NOT actuated.

BOPO

[Step 19]

DETERMINE SG levels between 35% and 85% NR using Main Feedwater.

BOPO

[Step 20]

• MAINTAIN Feedwater flow per HR-15, Main Feed Pump Operation.

[Step 20.a]

• CONTROL Feedwater flow per HR-11, Manual Feet Control (DCS).

[Step 20.b]

CAUTION Failure to place the Containment Spray Pumps to Pull to Lock may allow actuation of Spray into Containment. This can lead to Containment Sump Blockage.

ATCO SECURE all Containment Spray flow: [Step 21]

• PLACE SI-3A, Containment Spray Pump in PULL-TO-LOCK.

• PLACE SI-3B, Containment Spray Pump in PULL-TO-LOCK.

• PLACE SI-3C, Containment Spray Pump in PULL-TO-LOCK.

CAUTION

1) When TC is 178°F or greater, the maximum RCS cooldown rate is 100°F/hr. When TC is less than 178°F, the maximum RCS cooldown rate is 50°F/hr.

2) No more than three RCPs shall be in operation when RCS temperature is less than 500°F.

COMMENCE a Steam Generator cooldown per HR-12, Secondary Heat CRS Removal Operation. [Step 22]

NRC Simulator Scenario 1 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 6, 7, 8, & 9 Page 27 of 29 Event

Description:

Inadvertent Main Turbine Trip / Pressurizer Safety Valve Failure / Loss of Condenser Vacuum /

Pressurizer Pressure Low Signal Actuation Failure / Low Pressure Safety Injection Pumps Start Failure Time Position Applicants Actions or Behavior

• Time: __________

MAINTAIN RCS pressure per PC-12, Pressure-Temperature Limits.

CRS

[Step 23]

• CONTROL RCS heat removal per HR-12, Secondary Heat Removal BOPO Operation. [Step 23.a]

ATCO

• CONTROL RCS pressure per PC-11, Pressure Control. [Step 23.b]

• If HPSI Stop and Throttle criteria are met, CONTROL Charging, ATCO Letdown, and HPSI flow per IC-11, Inventory Control. [Step 23.c]

NOTE Voiding of the RCS is indicated by the inability to depressurize to SDC entry pressure.

Attachment IC-14, RCS Void Elimination, provides guidance to correct this condition.

COMMENCE depressurizing RCS to 300 psia using any of the following CRS per PC-11, Pressure Control: [Step 24]

• CONTROL Pressurizer Spray flow.

• CONTROL Charging and Letdown flow.

• THROTTLE HPSI Pumps.

• Time: __________

Commence a Cooldown and Depressurization of the Reactor Coolant System before Reactor Vessel Level Monitoring System (RVLMS) is less than 83%,

CRITICAL TASK indicating a bubble has formed in the head, to Reestablish RCS Inventory STATEMENT Control while maintaining RCS Heat Removal.

RVLMS at start of Cooldown: ______

CRITICAL IMPLEMENT HR-12, Secondary Heat Removal Operation, to lower RCS TASK BOPO temperature.

Examiner Note: The following steps are from HR-12, Secondary Heat Removal Operation.

BOPO ENSURE Turbine Control is in MANUAL. [Step 1]

BOPO * [CA] DETERMINE Turbine NOT online and GO TO Step 4.

NRC Simulator Scenario 1 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 6, 7, 8, & 9 Page 28 of 29 Event

Description:

Inadvertent Main Turbine Trip / Pressurizer Safety Valve Failure / Loss of Condenser Vacuum /

Pressurizer Pressure Low Signal Actuation Failure / Low Pressure Safety Injection Pumps Start Failure Time Position Applicants Actions or Behavior NOTES

1. In MANUAL, single arrows are 1% / double arrows are 5% change in OUTPUT value.

2. While Steam Dump and Bypass Control is in Temperature/Pressure Mode, the controllers PC0910 and TC0909_PI will alternate between controls, depending on the higher output signal. A red square outlining the controlling function signify which parameter is in control.

3. Steps 4 through 11 may be performed as needed, and in any order.

CAUTIONS When TC is 178°F or greater, the maximum RCS cooldown rate is 100°F/hr.

When TC is less than 178°F, the maximum RCS cooldown rate is 50°F/hr.

If Steam Dump and Bypass (SD&B) is available, CONTROL RCS BOPO temperature with a single SD&B Valve. [Step 4]

• [CA] DETERMINE Steam Dump and Bypass is NOT available and GO BOPO TO Step 9.

Examiner Note: HCV-1040, Atmospheric Dump Valve, may already be in service following the Loss of Condenser Vacuum that occurred on Reactor Trip.

BOPO If HCV-1040, is available, CONTROL RCS temperature as follows: [Step 9]

• DEPRESS the valve toggle to SELECT HCV-1040. [Step 9.a]

• PUSH UP and DOWN arrows as required to ADJUST HCV-1040 output as needed. [Step 9.b]

CRITICAL TASK ATCO IMPLEMENT PC-11, Pressure Control, to lower RCS pressure.

Examiner Note: The following steps are from PC-11, Pressure Control. PC-12, RCS Pressure-Temperature Limits (graph), is maintained on a Control Room hardcopy.

CAUTION A charging header flow path must be maintained at all times.

MAINTAIN RCS pressure per the PC-12, RCS Pressure-Temperature Limits ATCO graph: [Step 1]

• DETERMINE Steps 1.a through 1.d N/A. [Step 1.e]

NRC Simulator Scenario 1 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 6, 7, 8, & 9 Page 29 of 29 Event

Description:

Inadvertent Main Turbine Trip / Pressurizer Safety Valve Failure / Loss of Condenser Vacuum /

Pressurizer Pressure Low Signal Actuation Failure / Low Pressure Safety Injection Pumps Start Failure Time Position Applicants Actions or Behavior

• CONTROL Auxiliary Spray flow as necessary by operating the following:

[Step 1.e]

• HCV-240, PZR Auxiliary Spray Isolation Valve

• HCV-249, PZR Auxiliary Spray Isolation Valve

• HCV-238, Loop 1 Charging Isolation Valve

• HCV-239, Loop 2 Charging Isolation Valve

• If HPSI Stop and Throttle criteria is met, CONTROL Pressurizer level using Charging, Letdown, and/or HPSI flow per IC-11, Inventory Control.

[Step 1.f]

• CONTROL RCS heat removal per HR-12, Secondary Heat Removal Operation. [Step 1.g]

When RCS Cooldown and Depressurization is in progress, TERMINATE the scenario.

NRC Simulator Scenario 1 Outline Rev. Final

Appendix D Scenario Outline Form ES-D-1 Facility: Fort Calhoun Station Scenario No.: 3 Op Test No.: Dec 2015 NRC Examiners: Operators:

Initial Conditions: 100% power MOL - RCS Boron is 482 ppm (by sample).

Turnover: Maintain steady-state power conditions. Rotate Heater Drain Pumps FW-5B and FW-5C per OI-VD-1, Feedwater Heater Vents and Drains Normal Operation. Charging Pump CH-1C out of service for packing repair.

Critical Tasks:

• Stop One Reactor Coolant Pump in Each Loop when Reactor Coolant System Pressure is

< 1350 psia, Prior to losing Reactor Coolant Pump Net Positive Suction Head. (Event 7)

• Reduce and Maintain RCS THOT 510°F to Maintain SG Pressure Below Safety Valve Setpoint of 1000 psia Prior to Isolating SG RC-2B. (Event 7)

• Isolate the Affected Steam Generator with a Tube Rupture to Minimize Spread of Contamination Prior to Exiting EOP-04, Steam Generator Tube Rupture. (Event 7)

Event No. Malf. No. Event Type* Event Description 1 N (BOPO) Rotate Heater Drain Pumps per OI-VD-1, Feedwater Heater Vents

+10 min and Drains Normal Operation, Attachment 2.

2 I (ATCO, CRS) Pressurizer Level Channel Transmitter LT-101X Fails Low.

+20 min Transfer Pressurizer Level Control to LT-101Y.

3 I (BOPO, CRS) Steam Generator RC-2A Steam Flow Transmitter FT-907 Fails

+30 min High. Bypass Affected Transmitter.

4 C (ATCO, CRS) Charging Pump CH-1A Trip.

+40 min TS (CRS) Restore Letdown and Charging Flow.

5 C (ATCO,BOPO, Steam Generator RC-2B Tube Leak Greater Than 150 GPD.

+50 min CRS) TS (CRS) Isolate Blowdown Flow.

6 R (ATCO) Commence Plant Shutdown per AOP-05, Emergency Shutdown.

+60 min N (BOPO, CRS) 7 M (ATCO, BOPO, Steam Generator RC-2B Tube Rupture at 500 GPM on 10 Minute

+70 min CRS) Ramp Upon 3% to 5% Load Reduction.

8 I (BOPO) Diesel Generator DG-01 Start Failure on SIAS.

+70 min Manual Start Required.

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Technical Specifications Actual Target Quantitative Attributes 1 Malfunctions after EOP entry (1-2) 4 Abnormal events (2-4) 1 Major transients (1-2) 1 EOPs entered/requiring substantive actions (1-2) 0 EOP contingencies requiring substantive actions (0-2) 3 Critical tasks (2-3)

NRC Simulator Scenario 3 Outline Rev. Final

Scenario Event Description NRC Scenario 3 SCENARIO

SUMMARY

NRC 3 The crew will assume the shift at 100% power per OP-4, Load Change and Normal Power Operation.

The scheduled activity is to rotate Heater Drain Pumps by starting FW-5C and securing FW-5B per OI-VD-1, Feedwater Heater Vents and Drains Normal Operation, Attachment 2, Rotating Operating Heater Drain Pumps.

The next event is a low failure of Pressurizer Level Control Channel, LT-101X. Operator actions are per ARP-CB-1/2/3/A4, Window C PRESSURIZER LEVEL LO-LO CHANNEL X. The crew will transfer to the standby channel LT-101Y and restore Letdown per OI-RC-8, Reactor Coolant System Level Control Normal Operation, Attachment 8, Transferring Pressurizer Level Control Channel in CASCADE and Attachment 4, Transferring Letdown Controller from AUTOMATIC to MANUAL.

When plant conditions are stable, a high failure of Steam Generator RC-2A Steam Flow Transmitter FT-907 will occur. Initial operator actions are per ARP-DCS-FW, Feedwater DCS Annunciator Response Procedure and include verifying Feedwater Control is in Single Element Control, bypassing the failed input, and determining 3 Element Control is restored.

The next event is a trip of the running Charging Pump. Operator actions are per ARP-CB-1/2/3/A2, Window A-6L - CHARGING FLOW LO and include isolating of Letdown and verifying no system leaks exist. Charging Pump CH-1B is placed in service per OI-CH-1, Chemical and Volume Control System Normal Operation, Attachment 1, Startup of Charging and Letdown. The SRO will refer to Technical Specification LCO 2.2.4 - Charging Pumps - Operating.

When Charging flow is restored, a Steam Generator Tube Leak of greater than 150 gallons per day will occur on Steam Generator RC-2B. The crew will enter AOP-22, Reactor Coolant Leak, and implement Attachment B, Primary to Secondary Leak Rate Actions. RM-064, Main Steam Line Radiation Monitor, is placed in service to assist in determining leak size and location. Various Secondary Side valves are closed to minimize system contamination and HR-21, Blowdown Operation is performed to isolate blowdown flow from SG RC-2B. The SRO will refer to Technical Specification LCO 2.1.4 - Reactor Coolant System Leakage Limits.

Once blowdown is isolated, entry into AOP-05, Emergency Shutdown, is performed to bring the plant into MODE 4. When power has been reduced 3% to 5%, a Steam Generator Tube Rupture of 500 gpm will commence on a 10 minute ramp.

The crew enters EOP-00, Standard Post Trip Actions, and then transitions to EOP-04, Steam Generator Tube Rupture. Diesel Generator DG-01 fails to start upon SIAS and must be manually started. While in EOP-04, the Reactor Coolant System is cooled per HR-12, Secondary Heat Removal Operation, and the RCS is depressurized to less than 1000 psia per PC-11, Pressure Control, to allow isolating the affected Steam Generator. When SG RC-2B is isolated, the scenario is terminated.

Risk Significance:

• Failure of risk important system prior to trip: Loss of Charging Pump Steam Generator Tube Leak

• Risk significant core damage sequence: Steam Generator Tube Rupture

• Risk significant operator actions: Stop RCPs Upon Loss of Subcooling Isolate Affected Steam Generator Cooldown and Depressurize RCS NRC Simulator Scenario 3 Outline Rev. Final

Scenario Event Description NRC Scenario 3 BOOTH OPERATOR INSTRUCTIONS for SIMULATOR SETUP RESET to IC-103 and LOAD & EXECUTE NRC 3.sce for NRC Scenario 3.

Preset Item - CH-1C Removed from Service Type Item Value Condition Malfunction BUS_1B3B_4B_5_BKR_Trip True Scenario Event: CH-1C OOS Preset Item - Event 9 - Diesel Generator #1 Auto Start Failure Type Item Value Condition Expert H_PD1_033_3 Reset Scenario Event: DG-1 H_PD1_031_3 Reset Auto Start Failure Event 2 - Pressurizer Level Transmitter LT-101X Fails Low Type Item Value Condition Transmitter RCS_LT101X 0, ramp = 5 seconds When directed by examiner, trigger/activate this event.

Scenario Event: Pzr Level LT-101X Fail Low Event 3 - Steam Generator Flow Transmitter LT-907 Fails High Type Item Value Condition Transmitter FT-907 4000000, ramp = 5 sec When directed by examiner, trigger/activate this event.

FT-907 DCS Fail High Scenario Event: SG Flow FT-907-1 DCS Fail High FT-907 Fail High Event 4 - Charging Pump CH-1A trips Type Item Value Condition Malfunction BUS_1B3A_4_BKR_TRIP True When directed by examiner, trigger/activate this event.

Scenario Event: CH-1A Trip Event 5 - Primary-to-Secondary SG Tube Leak Develops in Steam Generator RC-2B Type Item Value Condition Malfunction RCS04B 0.001 When directed by examiner, trigger/activate this event.

Scenario Event: RC-2B S/G Tube Leak Event 7 - Steam Generator Tube Leak in RC-2B Grows to Tube Rupture Type Item Value Condition Malfunction RCS04B 1.4, ramp = 600 sec When directed by examiner, trigger/activate this event.

Scenario Event: RC-2B S/G Tube Rupture NRC Simulator Scenario 3 Outline Rev. Final

Scenario Event Description NRC Scenario 3 Booth Operator: INITIALIZE to IC-103 and LOAD NRC 3.sce.

ENSURE all Simulator Annunciator Alarms are ACTIVE.

ENSURE all Control Board Tags are removed.

ENSURE Charging Pump CH-1A in service.

ENSURE Charging Pump CH-1C OOS for emulsified oil replacement with Information Tag attached.

ENSURE Channel X Pressurizer Pressure and Level selected.

ENSURE Reactivity Briefing Sheet printout provided with Turnover.

ENSURE Middle-of-Life Thumb Rule Sheet provided with Turnover.

ENSURE ERF Computer System Display set to FWD for BOPO.

ENSURE procedures in progress provided to crew in Briefing Room:

- COPY of OI-VD-1, Feedwater Heater Vents and Drains Normal Operation, Attachment 2, Rotating Operating Heater Drains Pumps, INITIALED through Prerequisites and Procedure Step 2.

Control Room Annunciators in Alarm:

0B AI-30-ESF - CHARGING PUMP CH-1C OFF NORMAL Procedure List Event 1: OP-4, Load Change and Normal Power Operation.

Event 1: OI-VD-1, Feedwater Heater Vents and Drains Normal Operation Event 2: ARP-CB-1/2/3/A4, Window C-8, PRESSURIZER LEVEL LO-LO CHANNEL X Event 3: ARP-DCS-FW, Feedwater DCS Annunciator Response Procedure Event 4: ARP-CB-1/2/3/A2, Window A-6L, CHARGING FLOW LO Event 4: OI-CH-1, Chemical and Volume Control System Normal Operation, Attachment 1, Startup of Charging and Letdown Event 5: AOP-22, Reactor Coolant Leak Event 5: HR-21, Blowdown Operation Event 6: AOP-05, Emergency Shutdown Event 7: EOP-00, Standard Post Trip Actions Event 7: EOP-04, Steam Generator Tube Rupture Event 8: HR-12, Secondary Heat Removal Operation Event 8: PC-11, Pressure Control NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 1 Page 5 of 36 Event

Description:

Rotate Heater Drain Pumps Time Position Applicants Actions or Behavior Booth Operator: When directed, RESPOND to requests from Control Room.

Indications Available:

NONE Examiner Note: The following steps are from OI-VD-1, Feedwater Heater Vents and Drains Normal Operation, Attachment 2.

BOPO PERFORM the following at CB-10, 11: [Step 3]

• PLACE 43/FW Switch in OFF. [Step 3.a]

• VERIFY Annunciator CB-10,11/A10, Window B-6L - 43/FW TRANSFER SWITCH OFF AUTO in ALARM. [Step 3.b]

Examiner Note: XC105 is the Computer (DCS) generated value for Secondary Calorimetric.

CRS DECLARE XC105 invalid. [Step 4]

Make plant announcement, then:

BOPO PLACE FW-5C, Heater Drain Pump control switch to AFTER-START at CB-10, 11. [Step 5]

VERIFY FW-5C, Heater Drain Pump ammeter returns to less than 80 amps BOPO in less than 15 seconds and STABILIZES at ~ 66 amps. [Step 6]

Booth Operator: If contacted, REPORT FCV-1216C is closed.

VERIFY FCV-1216C, Heater Drain Pump FW-5C Recirculation Control Valve BOPO CLOSES. [Step 7]

PLACE FW-5B, Heater Drain Pump control switch to AFTER-STOP at BOPO CB-10, 11. [Step 8]

NOTE Verification of Cooling Water Flow to the Seal cooler will be used to ensure Stuffing Box pressure is < 250 psig when Pressure Gauge PI-1192A, B, or C is out of service.

NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 1 Page 6 of 36 Event

Description:

Rotate Heater Drain Pumps Time Position Applicants Actions or Behavior Booth Operator: If contacted, REPORT FW-5C discharge and stuffing box pressures normal.

BOPO MONITOR the following parameters on Heater Drain Pump FW-5C: [Step 9]

• Motor amperage at ~66 amps.

• PI-1269C, Pump Discharge pressure at ~160 psig on ERF Computer.

• Heater Drain Tank level ~54% on CB-10, 11.

• Bearing temperatures on ERF Display FWD normal.

• PI-1192C, Stuffing Box pressure < 250 psig read locally.

Booth Operator: If contacted, REPORT FW-5B is not rotating in reverse.

CONTACT Auxiliary Operator to VERIFY FW-5B, Heater Drain Pump NOT BOPO ROTATING in reverse direction. [Step 10]

BOPO PERFORM the following at CB-10, 11: [Step 11]

• PLACE 43/FW Switch in AUTO. [Step 11.a]

• VERIFY Annunciator CB-10, 11/A10, Window B-6L - 43/FW TRANSFER SWITCH OFF AUTO is CLEAR. [Step 11.b]

Booth Operator: If contacted, REPORT Shift Technical Advisor will restore GARDEL.

CONTACT Shift Technical Advisor to RESTORE GARDEL data feed per CRS OI-ERFCS-2. [Step 12]

When 12 minute validity period has passed and parameters are steady-state, STA DECLARE XC105 valid and ENTER in Control Room Log. [Step 13]

When restoration of XC105 is discussed, PROCEED to Event 2.

NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 2 Page 7 of 36 Event

Description:

Pressurizer Level Channel Transmitter Failure Time Position Applicants Actions or Behavior Booth Operator: When directed, EXECUTE Event 2.

- Pressurizer Level Channel Transmitter LT-101X fails low.

Indications Available:

CB-1,2,3/A4 - PRESSURIZER LEVEL LO-LO CHANNEL X CB-1,2,3/A4 - PRESSURIZER LEVEL HI-LO CHANNEL X Charging Pump CH-1B starts Letdown flow to minimum (~26 gpm)

+30 sec ATCO RESPOND to Annunciator Response Procedures.

ATCO INFORM CRS of Pressurizer Level Channel LT-101X failure.

Examiner Note: ATCO may Operate to Mitigate per OPD 4-09 and TRANSFER to Channel Y.

REFER to ARP-CB-1,2,3/A4, Window C PRESSURIZER LEVEL LO-LO CRS CHANNEL X.

Examiner Note: During this event, Pressurizer pressure may decrease to less than 2075 psia.

If this occurs, the crew should address Technical Specification LCO 2.10.4.5 for Pressurizer low pressure.

Examiner Note: The following steps are from ARP-CB-1,2,3/A4, Window C-8, PRESSURIZER LEVEL LO-LO CHANNEL X.

ATCO VERIFY Pressurizer Level on LR-101X/LR-101Y. [Step 1]

• If Pressurizer level is NOT low, PERFORM the following: [Step 1.1]

• PLACE HC-101 to Channel Y per OI-RC-8. [Step 1.1.1]

• If desired, PLACE HIC-101-1/101-2, Letdown Throttle Valves Controller to MANUAL per OI-RC-8. [Step 1.1.2]

• PLACE HC-101-1, Pzr Heater Cutout Channel Select Switch, to Channel Y. [Step 1.1.3]

Examiner Note: The following steps are from OI-RC-8, Reactor Coolant System Level Control Normal Operation, Attachment 8, Transferring Pressurizer Level Control Channel (X to Y or Y to X) in CASCADE.

NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 2 Page 8 of 36 Event

Description:

Pressurizer Level Channel Transmitter Failure Time Position Applicants Actions or Behavior ATCO ENSURE both Level Controllers are in (C) CASCADE: [Step 1]

• LC-101X-1, Pressurizer Level Controller

• LC-101Y-1, Pressurizer Level Controller If desired, PLACE Letdown Controller HIC-101-1/101-2, Letdown Throttle ATCO Valves Controller in MANUAL per Attachment 4. [Step 2]

Examiner Note: The following steps are from OI-RC-8, Attachment 4, Transferring Letdown Controller from AUTOMATIC to MANUAL.

ENSURE Letdown Controller HIC-101-1/101-2, Letdown Throttle Valves ATCO Controller in AUTO. [Step 1]

ATCO PLACE HIC-101-1/101-2 Manual/Auto Transfer Switch to BAL. [Step 2]

ADJUST Manual Control Knob on HIC-101-1/101-2 until TOP SCALE ATCO indicates 50% (zero deviation; red pointer aligned with the red dot). [Step 3]

ATCO PLACE HIC-101-1/101-2 Manual/Auto Transfer Switch to MAN. [Step 4]

If necessary, MAKE adjustments to HIC-101-1/101-2 Manual Control Knob to ATCO MAINTAIN desired Pressurizer Level. [Step 5]

Examiner Note: The following steps continue from OI-RC-8, Attachment 8.

CAUTION Transfer from the Selected Controller to the Non-Selected Controller should not be performed until both controller outputs are approximately equal.

VERIFY Controller LR-101Y has INDICATED Pressurizer Level and ATCO PROGRAMMED Pressurizer Level Setpoint MATCHED prior to transfer.

[Step 3]

PLACE HC-101, Pressurizer Level Channel Selector Switch, to Channel Y.

ATCO

[Step 4]

NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 2 Page 9 of 36 Event

Description:

Pressurizer Level Channel Transmitter Failure Time Position Applicants Actions or Behavior ENSURE Controller LC-101Y-1 is controlling INDICATED Pressurizer Level ATCO at PROGRAMMED Setpoint. [Step 5]

PUSH LC-101-1 & LC-101-2, Charging Pump Bistable Reset buttons on ATCO Reactor Regulating System Panel AI-4B and VERIFY all bistables are RESET. [Step 6]

If required, PLACE Letdown Controller HIC-101-1/101-2, Letdown Throttle ATCO Valves Controller, in AUTO per Attachment 3. [Step 7]

Examiner Note: The following steps are from OI-RC-8, Attachment 3, Transferring Letdown Controller from MANUAL to AUTOMATIC.

ENSURE Letdown Controller HIC-101-1/101-2, Letdown Throttle Valves ATCO Controller is in (M) MANUAL. [Step 1]

Manually ADJUST HIC-101-1/101-2, Letdown Throttle Valves Controller and ATCO PIC-210, Letdown Press Controller until following parameters are met:

[Step 2]

• Indicated Pressurizer Level matches the Programmed Pressurizer Level Setpoint on LR-101X or LR-101Y, Pressurizer Level Recorder.

• PIC-210 is maintaining 200 psi to 400 psi.

ADJUST bias knob on HIC-101-1/101-2 until the top scale indicates 50%

ATCO (zero deviation; red pointer aligned with the red dot). [Step 3]

PLACE HIC-101-1/101-2 Manual/Auto Transfer Switch to BAL, then to ATCO AUTO. [Step 4]

If necessary, ADJUST the bias knob of HIC-101-1/101-2 to ENSURE ATCO Indicated Pressurizer Level is maintained at Programmed Pressurizer Level setpoint. [Step 5]

NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 2 Page 10 of 36 Event

Description:

Pressurizer Level Channel Transmitter Failure Time Position Applicants Actions or Behavior Examiner Note: The following steps continue from ARP-CB-1,2,3/A4, Window C-8.

ATCO VERIFY RCS Pressure on PR-103X/PR-103Y > 1600 psia. [Step 2]

ATCO ENSURE all Pressurizer Heaters DEENERGIZED. [Step 3]

DETERMINE RCS Cold Leg temperatures on A-D/TI-112C and A-D/TI-122C ATCO are NOT lowering. [Step 4]

• CHECK VCT level on LI-219, for indication of lowering level. [Step 4.1]

• DETERMINE VCT level is NOT lowering. [Step 4.2]

ATCO VERIFY the following CVCS parameters: [Step 5]

• ENSURE Letdown at minimum flow of 26 gpm on FIC-212. [Step 5.1]

• ENSURE Charging Pumps CH-1A & CH-1B are RUNNING. [Step 5.2]

ATCO NOTIFY Work Week Manager of Pressurizer level instrument failure. [Step 6]

When Pressurizer level is in AUTO, PROCEED to Event 3.

NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 3 Page 11 of 36 Event

Description:

Steam Generator Steam Flow Transmitter Failure Time Position Applicants Actions or Behavior Booth Operator: When directed, EXECUTE Event 3.

- Steam Generator RC-2A Steam Flow Transmitter FT-907 fails high.

Indications Available:

Feedwater Digital Control System Alarm

+30 sec BOPO RESPOND to Annunciator Response Procedures.

INFORM CRS Steam Generator RC-2A Steam Flow Transmitter FT-907 BOPO failed high.

CRS DIRECT actions of ARP-DCS-FW, FT-907.

Examiner Note: The following steps are from ARP-DCS-FW, Feedwater Digital Control System.

BOPO PERFORM the following for Steam Flow Instrument FT-907 failure: [Step 1]

• VERIFY that FORCED TO 1 ELEM and 1 ELEM AUTO is displayed on Feedwater Regulating System display for RC-2A PT-907. [Step 1.1]

• TOUCH display with the BAD process. [Step 1.2]

• DETERMINE BAD input NOT automatically bypassed. [Step 1.3]

• TOUCH Bypass on verification faceplate to BYPASS BAD input.

[Step 1.3.1]

• VERIFY point displays GOOD status. [Step 1.3.2]

• ENSURE control SHIFT to 3 ELEMENT AUTO. [Step 1.3.3]

CRS DETERMINE Steam Generator level instruments NOT affected. [Step 2]

CRS DETERMINE BAD input bypassed MANUALLY. [Step 3]

BOPO MONITOR Steam Generator levels. [Step 4]

CRS VERIFY XC-105, Secondary Calorimetric, is valid. [Step 5]

CRS DETERMINE LT-903 or LT-906 NOT cause of alarm. [Step 6]

NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 3 Page 12 of 36 Event

Description:

Steam Generator Steam Flow Transmitter Failure Time Position Applicants Actions or Behavior BOPO NOTIFY Work Week Manager of FT-907 malfunction. [Step 7]

When Steam Generator levels are normal, PROCEED to Event 4.

NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 4 Page 13 of 36 Event

Description:

Charging Pump Trip Time Position Applicants Actions or Behavior Booth Operator: When directed, EXECUTE Event 4.

- Charging Pump CH-1A trip.

Indications Available:

CB-1,2,3/A2 - CHARGING PUMPS TRIP CB-1,2,3/A2 - CHARGING FLOW LO

+30 sec ATCO RESPOND to Annunciator Response Procedures.

Examiner Note: ATCO may Operate to Mitigate per OPD 4-09 and START CH-1B to avoid losing Letdown flow. Charging Pump CH-1B does not AUTO START until a level deviation exists. If Letdown is lost, steps to restore are included at the end of the event ATCO INFORM CRS of Charging Pump CH-1A trip.

CRS REFER to ARP-CB-1,2,3/A2, Window A-6L - CHARGING FLOW LO.

Examiner Note: The following steps are from ARP-CB-1,2,3/A2, Window A-6L - CHARGING FLOW LO.

ATCO OBSERVE Charging Header flow LOW. [Step 1]

If Charging flow is lost, CLOSE TCV-202 and HCV-204 to ISOLATE ATCO Letdown. [Step 2]

• DETERMINE TCV-202, Letdown to Regenerative Heat Exchanger Isolation Valve AUTO CLOSED or manually CLOSE.

• Manually CLOSE HCV-204, Reactor Coolant to Letdown Heat Exchanger Isolation Valve.

NOTE Based on plant conditions, XC-105 and GARDEL may be invalid.

Booth Operator: When contacted about the status of CH-1A, REPORT a breaker overcurrent trip. Investigation of CH-1A: The pump looks normal locally. If Maintenance or Work Week Manager is contacted, estimated time to restore CH-1C is 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 4 Page 14 of 36 Event

Description:

Charging Pump Trip Time Position Applicants Actions or Behavior If required, ROTATE Charging Pumps per OI-CH-1, CVCS Normal ATCO Operation, Attachment 1, Startup of Charging and Letdown. [Step 5]

EVALUATE Technical Specification LCO 2.2, Chemical and Volume Control CRS System

• LCO 2.2.4 - Charging Pumps - Operating

• CONDITION LCO 2.2.4.(1) - Two Charging Pumps inoperable

• ACTION LCO 2.2.4.(1) - RESTORE to at least two OPERABLE Charging Pumps within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

Examiner Note: The following steps are from OI-CH-1, CVCS Normal Operation, Attachment 1, Startup of Charging and Letdown.

ATCO START CH-1B-1, Packing Cooling Pump. [Step 1]

ATCO DETERMINE boron equalization not required. [Step 2]

ENSURE LCV-218-2, Volume Control Tank Outlet Valve, OPEN and in AUTO.

ATCO

[Step 3]

CAUTION HCV-247, Charg to RC Loop 1A Isolation Valve must remain open to provide an alternate makeup path for charging and ensure CH-202, Ltdn to Regen Ht Exch Isolation Valve will be able to relieve thermal expansion in the Regenerative Heat Exchanger ATCO ENSURE HCV-247, Charging to RC Loop 1A Isolation Valve, OPEN. [Step 4]

ENSURE one of the following combinations of Charging Isolation Valves ATCO OPEN: [Step 5]

• Charging to RC Loop 1A Isolation Valves

• HCV-247, Charging to RC Loop 1A Isolation Valve

• HCV-238, Charging to RC Loop 1A Isolation Valve

• Charging to RC Loop 2A Isolation Valves

• HCV-248, Charging to RC Loop 2A Isolation Valve NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 4 Page 15 of 36 Event

Description:

Charging Pump Trip Time Position Applicants Actions or Behavior

• HCV-239, Charging to RC Loop 2A Isolation Valve ATCO DETERMINE bypass of Ion Exchangers not required. [Step 6]

ATCO PLACE HC-101-3, Limiter Bypass Switch, in BYPASS. [Step 7]

PLACE HIC-101-1/101-2, Letdown Throttle Valve Controller, in MANUAL.

ATCO

[Step 8]

ATCO CLOSE LCV-101-1, Letdown Heat Exchanger Throttle Valve. [Step 9]

DETERMINE HC-101-2, Letdown Heat Exchanger Valves Selector Switch ATCO positioned as required. [Step 10]

ATCO PLACE PIC-210, Letdown Pressure Controller, in MANUAL. [Step 11]

ATCO THROTTLE PIC-210 to approximately 10% open. [Step 12]

PLACE Charging Pumps Mode Select that to CH-1C - CH-1A position.

ATCO

[Step 13]

CAUTION Pressurizer Level deviation will start standby charging pumps not in PULLOUT.

ATCO START CH-1B, Charging Pump. [Step 14]

ATCO OPEN HCV-204, RC to Letdown Heat Exchanger Isolation Valve. [Step 15]

OPEN TCV-202, Letdown to Regen Heat Exchanger Isolation Valve.

ATCO

[Step 16]

Using HIC-101-1/101-2 INITIATE Letdown flow while adjusting PCV-210 to ATCO maintain Letdown pressure approximately 300 psig. [Step 17]

NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 4 Page 16 of 36 Event

Description:

Charging Pump Trip Time Position Applicants Actions or Behavior BALANCE Charging and Letdown flows to maintain Pressurizer level.

ATCO

[Step 18]

ATCO PLACE HC-101-3 in NORMAL. [Step 19]

ATCO PERFORM the following to place PIC-210 in AUTO: [Step 20]

• PLACE HIC-101-1/101-2 Manual/Auto Transfer Switch to BAL, then to AUTO. [Step 20.a]

• PLACE PIC-210 in AUTO. [Step 20.b]

When Charging and Letdown flows are restored, PROCEED to Event 5.

NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 5 Page 17 of 36 Event

Description:

Steam Generator Tube Leak Time Position Applicants Actions or Behavior Booth Operator: When directed, EXECUTE Event 5.

- Steam Generator RC-2B Tube Leak greater than 150 gpd.

Indications Available:

RM-057, Condenser Off Gas Radiation Monitor in alarm and trending up RM-054B, Steam Generator RC-2B Blowdown Radiation Monitor in alarm and trending up

+30 sec ATCO RESPOND to Radiation Monitor Alarms.

ATCO INFORM CRS of indications of the tube leak on Steam Generator RC-2B.

CRS REFER to AOP-22, Reactor Coolant Leak.

Examiner Note: The following steps are from AOP-22, Reactor Coolant Leak,Section I, Leak Rate Determination and Leak Isolation.

CRS DETERMINE Shutdown Cooling is NOT in operation. [Step 4.1]

Booth Operator: When contacted as Shift Chemist, WAIT 2 minutes and REPORT Steam Generator RC-2B has increased activity and RC-2A has normal activity.

DETERMINE CIAS is NOT present and DIRECT Shift Chemist to PERFORM CRS the following: [Step 4.2]

• PERFORM rapid activity analysis of both Steam Generators. [Step 4.2.a]

• SAMPLE both SGs per CH-SMP-SE-0015, Steam Generator Sampling -

Room 60. [Step 4.2.b]

CRS IMPLEMENT the Emergency Plan. [Step 4.3]

CREW MONITOR the Floating Steps. [Step 4.4]

ATCO DETERMINE Pressurizer level is NOT below programmed level. [Step 4.5]

NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 5 Page 18 of 36 Event

Description:

Steam Generator Tube Leak Time Position Applicants Actions or Behavior ATCO DETERMINE RCS leakage rate per IC-17, RCS Manual Leak Rate and/or Calculation. [Step 4.6]

BOPO CRS DETERMINE RCS leak rate is NOT greater than 40 gpm. [Step 4.7]

Booth Operator: When contacted as Shift Chemist, WAIT 10 minutes, then REPORT initial Steam Generator RC-2B leak rate is greater than 150 gallon per day.

DIRECT Shift Chemist to verify primary to secondary leak rate < 1 gpd per CRS CH-AD-0007, Primary to Secondary Leak Rate Determination. [Step 4.8]

• [CA] If primary to secondary leak rate is > 1 gpd, IMPLEMENT Attachment B, Primary to Secondary Leak Rate Actions.

Examiner Note: The following steps are from AOP-22, Reactor Coolant Leak, Attachment B, Primary to Secondary Leak Rate Actions.

CRS IMPLEMENT SO-G-105, Steam Generator Tube Leakage. [Step 1]

Booth Operator: When contacted, REPORT Work Week Manager will implement SO-G-105.

Continuously MONITOR count rate trends for radiation monitors RM-054A, ATCO RM-054B and RM-057 on ERF Computer System. [Step 2]

PERFORM the following to PLACE RM-064, Main Steam Line Radiation ATCO Monitor, in service at AI-33C: [Step 3]

• PLACE Main Steam Line A/B Enable Switch for HCV-921 and HCV-922 in ON. [Step 3.a]

• PLACE Main Steam Line A/B Mode Selector Switch in AUTO. [Step 3.b]

CRS PERFORM the following to IDENTIFY SG with tube leak: [Step 4]

CRS

• DIRECT Shift Chemist to continue sampling. [Step 4.a]

CRS

• REVIEW Steam Generator analysis for activity. [Step 4.b]

• MONITOR RM-057 & RM-064, Steam Line Radiation Monitors and ATCO DETERMINE both radiation levels RISING. [Step 4.c]

NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 5 Page 19 of 36 Event

Description:

Steam Generator Tube Leak Time Position Applicants Actions or Behavior

• MONITOR RM-054A & RM-054B, SG Blowdown Radiation Monitors and ATCO DETERMINE RM-054B is RISING [Step 4.d]

BOPO

• MONITOR SG levels and DETERMINE no apparent change. [Step 4.e]

Booth Operator: When contacted, EXECUTE remote functions to position HC-2509 / HC-2508 /

FW-268 / FW-266 as required.

Direct Equipment Operators to PERFORM the following to MINIMIZE spread CREW of contamination: [Step 5]

• ENSURE HC-2509, SAMPLE DRAIN TO DRAIN HEADER, is OPEN at AI-107 in Room 60. [Step 5.a]

• ENSURE HC-2508, SAMPLE DRAIN TO CONDENSER C.W. TUNNEL, is CLOSED at AI-107 in Room 60. [Step 5.b]

• ENSURE FW-268, CONDENSATE DUMP VALVE LCV-1193 OUTLET ISOLATION VALVE, is CLOSED at Turbine Building Mezzanine.

[Step 5.c]

• ENSURE FW-266, CONDENSATE DUMP VALVE LCV-1193 BYPASS VALVE, is CLOSED at Turbine Building Mezzanine. [Step 5.d]

CRS

• DETERMINE SG RC-2A is least affected Steam Generator. [Step 5.e]

• DETERMINE SG RC-2B is most affected Steam Generator and BOPO PERFORM the following: [Step 5.f]

• PLACE YCV-1045B, RC-2B to FW-10 Isolation Valve in OVERRIDE.

• PLACE YCV-1045B, RC-2B to FW-10 Isolation Valve to CLOSE.

• CONSIDER stopping Turbine Building Sump Pumps VD-1A & VD-1B.

CRS

[Step 5.g]

CRS

• CONSIDER isolating steam generator blowdown. [Step 5.h]

BOPO

• PLACE RCV-978, 6th Stage Extraction Isolation Valve to STOP. [Step 5.i]

Booth Operator: When contacted, EXECUTE remote function to align Condenser Evacuation Discharge to Auxiliary Building Stack.

• CONTACT Auxiliary Operator to ALIGN Condenser Evacuation CRS Discharge to Auxiliary Building stack per OI-CE-1, Condenser Evacuation System Normal Operation. [Step 5.j]

• DIRECT Radiation Protection to develop a method for processing CRS contaminated Condensate. [Step 5.k]

NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 5 Page 20 of 36 Event

Description:

Steam Generator Tube Leak Time Position Applicants Actions or Behavior CRS DETERMINE primary to secondary leakage greater than 5 gpd. [Step 6]

CRS DETERMINE primary to secondary leakage greater than 30 gpd. [Step 7]

DETERMINE primary to secondary leakage greater than 30 gpd independent CRS of Xe-133 concentration. [Step 8]

DETERMINE primary to secondary leakage greater than 75 gpd independent CRS of Xe-133 concentration. [Step 9]

DETERMINE primary to secondary leakage greater than 75 gpd with a rate CRS increase greater than 30 gpd in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period. [Step 10]

DETERMINE primary to secondary leakage greater than 75 gpd with a rate CRS increase greater than 30 gpd in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. [Step 11]

DETERMINE primary to secondary leak rate greater than 150 gpd (0.10 gpm)

CRS and PERFORM the following: [Step 12]

• ISOLATE blowdown from SG RC-2B per HR-21, Blowdown Operation.

[Step 12.a]

• COMMENCE a Plant Shutdown to MODE 4 per AOP-05, Emergency Shutdown. [Step 12.b]

CRS EVALUATE Technical Specification LCO 2.1, Reactor Coolant System.

• LCO 2.1.4 - Reactor Coolant System Leakage Limits

• CONDITION LCO 2.2.4.(1).(d) - RCS operational LEAKAGE shall be limited to 150 gallons per day primary to secondary LEAKAGE through anyone Steam Generator.

• ACTION LCO 2.1.4.(3) - Primary to secondary LEAKAGE is not within limits, then be in MODE 3, Hot Shutdown, within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> AND be in MODE 4, Cold Shutdown within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

When Technical Specifications have been addressed, PROCEED to Event 6.

NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 6 Page 21 of 36 Event

Description:

Commence Plant Shutdown Time Position Applicants Actions or Behavior Booth Operator: When directed, RESPOND to requests from Control Room.

Indications Available:

NONE Examiner Note: The following steps are from AOP-05, Emergency Shutdown.

NOTE TDB-III-23a and the Power Ascension/Power Reduction Strategy (PAPRs) provide guidance for the shutdown.

CRS CONTACT Reactor Engineer if additional guidance is required. [Step 4.1]

NOTE Operation of more than one Charging Pump will raise the rate of the power reduction.

Examiner Note: Unless directed, boration will occur from the Safety Injection Refueling Water Tank (SIRWT) when in AOP-05 to avoid time constraints.

If borating from SIRWT, COMMENCE boration by performing the following:

CRS

[Step 4.2]

• DETERMINE Charging Pump, CH-1B RUNNING.

ATCO

[Step 4.2.a]

• OPEN LCV-218-3, Charging Pump Suction SIRWT Isolation Valve.

ATCO

[Step 4.2.b]

ATCO

• CLOSE LCV-218-2, VCT Outlet Valve. [Step 4.2.c]

CRS DETERMINE Boration alignment from CVCS NOT required. [Step 4.3]

CRS NOTIFY Energy Marketing of power reduction. [Step 4.4]

NOTE During the power reduction, maintain TC PER TDB Figure III.1, Tave Program.

MAINTAIN RCS Temperature Control via Turbine Load per HR-12, BOPO Secondary Heat Removal Operation: [Step 4.5]

NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 6 Page 22 of 36 Event

Description:

Commence Plant Shutdown Time Position Applicants Actions or Behavior

• MAINTAIN TCOLD 527°F to 547°F AND

• MAINTAIN TCOLD +0°F to -1°F of program.

MAINTAIN Pressurizer Level via Charging and Letdown per IC-11, Inventory ATCO Control: [Step 4.6]

• MAINTAIN Pressurizer Level 45% to 60% AND

• MAINTAIN Pressurizer Level within 4% of program.

PERFORM the following to MAINTAIN VCT level between 55% and 85%:

ATCO

[Step 4.7]

• As required, PLACE LCV-218-1, VCT Inlet Valve to RWTS. [Step 4.7.a]

• When diversion is complete, PLACE LCV-218-1, VCT Inlet Valve to AUTO. [Step 4.7.b]

PERFORM the following to MAXIMIZE Pressurizer Heaters and Spray:

ATCO

[Step 4.8]

• As required, PLACE Backup Heater Control Switches to ON. [Step 4.8.a]

• ADJUST PC-103X or PC-103Y, Pressurizer Pressure Controller Setpoint Pushbutton to maintain pressure between 2080 psia and 2145 psia.[Step 4.8.b]

CAUTION Do not insert CEAs below power dependent insertion limit.

As required, ADJUST Regulating Group 4 to CONTROL ASI per OI-RR-1, ATCO Attachment 4, Axial Shape Index (ASI) Control. [Step 4.9]

Examiner Note: The following steps are from HR-12, Secondary Heat Removal Operation.

BOPO ENSURE Turbine Control is in MANUAL. [Step 1]

NOTE Output will be highlighted by a yellow box when selected.

NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 6 Page 23 of 36 Event

Description:

Commence Plant Shutdown Time Position Applicants Actions or Behavior BOPO PUSH the OUT button to select OUTPUT. [Step 2]

NOTES

1. Depressing the single arrow will adjust turbine load by 0.1%. Depressing the double arrow will adjust turbine load by 0.5%.

2. Tc should be maintained within (+)0°F, (-)1°F of program per TDB-III.1, Tave Program.

PRESS single or double UP[] or DOWN[] arrow to maintain Turbine BOPO Load: [Step 3]

• MAINTAIN TCOLD 527°F to 547°F.

• MAINTAIN TCOLD +0°F to -1°F of program.

Examiner Note: Do not proceed to the next event during electrical plant realignment to 161KV.

When Reactor power is reduced 3% to 5%, PROCEED to Events 7 and 8.

NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 7&8 Page 24 of 36 Event

Description:

Steam Generator Tube Rupture / Diesel Generator Start Failure Time Position Applicants Actions or Behavior Booth Operator: When directed, EXECUTE Events 7 and 8.

- Steam Generator RC-2B Tube Rupture @ 500 gpm on 10 minute ramp.

- Diesel Generator DG-01 start failure on SIAS.

Indications Available:

Pressurizer pressure and level lowering.

RECOGNIZE Pressurizer pressure and level lowering, upward trending

+2 min ATCO Radiation Monitors and MANUALLY TRIP Reactor.

CRS DIRECT performance of EOP-00, Standard Post Trip Actions.

Examiner Note: The following steps are from EOP-00, Standard Post Trip Actions.

ATCO VERIFY Reactivity Control: [Step 1]

• VERIFY ALL of the following: [Step 1.a]

• VERIFY no more than one Regulating or Shutdown CEA NOT inserted.

• VERIFY Reactor Power is LOWERING.

• VERIFY Startup Rate is NEGATIVE.

• MONITOR plant for an uncontrolled RCS Cooldown. [Step 1.b]

CRS DETERMINE Reactivity Control criteria SATISFIED.

BOPO VERIFY Turbine Trip: [Step 2]

• VERIFY HP & LP Stop and Intercept Valves CLOSED.

BOPO ENSURE all Generator Breakers are tripped: [Step 3]

• DETERMINE Generator Output Breaker 3451-4 tripped.

• DETERMINE Generator Output Breaker 3451-5 tripped.

• DETERMINE Generator Field Breaker 41E/G1F tripped.

BOPO VERIFY 4160 V Safeguards Buses 1A3 and 1A4 are ENERGIZED. [Step 4]

NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 7&8 Page 25 of 36 Event

Description:

Steam Generator Tube Rupture / Diesel Generator Start Failure Time Position Applicants Actions or Behavior CRS DETERMINE Maintenance of Vital Auxiliaries criteria SATISFIED.

Examiner Note: The following step (Verify Diesel Generators running) is not required until Reactor Coolant System Pressure is less than 1600 psia and PPLS has actuated.

VERIFY both Diesel Generators RUNNING on Safety Injection Actuation BOPO Signal. [Step 5]

• [CA] DEPRESS DG-01 Emergency Start pushbutton and VERIFY DG-01 running at 900 RPM.

VERIFY 4160 V Non-Safeguards Buses 1A1 and 1A2 are ENERGIZED.

BOPO

[Step 6]

BOPO VERIFY 125 VDC Buses 1 and 2 are ENERGIZED. [Step 7]

BOPO VERIFY Instrument Air is AVAILABLE: [Step 8]

• DETERMINE Instrument Air pressure 90 psig.

• DETERMINE Instrument Air Compressor CA-1C RUNNING.

ATCO VERIFY Component Cooling Water System operation NORMAL: [Step 9]

• DETERMINE at least one CCW pump RUNNING. [Step 9.a]

• DETERMINE CCW Pump discharge pressure 60 psig. [Step 9.b]

• DETERMINE HCV-438A/B/C/D, CCW to RCP Coolers OPEN. [Step 9.c]

• DETERMINE at least one Raw Water Pump RUNNING. [Step 9.d]

CRS VERIFY RCS Inventory Control criteria satisfied: [Step 10]

• DETERMINE PZR level NOT between 30% and 70% and NOT TRENDING to between 45% and 60%.

• [CA] RESTORE Inventory Control by manually controlling Charging and Letdown. [Step 10.1.a]

• DETERMINE RCS subcooling > 20°F.

NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 7&8 Page 26 of 36 Event

Description:

Steam Generator Tube Rupture / Diesel Generator Start Failure Time Position Applicants Actions or Behavior CRS DETERMINE RCS Inventory Control criteria NOT SATISFIED.

CRS VERIFY RCS Pressure Control criteria satisfied: [Step 11]

ATCO

• DETERMINE RCS pressure less than 1600 psia.

• [CA] VERIFY RCS pressure < 2300 psia and PORV NOT open.

[Step 11.1]

• [CA] When RCS pressure < 1350 psia, PERFORM the following:

[Step 11.2]

ATCO * [CA] STOP one RCP in each Loop.

• [CA] DETERMINE RCS pressure < 1600 psia and VERIFY Engineered Safeguards ACTUATED. [Step 11.3]

• [CA] DETERMINE PPLS relays 86A/PPLS / 86B/PPLS /

86A1/PPLS / 86B1/PPLS have TRIPPED.

[Step 11.3.a]

• [CA] DETERMINE all PPLS relays have TRIPPED.

[Step 11.3.b]

• [CA] DETERMINE VIAS relays 86A/VIAS / 86B1/VIAS /

86B/VIAS / 86A1/VIAS have TRIPPED. [Step 11.3.c]

• [CA] DETERMINE SIAS relays 86A/SIAS / 86AX/SIAS /

86B1/SIAS / 86B1X/SIAS / 86B/SIAS / 86BX/SIAS /

86A1/SIAS / 86A1X/SIAS have TRIPPED.

[Step 11.3.d]

• [CA] DETERMINE CIAS relays 86A/CIAS / 86B1/CIAS /

86B/CIAS / 86A1/CIAS have TRIPPED. [Step 11.e]

• [CA] ENSURE required pumps RUNNING [Step 11.3.f]

• DETERMINE HPSI Pumps SI-2A & SI-2B RUNNING.

• DETERMINE LPSI Pumps SI-1A and SI-1B RUNNING.

• DETERMINE Charging Pumps CH-1B RUNNING.

• [CA] ENSURE acceptable SI flow per Attachment IC-13, SI Flow vs. Pressurizer Pressure. [Step 11.3.g]

• [CA] ENSURE Emergency Boration in progress.

ATCO

[Step 11.3.h]

CRS DETERMINE RCS Pressure Control criteria NOT SATISFIED.

NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 7&8 Page 27 of 36 Event

Description:

Steam Generator Tube Rupture / Diesel Generator Start Failure Time Position Applicants Actions or Behavior Examiner Note: The following steps are from RC-11, Emergency Boration Verification.

ATCO ENSURE the following valves are CLOSED: [Step 1]

• FCV-269X, Demin Water Makeup Valve

• FCV-269Y, Boric Acid Makeup Valve

• HCV-264, CH-4A Recirc Valve

• HCV-257, CH-4B Recirc Valve ATCO VERIFY all the following valves OPEN: [Step 2]

• HCV-268, Boric Acid Pump Header to Charging Pumps Isolation Valve

• HCV-265, CH-11A Gravity Feed Valve

• HCV-258, CH-11B Gravity Feed Valve ATCO ENSURE all available Boric Acid Pumps RUNNING: [Step 3]

• CH-4A, Boric Acid Pump

• CH-4B, Boric Acid Pump ATCO ENSURE all available Charging Pumps RUNNING: [Step 4]

• CH-1A, Charging Pump is tripped.

• CH-1B, Charging Pump is RUNNING.

• CH-1C, Charging Pump is OOS.

ATCO ENSURE the following valves are CLOSED: [Step 5]

• LCV-218-2, VCT Outlet Valve

• LCV-218-3, Charging Pump Suction SIRWT Isolation Valve

• HCV-257, CH-4B Recirculation Valve

• HCV-264, CH-4A Recirculation Valve ATCO DETERMINE Emergency Boration is in progress. [Step 6]

NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 7&8 Page 28 of 36 Event

Description:

Steam Generator Tube Rupture / Diesel Generator Start Failure Time Position Applicants Actions or Behavior Examiner Note: The following steps continue from EOP-00, Standard Post Trip Actions.

CRS VERIFY Core Heat Removal criteria satisfied: [Step 12]

ATCO

• DETERMINE RCP NPSH requirements met.

• DETERMINE at least one RCP operating.

• DETERMINE Core T 10°F.

Stop One Reactor Coolant Pump in Each Loop when Reactor Coolant System CRITICAL TASK Pressure is < 1350 psia, Prior to losing Reactor Coolant Pump Net Positive STATEMENT Suction Head.

CRITICAL DETERMINE Reactor Coolant System pressure < 1350 psia and PERFORM TASK ATCO the following:

ATCO

• STOP one RCP in each Loop.

CRS DETERMINE Core Heat Removal criteria SATISFIED.

NOTE If Instrument Air to valves HCV-1105, HCV-1106, HCV-1107A/B and HCV-1108A/B is not available, throttling of these valves is not possible. Open or close operation of these valves is possible for a minimum of three cycles.

CRS VERIFY RCS Heat Removal criteria satisfied:

VERIFY Main Feedwater is restoring SG levels to 35% to 80% NR and 73%

BOPO to 94% WR. [Step 13]

• DETERMINE FCV-1101 & FCV-1102 Feed Regulating Valves CLOSED.

[Step 13.a]

• DETERMINE HCV-1105 & HCV-1106 Feed Regulating Bypass Valves ramped to between 40% & 45% OPEN. [Step 13.b]

BOPO

• PLACE 43/FW switch in OFF. [Step 13.c]

• ENSURE no more than one Main Feedwater Pump RUNNING.

[Step 13.d]

• ENSURE no more than one Condensate Pump RUNNING. [Step 13.e]

NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 7&8 Page 29 of 36 Event

Description:

Steam Generator Tube Rupture / Diesel Generator Start Failure Time Position Applicants Actions or Behavior

• STOP running Heater Drain Pumps FW-5A, FW-5B, and/or FW-5C.

BOPO

[Step 13.f]

• ENSURE SG Blowdown Isolation Valves CLOSED. [Step 13.g]

• HCV-1387A & HCV-1387B

• HCV-1388A & HCV-1388B VERIFY Steam Dump and Bypass Valves controlling both of the following:

BOPO

[Step 14]

• DETERMINE RCS TCOLD between 525°F and 535°F.

• DETERMINE Steam Generator pressure between 850 psia & 925 psia.

CRS DETERMINE RCS Heat Removal criteria SATISFIED.

CRS VERIFY Containment Isolation criteria satisfied:

ATCO VERIFY Normal Containment conditions exist: [Step 15]

• DETERMINE no unexpected rise in Containment Sump level. [Step 15.a]

• DETERMINE Containment Area Radiation Monitors NOT in alarm.

[Step 15.b]

• DETERMINE Containment Ventilation Radiation Monitors NOT in alarm.

[Step 15.c]

• DETERMINE RM-054B, SG Blowdown Radiation Monitor ALARMING.

ATCO

[Step 15.d]

• [CA] MINIMIZE spread of contamination: [Step 15.d.1]

• [CA] VERIFY RCV-978, 6th Stage Extraction Isolation Valve BOPO CLOSED. [Step 15.d.1.1)]

• [CA] VERIFY all Blowdown Isolation Valves CLOSED.

[Step 15.d.1.2)]

• [CA] HCV-1387A & HCV-1387B

• [CA] HCV-1388A & HCV-1388B

• DETERMINE RM-054B, SG Blowdown Radiation Monitor and RM-057, ATCO Condenser Off Gas Radiation Monitor TRENDING upward. [Step 15.e]

• [CA] DETERMINE if Steam Generator Tube Rupture is in progress:

CRS

[Step 15.e.1]

NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 7&8 Page 30 of 36 Event

Description:

Steam Generator Tube Rupture / Diesel Generator Start Failure Time Position Applicants Actions or Behavior

• [CA] DIRECT Shift Chemist to perform rapid activity analysis of both SGs. [Step 15.e.1.1)]

• [CA] DETERMINE SG RC-2B has an abnormal rise in level.

BOPO

[Step 15.e.1.2)]

ATCO

• VERIFY Containment conditions: [Step 15.f]

• DETERMINE Containment pressure < 3 psig.

• DETERMINE Containment temperature < 120°F.

CRS DETERMINE Containment Integrity criteria NOT SATISFIED.

CRS DIAGNOSE event in progress: [Step 16]

• DETERMINE Reactivity Control requirements met.

• DETERMINE both DC buses energized.

• DETERMINE at least one Vital 4160 V Bus energized.

• DETERMINE at least one Non-Vital 4160 V Bus energized.

• DETERMINE at least one RCP running.

• DETERMINE adequate Feedwater flow to at least one SG.

• VERIFY Pressurizer pressure > 1800 psia with high subcooled margin, normal SG pressure, and no indications of primary to secondary leakage.

• If not, CONSIDER EOP-04, Steam Generator Tube Rupture.

NOTE Certain events (i.e., LOCA, SGTR, UHE and Loss of All Feedwater) do not require offsite power in order to adequately, mitigate the effects of the accident. For this reason, the LOCA, SGTR, UHE or Loss of All Feedwater procedure may be implemented even if a Loss of Offsite Power has also occurred.

• DETERMINE all Safety Function Acceptance Criteria NOT SATISFIED.

• DETERMINE single event in progress and TRANSITION to EOP-04, Steam Generator Tube Rupture.

Examiner Note: The following steps are from EOP-04, Steam Generator Tube Rupture.

CRS CONFIRM Standard Post Trip Actions have been performed. [Step 1]

NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 7&8 Page 31 of 36 Event

Description:

Steam Generator Tube Rupture / Diesel Generator Start Failure Time Position Applicants Actions or Behavior CRS CONFIRM Steam Generator Tube Rupture Diagnosis: [Step 2]

• VERIFY Safety Function Status Check Acceptance Criteria being satisfied. [Step 2.a]

• VERIFY CIAS is present and SAMPLE both SGs. [Step 2.c]

CRS IMPLEMENT the Emergency Plan. [Step 3]

• Time: __________

CREW MONITOR the Floating Steps. [Step 4]

DETERMINE RCS pressure 1600 psia and VERIFY Engineered CRS Safeguards are ACTUATED: [Step 5]

• DETERMINE PPLS relays 86A/PPLS / 86B/PPLS / 86A1/PPLS /

86B1/PPLS have TRIPPED. [Step 5.a]

• DETERMINE SIAS relays 86A/SIAS / 86AX/SIAS / 86B1/SIAS /

86B1X/SIAS / 86B/SIAS / 86BX/SIAS / 86A1/SIAS / 86A1X/SIAS have TRIPPED. [Step 5.b]

• DETERMINE CIAS relays 86A/CIAS / 86B1/CIAS / 86B/CIAS /

86A1/CIAS relays TRIPPED. [Step 5.c]

• DETERMINE VIAS relays 86A/VIAS / 86B1/VIAS / 86B/VIAS /

86A1/VIAS relays TRIPPED. [Step 5.d]

OPTIMIZE Safety Injection and Charging flow and PERFORM the following:

ATCO

[Step 6]

• ENSURE required Safety Injection Pumps RUNNING: [Step 6.a]

• DETERMINE HPSI Pumps SI-2A & SI-2B RUNNING.

• DETERMINE LPSI Pumps SI-1A and SI-1B RUNNING.

• DETERMINE Charging Pumps CH-1B RUNNING.

• DETERMINE Emergency Boration already in progress per RC-11, ATCO Emergency Boration Verification. [Step 6.b]

• ENSURE adequate SI flow per IC-13, Safety Injection Flow vs.

Pressurizer Pressure. [Step 6.c]

NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 7&8 Page 32 of 36 Event

Description:

Steam Generator Tube Rupture / Diesel Generator Start Failure Time Position Applicants Actions or Behavior NOTE Main PZR Spray flow will be reduced with less than four-pump operation. Pressure should be controlled using Main and Auxiliary PZR Spray whenever the Plant is placed in a two-pump configuration.

ATCO VERIFY RCP operating parameters: [Step 7]

• ENSURE at least one RCP stopped if TCOLD < 500°F. [Step 7.a]

• DETERMINE one RCP stopped in each loop when RCS pressure 1350 psia following SIAS. [Step 7.b]

• DETERMINE all RCPs STOPPED on low subcooling. [Step 7.c]

• Time: _______

DETERMINE Condenser vacuum greater than 10.92 inches Hg absolute or CRS 19 inches Hg. [Step 8]

NOTE Reducing RCS TH to less than or equal to 510°F will maintain adequate RCP NPSH and RCS subcooling when RCS pressure is reduced below SG safety valve setpoint of 1000 psia.

CAUTION When TC is 178°F or greater, the maximum RCS cooldown rate is 100°F/hr.

When TC is less than 178°F, the maximum RCS cooldown rate is 50°F/hr.

COMMENCE a cooldown using both SGs to reduce RCS THOT to 510°F per BOPO Attachment HR-12, Secondary Heat Removal Operation. [Step 9]

COMMENCE a depressurization of RCS to less than 1000 psia per ATCO Attachment PC-11, Pressure Control. [Step 10]

Examiner Note: The following steps are from HR-12, Secondary Heat Removal Operation.

BOPO ENSURE Turbine Control is in MANUAL. [Step 1]

BOPO * [CA] DETERMINE Turbine NOT online and GO TO Step 4.

NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 7&8 Page 33 of 36 Event

Description:

Steam Generator Tube Rupture / Diesel Generator Start Failure Time Position Applicants Actions or Behavior NOTES

1. In MANUAL, single arrows are 1% / double arrows are 5% change in OUTPUT value.

2. While Steam Dump and Bypass Control is in Temperature/Pressure Mode, the controllers PC0910 and TC0909_PI will alternate between controls, depending on the higher output signal. A red square outlining the controlling function signify which parameter is in control.

3. Steps 4 through 11 may be performed as needed, and in any order.

CAUTIONS When TC is 178°F or greater, the maximum RCS cooldown rate is 100°F/hr.

When TC is less than 178°F, the maximum RCS cooldown rate is 50°F/hr.

CRITICAL TASK Reduce and Maintain RCS THOT 510°F to Maintain SG Pressure Below Safety STATEMENT Valve Setpoint of 1000 psia Prior to Isolating SG RC-2B.

CRITICAL DETERMINE Steam Dump and Bypass (SD&B) available and CONTROL TASK BOPO RCS temperature with a single SD&B Valve. [Step 4]

• DEPRESS Valve Toggle to SELECT valve to be operated: [Step 4.a]

• PCV-910 / TCV-909-1 / TCV-909-2 / TCV-909-3 / TCV-909-4

• PLACE Controller for selected valve in MANUAL. [Step 4.b]

• PUSH UP and DOWN arrows to ADJUST Controller Output. [Step 4.c]

• When no longer required, PLACE Controller for selected valve in AUTO.

[Step 4.d]

Examiner Note: The following steps are from PC-11, Pressure Control. PC-12, RCS Pressure-Temperature Limits (graph), is maintained on a Control Room hardcopy.

CAUTION A charging header flow path must be maintained at all times.

MAINTAIN RCS pressure per the PC-12, RCS Pressure-Temperature Limits ATCO graph: [Step 1]

• DETERMINE Steps N/A due to RCS pressure. [Step 1.a to 1.d]

NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 7&8 Page 34 of 36 Event

Description:

Steam Generator Tube Rupture / Diesel Generator Start Failure Time Position Applicants Actions or Behavior CRITICAL

• OPERATE the following to CONTROL Auxiliary Spray flow and TASK ATCO REDUCE RCS pressure to < 1000 psia: [Step 1.e]

• HCV-240, PZR Auxiliary Spray Isolation Valve

• HCV-249, PZR Auxiliary Spray Isolation Valve

• HCV-238, Loop 1 Charging Isolation Valve

• HCV-239, Loop 2 Charging Isolation Valve

• If HPSI Stop and Throttle criteria is met, CONTROL Pressurizer level using Charging, Letdown, and/or HPSI flow per IC-11, Inventory Control.

[Step 1.f]

• CONTROL RCS heat removal per HR-12, Secondary Heat Removal Operation. [Step 1.g]

MAINTAIN RCS Pressure per PC-12, RCS Pressure-Temperature Limits by ATCO performing ANY of the following: [Step 11]

• CONTROL RCS Heat Removal per HR-12, Secondary Heat Removal Operation. [Step 11.a]

• CONTROL Pressurizer Heaters and Spray per PC-11 Pressure Control.

[Step 11.b]

• If HPSI Stop and Throttle criteria are met, CONTROL Charging, Letdown, and/or HPSI flow per IC-11, Inventory Control. [Step 11.c]

If feeding through Feed Ring, MAINTAIN SG levels 44% to 85% NR (77% to BOPO 94% WR) using Main Feedwater or FW-54. [Step 12]

• FEED SGs using HR-15, Main Feed Pump Operation or HR-16, FW-54 Operation. [Step 12.a]

• CONTROL feed flow per HR-11, Manual Feed Control. [Step 12.b]

PERFORM the following to PLACE RM-064, Main Steam Line Radiation ATCO Monitor in service at AI-33C. [Step 13]

• PLACE Main Steam Line A/B Enable Switch for HCV-921 and HCV-922 in ON. [Step 13.a]

• PLACE Main Steam Line A/B Mode Selector Switch in AUTO. [Step 13.b]

CRS DETERMINE Steam Generator RC-2B has the tube rupture. [Step 14]

BOPO PERFORM the following to MINIMIZE spread of contamination: [Step 15]

NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 7&8 Page 35 of 36 Event

Description:

Steam Generator Tube Rupture / Diesel Generator Start Failure Time Position Applicants Actions or Behavior

• CONTACT Auxiliary Operator to POSITION following valves: [Step 15.a]

• ENSURE HC-2509, SAMPLE DRAIN TO DRAIN HEADER is OPEN at AI-107, Room 60. [Step 15.a]

• ENSURE HC-2508, SAMPLE DRAIN TO CONDENSER C.W.

TUNNEL is CLOSED at AI-107, Room 60. [Step 15.b]

• ENSURE FW-268, CONDENSATE DUMP VALVE LCV-1193 OUTLET ISOLATION VALVE, is CLOSED at Turbine Building Mezzanine. [Step 15.c]

• ENSURE FW-266, CONDENSATE DUMP VALVE LCV-1193 BYPASS VALVE, is CLOSED at Turbine Building Mezzanine.

[Step 15.d]

BOPO When RCS THOT is 510°F, ISOLATE SG RC-2B. [Step 16]

• ISOLATE SG RC-2B per HR-20, Isolate/Restore Steam Generator B.

[Step 16.a]

Examiner Note: The following steps are from HR-20, Isolate/Restore Steam Generator B.

NOTE RCS Heat Removal takes precedence over isolation of a S/G with a tube rupture.

CRITICAL TASK Isolate the Affected Steam Generator with a Tube Rupture to Minimize Spread STATEMENT of Contamination Prior to Exiting EOP-04, Steam Generator Tube Rupture.

CRITICAL TASK BOPO PERFORM the following to isolate Steam Generator RC-2B: [Step 1]

• ENSURE all the following valves are CLOSED: [Step 1.a]

• CLOSE HCV-1042A, RC-2B MSIV.

• VERIFY HCV-1042C, RC-2B MSIV Bypass Valve CLOSED.

• VERIFY MS-292, Air Assisted Main Steam Safety Valve CLOSED.

• CLOSE FCV-1102, RC-2B Feed Regulating Valve.

• CLOSE HCV-1106, Feed Regulating Bypass Valve.

BOPO

• CLOSE HCV-1385, RC-2B Feed Header Isolation Valve.

• CLOSE HCV-1104, Feed Regulating Block Valve.

• VERIFY HCV-1387A, Blowdown Isolation Valve CLOSED.

• VERIFY HCV-1387B, Blowdown Isolation Valve CLOSED.

• CLOSE HCV-1108A, AFW Isolation Valve.

• CLOSE HCV-1108B, AFW Isolation Valve.

NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 7&8 Page 36 of 36 Event

Description:

Steam Generator Tube Rupture / Diesel Generator Start Failure Time Position Applicants Actions or Behavior

• CONTACT Auxiliary Operator to CLOSE MS-298, Steam Valves HCV-1041A & 1042A Packing Leakoff Line Isolation Valve in Room 81.

[Step 1.b]

• If sampling is NOT in progress, CLOSE both Sample Valves:

[Step 1.c]

• HCV-2507A, RC-2B Blowdown Sample Isolation Valve

• HCV-2507B, RC-2B Blowdown Sample Isolation Valve BOPO

• PERFORM the following to CLOSE YCV-1045B: [Step 1.d]

• DETERMINE Isolation Valve YCV-1045B OVERRIDE SW in OVERRIDE. [Step 1.d.1)]

• DETERMINE SG RC-2B STM TO FW-10 HDR A ISOLATION VALVE YCV-1045B in CLOSE. [Step 1.d.2)]

NOTE Air accumulators will maintain the valve in a closed position for 30 minutes after a loss of Instrument Air.

• CONTACT Auxiliary Operator to HANDJACK YCV-1045B, MAIN STEAM LINE "B" TO AUX FEEDWATER PUMP FW-10 SUPPLY VALVE to CLOSE in Room 81. [Step 1.d.3)]

CRS

• RECORD time Steam Generator RC-2B ISOLATED. [Step 1.e]

• Time: ________

VERIFY RC-2B is most affected SG per Attachment HR-18, Most Affected CRS Steam Generator Determination. [Step 2]

When Steam Generator RC-2B is isolated, TERMINATE the scenario.

NRC Simulator Scenario 3 Outline Rev. Final

Appendix D Scenario Outline Form ES-D-1 Facility: Fort Calhoun Station Scenario No.: 4 Op Test No.: Dec 2015 NRC Examiners: Operators:

Initial Conditions: MODE 2 at ~1% power - RCS Boron is 959 ppm (by sample).

Turnover: Continue in OP-2A, Plant Startup and OI-RR-1, Reactor Regulating System Operation to raise Reactor power to 7% power. When MODE 1 is entered, place Steam Dump and Bypass Valves in AUTO per OI-MS-1A, Main Steam System Operation.

Critical Tasks:

• Upon Loss of Cooling to Reactor Coolant Pumps (RCPs), Trip the Reactor and Trip Associated RCP(s) within 5 minutes of Total Loss of CCW flow or Motor Bearing Temperatures > 203°F but prior to exceeding 210°F or within 5 minutes of CCW temperature

> 110°F. (Event 2). OR

• Manually Trip Reactor to meet Core Design Criteria of Lowering Reactor Power and Negative Startup Rate to Verify Reactivity Control Established During ATWS Event Prior to Exiting EOP-00, Standard Post Trip Actions. (Event 5)

• Isolate the Affected Steam Generator to Prevent Excess Plant Cooldown and Reactivity Additions Prior to Steam Generator Level = 0% Wide Range Level. (Event 7)

Event No. Malf. No. Event Type* Event Description 1 R (ATCO) Raise Power Using Control Rods to 7% per OP-2A, Plant Startup.

+20 min N (BOPO, CRS) Place Steam Dump and Bypass Valves in AUTO per OI-MS-1A.

2 C (ATCO, CRS) Raw Water Pump Discharge Line Leak Upstream of HCV-2879A in

+30 min TS (CRS) the Auxiliary Building.

3 I (BOPO, CRS) Inadvertent Channel B Auxiliary Feedwater Actuation Signal On

+45 min TS (CRS) Steam Generator RC-2A.

4 C (ATCO, CRS) Loss of Instrument Bus AI-40A.

+60 min TS (CRS) Loss of Letdown and Pressurizer Level Control.

(Alternate Path Event 8) 5 M (ATCO, BOPO, Reactor Coolant Pump RC-3A Trip.

+60 min CRS) Automatic Reactor Trip Failure, Manual Reactor Trip Required.

6 C (BOPO) Instrument Air Compressor CA-1B and CA-1C Trip.

+65 min Bearing Cooling Water Pump AC-9B Trip.

7 M (ATCO, BOPO, Steam Line Break inside Containment on RC-2A @ 0.65% Severity

+70 min CRS) on 5 Minute Ramp.

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Technical Specifications Actual Target Quantitative Attributes 1 Malfunctions after EOP entry (1-2) 3 Abnormal events (2-4) 2 Major transients (1-2) 1 EOPs entered/requiring substantive actions (1-2) 0 EOP contingencies requiring substantive actions (0-2) 2 Critical tasks (2-3)

NRC Simulator Scenario 4 Outline Rev. Final As Run

Scenario Event Description NRC Scenario 4 SCENARIO

SUMMARY

NRC 4 The crew will assume the shift at 1% power and raise Power to ~7% using CEAs per OP-2A, Plant Startup, Attachment 4, Hot Standby, MODE 2 to Minimum Load, MODE 1 and OI-RR-1, Reactor Regulating System Normal Operation, Attachment 7, Manual Sequential Mode Checklist. When MODE 1 is entered, temperature control is placed in AUTO per OI-MS-1A, Main Steam System Operation, , Steam Dump and Bypass Manual Control Function.

The next event is a Raw Water Pump AC-10C discharge line leak in the Auxiliary Building upstream of HCV-2879A. The crew enters AOP-18, Loss of Raw Water, and must observe Raw Water System indications in order to determine the location of the leak. Once identified, the leak is isolated per AOP-18, Attachment C, Equipment Isolation, and Raw Water flow is restored. If the leak is not isolated, the Reactor and affected RCPs will be tripped. The SRO will refer to Technical Specification LCO 2.4(1)

- Raw Water Header.

The next event is an inadvertent Channel B Auxiliary Feedwater Actuation Signal (AFAS) on Steam Generator RC-2A. The crew responds per ARP-AI-66B/A66B, Window 41 and verifies Auxiliary Feedwater Pumps FW-6 and FW-10 are running. Once it is determined the AFAS was inadvertent, AOP-23, Reset of Engineered Safeguards,Section IX, Reset of Inadvertent AFAS, is performed. The SRO will refer to Technical Specification LCO 2.15.1(1) - Automatic Initiation Steam Generator Water Level Logic Subsystem B.

When plant conditions are stable, a loss of Instrument Bus AI-40A occurs. The crew enters AOP-16, Loss of Instrument Bus Power,Section I, Loss of Instrument Bus Power, then Section II, Loss of Instrument Bus AI-40A. Actions include isolating Letdown, transferring Pressurizer Level Control, and operating Charging Pumps as required. Electrical Maintenance is notified and the Plant remains in this configuration through the end of the Scenario. The SRO will refer to Technical Specification LCO 2.15.2

- Reactor Protective System Logic and Trip Initiation and LCO 2.7(1) - 120 VAC Instrument Bus A.

The next event is a trip of Reactor Coolant Pump RC-3A. The crew should recognize failure of the Reactor Protection System Low Flow trips and manually trip the Reactor and enter EOP-00, Standard Post Trip Actions. When the Reactor is tripped, a Steam Line Break inside Containment initiates on a 5 minute ramp. Due to the small size of this break, RCS pressure remains above the SIAS initiation setpoint of 1600 psia. The crew will transition to EOP-05, Uncontrolled Heat Extraction, and identify and isolate the affected Steam Generator RC-2A.

The event is complicated by a trip of the running and standby Instrument Air Compressors CA-1B and CA-1C and a trip of Bearing Water Cooling Pump AC-9B. The crew must restore a Bearing Cooling Water Pump and Instrument Air Compressor while in EOP-00. The scenario is terminated when Steam Generator RC-2A is isolated per HR-19, Isolate/Restore Steam Generator A while in EOP-05.

Risk Significance:

• Failure of risk important system prior to trip: Loss of Raw Water System Header Loss of Instrument Bus

• Risk significant core damage sequence: Automatic Reactor Trip Failure Steam Line Break Inside Containment

• Risk significant operator actions: Isolate Raw Water East Header Manually Trip Reactor Restore Instrument Air Isolate Affected Steam Generator NRC Simulator Scenario 4 Outline Rev. Final As Run

Scenario Event Description NRC Scenario 4 BOOTH OPERATOR INSTRUCTIONS for SIMULATOR SETUP RESET to IC-122 and LOAD & EXECUTE NRC 4.sce for NRC Scenario 4.

Preset item - Event 5 - Reactor Fails to Trip Automatically, CB-4 Trip Button Works Type Item Value Condition Expert RPS02 Energized Scenario Event: Rx Fail to RPS01 Energized Trip, CB-4 works RPS03 Energized RPS04 Energized P6A_026_1 True P6B_028_1 True ANN-P6A_0026R1C_Fail Alarm Off ANN-P6A_0027R1C_Fail Alarm Off ANN-P6B_0026R5C_Fail Alarm Off ANN-P6B_0027R5C_Fail Alarm Off ANN-P6B_0025R5C_Fail Alarm Off ANN-P6A_0025R1C_Fail Alarm Off H_P6A_022A_1 True H_P6B_024A_1 True Event 2 - Raw Water leak in the Auxiliary Building Type Item Value Condition Malfunction RWS02B 25 When directed by examiner, trigger/activate this event.

Scenario Event: Raw Water Leak in Aux Building Event 3 - Inadvertent AFAS on RC-2A Type Item Value Condition Expert B_RC_2A_AFWS True When directed by examiner, trigger/activate this event.

Scenario Event:

Inadvertent AFAS Event 4 - Loss of Instrument Bus AI-40A Type Item Value Condition Malfunction EDA08 10 When directed by examiner, trigger/activate this event.

Scenario Event: Loss of AI-40A Event 5 - A Reactor Coolant Pump Trips Type Item Value Condition Malfunction BUS_1A1_5_BKR_TRIP True When directed by examiner, trigger/activate this event.

Scenario Event: A RCP Trip Event 6 - Following RX Trip, Loss of Instrument Air and Bearing Cooling Water NRC Simulator Scenario 4 Outline Rev. Final As Run

Scenario Event Description NRC Scenario 4 Type Item Value Condition Remote BCW_AC9B_BRKR Trip Event is triggered Malfunction BUS_1B3A_4A_2_BKR_Trip True automatically after reactor BUS_1B4B_4_BKR_TRIP True trip. Scenario Event: Loss of Inst Air and Bearing Water Event 7 - Main Steam Break Inside Containment Type Item Value Condition Malfunction SGN01A 0.65% Event is triggered Ramp = 300 sec automatically after reactor trip. Scenario Event:

Steam Line Break in Containment NRC Simulator Scenario 4 Outline Rev. Final As Run

Scenario Event Description NRC Scenario 4 Booth Operator: INITIALIZE to IC-122 and LOAD NRC 4.sce.

ENSURE all Simulator Annunciator Alarms are ACTIVE.

ENSURE all Control Board Tags are removed.

ENSURE Bearing Water Pump AC-9B running.

ENSURE Charging Mode Select Switch is in CH-1A - CH-1C position.

ENSURE Turbine speed is approximately 3 RPM.

ENSURE Air Compressors CA-1B & CA-1C alignment: 1 in Standby, 1 running.

PLACE Steam Dump & Bypass Controllers in Manual.

ENSURE Lead Examiner has AFAS Keys 55 & 57 for Event 3.

ENSURE Lead Examiner has RPS Trip Unit Keys 1-12 for Event 4.

ENSURE Operator Aid Tags reflect current boron conditions.

ENSURE Reactivity Briefing Sheet printout provided with Turnover.

ENSURE Middle-of-Life Thumb Rule Sheet provided with Turnover.

ENSURE Steam Dump and Turbine Bypass System in MANUAL control.

ENSURE Control Room hard copy for OI-RR-1 is CLEAN.

ENSURE CEA Regulating Group 4 @ 72.

ENSURE procedures in progress provided to crew in Briefing Room:

- COPY of ReMA Data for Reactor Power Ascension.

- COPY of OP-2A, Plant Startup, Attachment 4, Hot Standby, MODE 2 to Minimum Load, MODE 1, INITIALED through Step 6.b.

- COPY of OI-RR-1, Reactor Regulating System Normal Operation, Attachment 7, Manual Sequential Mode Checklist.

- Copy of OI-MS-1A, Main Steam System Operation, Attachment 5, Steam Dump and Bypass Manual Control Function, INITIALED through Prerequisites and Steps 1.a & 2.a.

Control Room Annunciators in Alarm:

A9-B-1(U) - TURBINE DIFFERENTIAL EXPANSION A10-A-1(U) - MOTOR SUCT PUMP RUNNING OR NOT IN AUTO A10-B-6(L) - 43/FW TRANSFER SWITCH OFF-AUTO A11-A-4(U) - HEATER 5A HEATER HI-LO A11-A-4(L) - HEATER 5B HEATER HI-LO A11-B-3(U) - HEATER DRAIN TANK LEVEL HI-LO A20-D LOSS OF LOAD CHANNEL TRIP BYPASSED A20-E HIGH POWER RATE OF CHANGE TRIP ENABLED A21-B-1(U) - HC-909 INHIBIT A21-C-6(U) - HEATING STEAM PRESS LO AI-66B/A66B-Window 3 - FW-10 TURBINE DRIVEN FEEDWATER PUMP RUNNING NRC Simulator Scenario 4 Outline Rev. Final As Run

Scenario Event Description NRC Scenario 4 Procedure List Event 1: OP-2A, Plant Startup, Attachment 4, Hot Standby, MODE 2 to Minimum Load, MODE 1 Event 1: OI-RR-1, Reactor Regulating System Normal Operation, Attachment 7, Manual Sequential Mode Checklist Event 1: OI-MS-1A, Main Steam System Operation, Attachment 5, Steam Dump and Bypass Manual Control Function Event 2: AOP-18, Loss of Raw Water Event 3: ARP-AI-66B/A66B, Window 41, AFWS STEAM GEN RC-2A CHANNEL B ACTUATED Event 3: AOP-23, Reset of Engineered Safeguards,Section IX, Reset of Inadvertent AFAS Event 3: OI-AFW-2, Auxiliary Feedwater System Actuation and Bypass, Attachment 1, Bypass of the Auxiliary Feedwater Actuation Signal (AFAS) (Modes 1 or 2)

Event 3: OI-AFW-2, Auxiliary Feedwater System Bypass, Table 2, AFAS Logic Subsystem Channel Bypass Switch Alignment Event 4: AOP-16, Loss of Instrument Bus Power,Section I - Loss of Instrument Bus Power Event 4: AOP-16, Loss of Instrument Bus Power,Section II, Loss of Instrument Bus AI-40A Event 5: EOP-00, Standard Post Trip Actions Event 7: EOP-05, Uncontrolled Heat Extraction NRC Simulator Scenario 4 Outline Rev. Final As Run

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 1 Page 7 of 37 Event

Description:

Raise Reactor Power Time Position Applicants Actions or Behavior Booth Operator: When directed, RESPOND to requests from Control Room.

Examiner Note: This Scenario Section contains guidance for the following Operator actions:

1. Raising power per OP-2A.

2. Withdrawing Control Rods per OI-RR-1.

3. Control of Steam Dumps and Bypass per OI-MS-1A.

Examiner Note: The following steps are from OP-2A, Plant Startup, Attachment 4, Hot Standby, MODE 2 to Minimum Load, MODE 1, Step 6.

RAISE Reactor power to ~ 10% while performing the following: [Step 6]

• DETERMINE Main Feedwater Pump FW-4B is RUNNING. [Step 6.a]

• DETERMINE Auxiliary Feedwater SECURED. [Step 6.b]

• MAINTAIN RCS temperature 527°F to 535°F using Steam Dump and Bypass Valves. [Step 6.c]

• Prior to exceeding 15% power, VERIFY Secondary Chemistry parameters. [Step 6.d]

• Prior to exceeding 15% power, VERIFY Condensate Pump Discharge Suspended Solids within specification. [Step 6.e]

• PERFORM daily grab samples for Secondary activity or DECLARE RM-057 Radiation Monitor in service. [Step 6.f]

NOTE This step is performed to ensure that the DVM NI indication is greater than or equal to actual power.

• When power is stable at approximately 10% (as indicated by highest of NI and T power), ADJUST RPS power per OI-NI-1. [Step 6.g]

• OPEN MFW Isolation Valves HCV-1103 & HCV-1104. [Step 6.h]

Examiner Note: The following steps are from OI-RR-1, Reactor Regulating System Normal Operation, Attachment 7, Manual Sequential Mode Checklist, and is maintained as a Control Room hard copy.

ENSURE an out-of-scan CEA is NOT selected as Target Rod on CB-4.

ATCO

[Step 1]

VERIFY alarm REGULATING GROUP WITHDRAWAL PROHIBIT is clear.

ATCO

[Step 2]

NRC Simulator Scenario 4 Outline Rev. Final As Run

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 1 Page 8 of 37 Event

Description:

Raise Reactor Power Time Position Applicants Actions or Behavior PLACE Rod Control Mode Selector Switch in Manual Sequential (MS).

ATCO

[Step 3]

NOTE Continuous CEA motion shall be avoided whenever possible. CEA motion should be stopped at least every 33 inches (43 seconds of continuous CEA motion) to check position of CEAs in Group and Reactor response.

MOVE Manual Rod Control Switch to RAISE or LOWER as required.

ATCO

[Step 4]

DETERMINE appropriate Group Overlap during WITHDRAWAL is N/A.

ATCO

[Step 5]

When CEAs are at desired position, RELEASE Manual Rod Control Switch.

ATCO

[Step 6]

ATCO VERIFY all CEA motion has stopped. [Step 7]

ATCO If additional movement is required, GO TO Step 3. [Step 10]

When completed, PLACE Rod Control Mode Selector Switch in OFF.

ATCO

[Step 11]

Examiner Note: The following steps are from OI-MS-1A, Main Steam System Operation, Attachment 5, Steam Dump and Bypass Manual Control Function.

If operating all Steam Dump and Bypass Valves via the Pressure Controller BOPO (PC0910), PERFORM the following (SEC/MS/SD&B Control): [Step 1]

• DETERMINE PC0910, STM DMP & BYP PRESS CONTROL, in MANUAL control. [Step 1.a]

• As required, ADJUST output to Steam Dump and Bypass Valves.

[Step 1.b]

NRC Simulator Scenario 4 Outline Rev. Final As Run

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 1 Page 9 of 37 Event

Description:

Raise Reactor Power Time Position Applicants Actions or Behavior Booth Operator: When power has been raised approximately 3%, and prior to transitioning to the next event, CONTACT the Control Room as the Shift Manager and direct placing Steam Dump and Turbine Bypass System (pressure and temperature control) in AUTO.

• If desired to transfer back to AUTO at Output that has been selected, BOPO COMPLETE the following on Digital Control System: [Step 1.c]

• PLACE PC0910 in LOCAL. [Step 1.c.1)]

• ADJUST PC0910 SPT to approximately match PC0910 MEAS value.

[Step 1.c.2)]

• PLACE PC0910 back in AUTO. [Step 1.c.3)]

If operating all Steam Dump and Bypass Valves via the Temperature BOPO Controller (TC0909_PI), PERFORM the following (SEC/MS/SD&B Control):

[Step 2]

• DETERMINE TC0909_PI, STM DMP & BYP TEMP CONTROL, in MANUAL control. [Step 2.a]

• As required, ADJUST output to Steam Dump and Bypass Valves.

[Step 2.b]

• If desired to transfer back to AUTO at Output that has been selected,

+20 min BOPO COMPLETE the following on Digital Control System: [Step 2.c]

• PLACE TC0909_PI in LOCAL. [Step 2.c.1)]

• ADJUST TC0909_PI SPT to approximately match TC0909_PI MEAS value. [Step 2.c.2)]

• PLACE TC0909_PI back in AUTO. [Step 2.c.3)]

When Reactor power is raised 3% to 5%, PROCEED to Event 2.

NRC Simulator Scenario 4 Outline Rev. Final As Run

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 2 Page 10 of 37 Event

Description:

Raw Water Pump Discharge Line Leak Time Position Applicants Actions or Behavior Booth Operator: When directed, EXECUTE Event 2.

- Raw Water Pump discharge line leak upstream of HCV-2879A.

Indications Available:

CB-1,2,3/A1 - RAW WATER SUPPLY HEADER FLOW LO CB-1,2,3/A1 - RAW WATER SUPPLY HEADER PRESS LO All Raw Water System 10 psig and 25 psig pressure indicating lights OUT

+30 sec ATCO RESPOND to Annunciator Response Procedures.

ATCO INFORM CRS of Raw Water System low pressure and low flow.

Examiner Note: ATCO may Operate to Mitigate per OPD 4-09 and START another Raw Water Pump.

CRS REFER to AOP-18, Loss of Raw Water.

Examiner Note: The following steps are from AOP-18, Loss of Raw Water.

ATCO DETERMINE Raw Water Pump AC-10C is RUNNING. [Step 4.1]

Booth Operator: If not already contacted, 1 minute after Control Room Receipt of alarms, REPORT as Auxiliary Building Operator that he observed water flowing out of Room 18, and he is going in to investigate.

WAIT 30 seconds and REPORT Raw Water System leak in Room 18, upstream of HCV-2879A/B on the header side of the system.

If Raw Water System rupture is indicated, DIRECT Operators to identify ATCO location of leak: [Step 4.2]

• OBSERVE East RW Header Flow FIC-2890 OSCILLATING.

• OBSERVE West RW Header Flow FIC-2891 OSCILLATING.

• OBSERVE RW Pump(s) Current OSCILLATING.

• OBSERVE RW System Pressure PIC-2892 OSCILLATING.

• OBSERVE RW Pump Room Water Level LIC-2889/LC-2825 Level NORMAL.

NRC Simulator Scenario 4 Outline Rev. Final As Run

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 2 Page 11 of 37 Event

Description:

Raw Water Pump Discharge Line Leak Time Position Applicants Actions or Behavior ATCO DETERMINE Raw Water vault flooding is NOT occurring. [Step 4.3]

DETERMINE Raw Water leak in Auxiliary Building and PERFORM the ATCO following: [Step 4.4]

• ENSURE only one Raw Water Pump RUNNING. [Step 4.4.a]

• IMPLEMENT Attachment C, Equipment Isolation. [Step 4.4.b]

Examiner Note: If the leak has NOT been isolated and another Raw Water Pump NOT started, the crew will either continue in AOP-18 (CCW temperature > 110°F) OR transition to AOP-35, RCP Malfunctions (RCP motor bearing temps > 203°F).

ATCO DETERMINE CCW temperature 110°F. [Step 4.5]

CRS * [CA] If CCW temperature > 110°F, GO TO Step 10. [Step 4.5.1]

CRS IMPLEMENT the Emergency Plan. [Step 4.6]

Examiner Note: The following steps are from AOP-18, Attachment C, Equipment Isolation.

CRS If leak is on Raw Water System, GO TO Step 8. [Step 1]

NOTE The leak isolation Steps 8 through 15 may be performed in any logical order.

ATCO DETERMINE leak is NOT on any of the following: [Step 8]

• WEST RW header

• AC-12A, Raw Water Strainer

• AC-1C, RW Heat Exchanger DETERMINE leak is on East Raw Water Header and PERFORM the ATCO following to ISOLATE Header: [Step 9]

• PLACE AC-10D, Raw Water Pump, in PULL-TO-LOCK. [Step 9.a]

NRC Simulator Scenario 4 Outline Rev. Final As Run

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 2 Page 12 of 37 Event

Description:

Raw Water Pump Discharge Line Leak Time Position Applicants Actions or Behavior ATCO

• CLOSE all Raw Water Header Isolation Valves: [Step 9.b]

• CLOSE HCV-2876A.

• CLOSE HCV-2876B.

• CLOSE HCV-2894.

• CLOSE HCV-2879A.

• CLOSE HCV-2879B.

• CLOSE HCV-2883A.

• CLOSE HCV-2883B.

Booth Operator: When contacted, REPORT RW-145 is CLOSED.

When contacted, EXECUTE local actions and REPORT handjacks applied to Raw Water System Valves as directed.

• Locally CLOSE RW-145, RAW WATER STRAINER AC-12B ATCO BACKWASH VALVE HCV-2805B OUTLET ISOLATION VALVE in RW Vault. [Step 9.c]

• DETERMINE leak is isolated and one Raw Water Pump RUNNING.

CRS

[Step 9.d]

Examiner Note: The following steps continue from AOP-18.

CRS DETERMINE Raw Water System restored to service. [Step 4.8]

CRS EVALUATE Technical Specification LCO 2.4, Containment Cooling

• LCO 2.4.(1).a.iv - Raw Water Header

• CONDITION 2.4.(2).d - Raw Water Header inoperable.

• ACTION 2.4.(2).d - RESTORE Raw Water Header within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> OR PLACE Reactor in HOT SHUTDOWN condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Examiner Note: If the leak has been isolated and Raw Water is restored, CONTINUE to the next event.

Examiner Note: If the leak has NOT been isolated and another Raw Water Pump NOT started, the crew will either continue in AOP-18 (CCW temperature > 110°F) OR transition to AOP-35, RCP Malfunctions (RCP motor bearing temps > 203°F).

NRC Simulator Scenario 4 Outline Rev. Final As Run

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 2 Page 13 of 37 Event

Description:

Raw Water Pump Discharge Line Leak Time Position Applicants Actions or Behavior Examiner Note: The following steps continue from AOP-18.

CRS DETERMINE Raw Water System NOT restored to service. [Step 4.9]

CRS If the Reactor is critical, PERFORM the following: [Step 4.10]

• TRIP the Reactor. [Step 4.10.a]

• IMPLEMENT EOP-00, Standard Post Trip Actions. [Step 4.10.b]

Upon Loss of Cooling to Reactor Coolant Pumps (RCPs), Trip the Reactor and Trip Associated RCP(s) within 5 minutes of Total Loss of CCW flow or Motor CRITICAL TASK Bearing Temperatures > 203°F but prior to exceeding 210°F or within 5 minutes STATEMENT of CCW temperature > 110°F.

Time CCW > 110°F: _____ minutes.

CRITICAL TASK ATCO

• Manually TRIP Reactor at CB-4.

Examiner Note: The following steps are from AOP-35, RCP Malfunctions,Section II, Motor Bearing System Failures.

CRS VERIFY none of the following conditions exist: [Step 4.1]

• Motor guide or thrust bearing temperatures > 203°F for RC-3A/3C/3D.

• Motor guide or thrust bearing temperatures > 230°F for RC-3B.

• [CA] If any bearing temperature exceeds its limit and the Reactor ATCO is critical, PERFORM the following: [Step 4.1.1]

• [CA] TRIP the Reactor. [Step 4.1.1.a]

• [CA] IMPLEMENT EOP-00, Standard Post Trip Actions.

[Step 4.1.1.b]

• [CA] STOP the affected RCPS. [Step 4.1.1.c]

• [CA] GO TO Section 5.0, Exit Conditions. [Step 4.1.1.d]

NRC Simulator Scenario 4 Outline Rev. Final As Run

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 2 Page 14 of 37 Event

Description:

Raw Water Pump Discharge Line Leak Time Position Applicants Actions or Behavior Upon Loss of Cooling to Reactor Coolant Pumps (RCPs), Trip Associated RCP(s) within 5 minutes of Total Loss of CCW flow or Motor Bearing CRITICAL TASK Temperatures > 203°F but prior to exceeding 210°F or within 5 minutes of CCW STATEMENT temperature > 110°F.

Time RCPs exceeding > 203°F: _____ minutes CRITICAL TASK ATCO

• Manually TRIP Reactor at CB-4.

CRITICAL TASK ATCO

• Manually TRIP any affected Reactor Coolant Pumps.

When the Reactor and RCPs have been tripped, PROCEED to Events 6 & 7.

NRC Simulator Scenario 4 Outline Rev. Final As Run

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 3 Page 15 of 37 Event

Description:

Inadvertent Auxiliary Feedwater Actuation Signal Time Position Applicants Actions or Behavior Booth Operator: When directed, EXECUTE Event 3.

- Inadvertent Auxiliary Feedwater Actuation Signal.

Indications Available:

AI-66B/A66B - AFWS STEAM GEN RC-2A CHANNEL B ACTUATED AI-66B/A66B - FW-10 TURBINE DRIVEN FEEDWATER PUMP RUNNING (~30 seconds later)

AI-66B/A66B - FW-10 TURBINE OIL PUMP RUNNING (~30 seconds later)

+30 sec BOPO RESPOND to Annunciator Response Procedures.

BOPO INFORM CRS of Auxiliary Feedwater Actuation Signal initiation.

Examiner Note: BOPO may Operate to Mitigate per OPD 4-09 and CLOSE HCV-1107A and HCV-1107B to stop FW-10, Turbine Driven Auxiliary Feedwater Pump.

REFER to ARP-AI-66B/A66B, Window 41, AFWS STEAM GEN RC-2A CRS CHANNEL B ACTUATED.

Examiner Note: The following steps are from ARP-AI-66B/A66B, Window 41 - AFWS STEAM GEN RC-2A CHANNEL B ACTUATED.

CHECK A/B/LI-911, Steam Generator RC-2A Level at AI-66A and AI-66B.

BOPO

[Step 1]

• DETERMINE SG level LI-911A at Panel AI-66A NORMAL.

• DETERMINE SG level LI-911B at Panel AI-66B NORMAL.

Booth Operator: When contacted, REPORT LI-911D, RC-2A level at AI-179 is ~ 64% and LI-911C, RC-2A pressure is ~ 884 psia (or as indicated).

BOPO DISPATCH Operator to check C/D/LI-911, RC-2A Level at AI-179. [Step 2]

BOPO DETERMINE Steam Generator Wide Range level is > 32%. [Step 3]

DETERMINE AFAS initiation is inadvertent and IMPLEMENTS AOP-23, CRS Reset of Engineered Safeguards,Section IX, Reset of Inadvertent AFAS.

[Step 4]

NRC Simulator Scenario 4 Outline Rev. Final As Run

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 3 Page 16 of 37 Event

Description:

Inadvertent Auxiliary Feedwater Actuation Signal Time Position Applicants Actions or Behavior CRS REFER to Technical Specification LCOs 2.14 and 2.15. [Step 5]

EVALUATE Technical Specification LCO 2.15.1, Instrumentation and Control CRS Systems

• LCO 2.15.1.(3) - Automatic Initiation Steam Generator Water Level Logic Subsystem B (Table 2.3 / Function 5).

• CONDITION 2.15.1.(3) - Logic Subsystem B inoperable

• ACTION 2.15.1.(3) - RESTORE inoperable channel within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> OR PLACE Reactor in HOT SHUTDOWN condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Examiner Note: The following steps are from AOP-23, Reset of Engineered Safeguards,Section IX, Reset of Inadvertent AFAS.

CRS DETERMINE the AFAS is inadvertent. [Step 4.1]

CRS REFER to the following Technical Specifications: [Step 4.2]

• LCO 2.5, Steam and Feedwater Systems

• LCO 2.15, Instrumentation and Control Systems Examiner Note: Entry into Technical Specification LCO 2.5.(1).d is required until FW-10, TDAFW Pump is reset and returned to AUTO at the end of this event.

EVALUATE Technical Specification LCO 2.5, Steam and Feedwater CRS Systems

• LCO 2.5.(1) - Two AFW Trains OPERABLE

• CONDITION 2.5.(1).d - Both AFW Trains inoperable

• ACTION 2.5.(1).d - RESTORE one train to OPERABLE status immediately.

BOPO ENSURE both of the following valves in AUTO: [Step 4.3]

• DETERMINE FCV-1368, FW-6 Recirc Valve in AUTO.

• DETERMINE FCV-1369, FW-10 Recirc Valve in AUTO.

NRC Simulator Scenario 4 Outline Rev. Final As Run

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 3 Page 17 of 37 Event

Description:

Inadvertent Auxiliary Feedwater Actuation Signal Time Position Applicants Actions or Behavior PLACE control switches for the following AFW Isolation Valves in CLOSE:

BOPO

[Step 4.4]

• PLACE HCV-1107A in CLOSE.

• PLACE HCV-1107B in CLOSE.

• PLACE HCV-1108A in CLOSE.

• PLACE HCV-1108B in CLOSE.

BYPASS affected logic subsystem per OI-AFW-2, Auxiliary Feedwater CRS System Actuation and Bypass. [Step 4.5]

Examiner Note: The following steps are from OI-AFW-2, Auxiliary Feedwater System Actuation and Bypass, Attachment 1, Bypass of the Auxiliary Feedwater Actuation Signal (AFAS) (Modes 1 or 2).

BOPO DETERMINE AFAS is aligned for automatic initiation. [Step 2]

BOPO DETERMINE plant is in Mode 1. [Step 3]

DETERMINE if an Instrument Channel or a Logic Subsystem Channel is to CRS be bypassed. [Step 1]

• DETERMINE an Instrument Channel will NOT be bypassed. [Step 1.a]

• DETERMINE a Logic Subsystem Channel will be bypassed and GO TO Step 3. [Step 1.b]

If a Logic Subsystem Channel of AFAS is to be bypassed, COMPLETE the CRS following: [Step 3]

SM/CRS

• LOG entry into Technical Specification 2.15.1(3), 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> LCO.

NRC Simulator Scenario 4 Outline Rev. Final As Run

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 3 Page 18 of 37 Event

Description:

Inadvertent Auxiliary Feedwater Actuation Signal Time Position Applicants Actions or Behavior NOTE The following alarms are expected depending on the Logic Subsystem Channel that is bypassed:

• AFWS RC-2A CH A MATRIX TS-A/RC-2A/AFWS TEST SWITCH OFF NORM (AI-66A, Window 24)

• AFWS RC-2B CH A MATRIX TS-A/RC-2B/AFWS TEST SWITCH OFF NORM (AI-66A, Window 25)

• AFWS OVERRIDE SWITCH A/OR-RC-2A/AFWS OFF NORMAL (AI-66A, Window 29)

• AFWS OVERRIDE SWITCH A/OR-RC-2B/AFWS OFF NORMAL (AI-66A, Window 30)

• HCV-1107A & B AFWS OVERRIDE SWITCH CH A OR B OFF NORM (AI-66A, Window 35)

• AFWS RC-2A CH B MATRIX TS-B/RC-2A AFWS TEST SWITCH OFF NORM (AI-66B, Window 21)

• AFWS RC-2B CH B MATRIX TS-B/RC-2B/AFWS TEST SWITCH OFF NORM (AI-66B, Window 22)

• AFWS OVERRIDE SWITCH B/OR-RC-2A/AFWS OFF NORMAL (AI-66B, Window 26)

• AFWS OVERRIDE SWITCH B/OR-RC-2B/AFWS OFF NORMAL (AI-66B, Window 27)

• HCV-1108A & B AFWS OVERRIDE SWITCH CHA OR B OFF NORMAL (AI-66A, Window 32)

BYPASS selected Logic Subsystem using Table 2, AFAS Logic Subsystem BOPO Bypass Switch Alignment, and RECORD as left information in appropriate slots. [Step 3.b]

Examiner Note: The following steps are from OI-AFW-2, Table 2, AFAS Logic Subsystem Channel Bypass Switch Alignment.

Table 2 - AFAS Logic Subsystem Channel Bypass Switch Alignment As-Left Switch Bypassing Channel Panel No. Switch Position Position RC-2A Channel B AI-66B S/G RC-2A Chan. B Auto Sig Bypass (Amber lamps S/G RC- Override Relay Test Sw 2A Chan B/B1)

S/G RC-2A Chan. B Auto Sig Override Override Sw AFW Pumps FW-6/FW-10 Chan. B AFW Auto Sig B/OR -1107 Override S/G Feed Valves AFWS Examiner Note: Acting as Shift Manager, PROVIDE Keys #55 and #57 when requested.

BOPO PERFORM the following at Panel AI-66B for RC-2A Channel B:

• INSERT key #57 and PLACE S/G RC-2A Channel B Auto Signal Override Relay Test Switch in BYPASS.

NRC Simulator Scenario 4 Outline Rev. Final As Run

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 3 Page 19 of 37 Event

Description:

Inadvertent Auxiliary Feedwater Actuation Signal Time Position Applicants Actions or Behavior

• INSERT key #55 and PLACE S/G RC-2A Channel B Auto Signal Override Switch AFW Pumps FW-6/FW-10 in OVERRIDE.

• PLACE Channel B AFW Auto Signal Override S/G Feed Valves to B/OR

-1107 AFWS position.

Examiner Note: The following steps continue from AOP-23,Section IX, Reset of Inadvertent AFAS.

BOPO PERFORM the following to STOP all AFW Pumps: [Step 4.6]

• CLOSE YCV-1045, FW-10 Steam Inlet Valve. [Step 4.6.a]

• PLACE both Override Switches in OVERRIDE: [Step 4.6.b]

• ISOLATION VALVE YCV-1045A OVERRIDE SW.

• ISOLATION VALVE YCV-1045B OVERRIDE SW.

• CLOSE both FW-10 Steam Supply Valves: [Step 4.6.c]

• YCV-1045A, RC-2A to FW-10 Isolation Valve.

• YCV-1045B, RC-2B to FW-10 Isolation Valve.

• ENSURE FIC-1369, AUX FW PUMP FW-10 SUCTION FLOW drops to zero. [Step 4.6.d]

• STOP FW-6, Electric AFW Pump, and PLACE HC-1367, FW-6 Control Switch, in PULL-TO-LOCK. [Step 4.6.e]

• ENSURE FIC-1368, AUX FW PUMP FW-6 SUCTION FLOW drops to zero. [Step 4.6.f]

PERFORM the following to return the AFW System to automatic operation:

BOPO

[Step 4.7]

• PLACE Control Switches for AFW Isolation Valves in RESET:

[Step 4.7.a]

• PLACE HCV-1107A in RESET.

• PLACE HCV-1107B in RESET.

• PLACE HCV-1108A in RESET.

• PLACE HCV-1108B in RESET.

• PLACE Control Switches for AFW Isolation Valves in AUTO: [Step 4.7.b]

NRC Simulator Scenario 4 Outline Rev. Final As Run

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 3 Page 20 of 37 Event

Description:

Inadvertent Auxiliary Feedwater Actuation Signal Time Position Applicants Actions or Behavior

• PLACE HCV-1107A in AUTO.

• PLACE HCV-1107B in AUTO.

• PLACE HCV-1108A in AUTO.

• PLACE HCV-1108B in AUTO.

• PLACE Control Switch for YCV-1045, FW-10 Steam Inlet Valve, in RESET. [Step 4.7.c]

• PLACE Control Switch for YCV-1045, FW-10 Steam Inlet Valve, in AUTO. [Step 4.7.d]

• PLACE both Override Switches in NORMAL. [Step 4.7.e]

• ISOLATION VALVE YCV-1045A OVERRIDE SW

• ISOLATION VALVE YCV-1045B OVERRIDE SW

• PLACE HC-1367, FW-6 Control Switch, in AFTER-STOP. [Step 4.7.f]

Booth Operator: When contacted, EXECUTE remote functions to RESET FW-10 and Trip Latch Clamp is finger tight.

CONTACT Auxiliary Operator ENSURE FW-64-RL, AUX FEED PUMP BOPO FW-10 MANUAL TRIP LATCH RESET LEVER is latched: [Step 4.8]

• VERIFY Reset Lever is seated.

• ENSURE FW-64-C, AUX FEED PUMP FW-10 MANUAL TRIP LATCH CLAMP is installed finger tight.

CRS EXIT Technical Specification LCO 2.5, Steam and Feedwater. [Step 4.9]

When AFAS has been RESET, PROCEED to Event 4.

NRC Simulator Scenario 4 Outline Rev. Final As Run

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 4 Page 21 of 37 Event

Description:

Loss of Instrument Bus Time Position Applicants Actions or Behavior Booth Operator: When directed, EXECUTE Event 4. (Alternate Path Event 8)

- Loss of Instrument Bus AI-40A.

Indications Available:

CB-20/A15 - INVERTER A TROUBLE CB-20/A15 - INSTRUMENT BUS A LOW VOLTAGE/GROUND (~10 seconds later)

Multiple Loss of Instrument Bus alarms

+30 sec BOPO RESPOND to Annunciator Response Procedures.

CREW INFORM CRS of Loss of Instrument Bus AI-40A.

Booth Operator: When contacted, REPORT Inverter A Output Breaker is TRIPPED.

REFER to AOP-16, Loss of Instrument Bus Power,Section I, Loss of CRS Instrument Bus Power.

Examiner Note: The following steps are from AOP-16, Loss of Instrument Bus Power,Section I, Loss of Instrument Power.

CRS DETERMINE a Reactor Trip has NOT occurred: [Step 4.1]

CRS DETERMINE appropriate AOP-16 Section: [Step 4.2]

• OBSERVE an INVERTER A TROUBLE alarm.

• OBSERVE an INSTRUMENT BUS A LOW VOLTAGE/GROUND alarm.

CRS GO TO AOP-16,Section II, Loss of Instrument Bus AI-40A.

Examiner Note: The following steps are from AOP-16, Loss of Instrument Bus Power,Section II, Loss of Instrument Bus AI-40A.

CRS VERIFY Loss of Instrument Bus AI-40A by the following: [Step 4.1]

• INVERTER A TROUBLE alarm.

• INSTRUMENT BUS A LOW VOLTAGE/GROUND alarm.

NRC Simulator Scenario 4 Outline Rev. Final As Run

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 4 Page 22 of 37 Event

Description:

Loss of Instrument Bus Time Position Applicants Actions or Behavior NOTE

1. Upon loss of Instrument Bus A, ALL of the following instrumentation or equipment associated with the Reactivity Control Safety Function is affected as follows:

• All RPS Channel A is in trip

• Channel A "VARIABLE OVER POWER TRIP POWER MARGIN A/JI-007" meter is inoperable

• Channel A Wide Range Log Power Meter and Rate Meter are inoperable

• The Diverse Scram System is in half-trip

• Three RPS Logic Matrix channels are inoperable

2. Loss of more than one RPS Logic Matrix channel requires entry into T.S. 2.15.2.

3. If the associated clutch power supply is selected to Instrument Bus A then two RPS Trip Initiation Logic channels (AB, AC, AD) are inoperable.

DETERMINE clutch power supply selected to AI-40A and VERIFY clutch ATCO power supply is DEENERGIZED: [Step 4.2]

• OBSERVE AI-3-PS1 output current is 0.

• OBSERVE AI-3-PS3 output current is 0.

• OBSERVE AI-3-PS1 Indicating lights are out.

• OBSERVE AI-3-PS3 Indicating lights are out.

• OBSERVE clutch power supply breaker in half trip position.

Examiner Note: Acting as Shift Manager, PROVIDE Trip Unit Keys #1 to #12 when requested.

ATCO INSERT keys and BYPASS all RPS Channel A Bistable Trip Units. [Step 4.3]

CRS COMPLY with Technical Specification 2.15.2(5). [Step 4.4]

EVALUATE Technical Specification LCO 2.15, Instrumentation and Control CRS Systems

• LCO 2.15.2 - Reactor Protective System Logic and Trip Initiation

• CONDITION 2.15.2.(2) - One RPS Trip Initiation Logic channel inoperable.

• ACTION 2.15.2.(2) - Deenergize the affected clutch power supply within one hour (in 1/2 trip).

• ACTION 2.15.2.(5) - With the required actions of (2) not met, be in HOT SHUTDOWN and verify no more than one CEA is capable of being withdrawn within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

NRC Simulator Scenario 4 Outline Rev. Final As Run

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 4 Page 23 of 37 Event

Description:

Loss of Instrument Bus Time Position Applicants Actions or Behavior NOTE Upon loss of Instrument Bus A, ALL of the following instrumentation or equipment associated with the Vital Auxiliaries Safety Function are inoperable:

• "WEST RW SUPPLY HEADER FLOW FIC-2891" indicator

• "CC HT EXCH AC-1A RW OUTLET TEMP TIC-2885"

• "CNTMT CLG COIL VA-1A OUTLT ISOL VLV CNTRLR HCV-400C"

• "CNTMT CLG COIL VA-1B OUTLT ISOL VLV CNTRLR HCV-401C"

• "CNTMT CLG COIL VA-8A OUTLT ISOL VLV CNTRLR HCV-402C"

• "CNTMT CLG COIL VA-8B OUTLT ISOL VLV CNTRLR HCV-403C" ATCO ENSURE CCW System operation satisfactory: [Step 4.5]

• DETERMINE one CCW Pump RUNNING.

• DETERMINE CCW pressure 60 psig.

ATCO DETERMINE one Raw Water Pump RUNNING. [Step 4.6]

BOPO DETERMINE Instrument Air pressure 90 psig. [Step 4.7]

NOTE Upon loss of Instrument Bus A, ALL of the following instrumentation or equipment associated with the RCS Inventory Control Safety Function is affected as follows:

• Letdown is isolated

• Pressurizer Level Channel X is inoperable

• CH-1A, Charging Pump, is inoperable

• Charging Pump Backup Auto starts are disabled MAINTAIN Pressurizer level between 30% and 70% and TRENDING to ATCO between 45% percent by operating Charging Pumps CH-1B and/or CH-1C per IC-11, Inventory Control. [Step 4.8]

ATCO CLOSE TCV-202, Letdown Isolation Valve. [Step 4.9]

Examiner Note: At Chief Examiner discretion, PROCEED to the next event.

NRC Simulator Scenario 4 Outline Rev. Final As Run

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 4 Page 24 of 37 Event

Description:

Loss of Instrument Bus Time Position Applicants Actions or Behavior PLACE HC-101, Pressurizer Level Channel Selector Switch, in CHAN Y ATCO position. [Step 4.10]

NOTE Upon loss of Instrument Bus A, ALL of the following instrumentation or equipment associated with the RCS Pressure Control Safety Function is affected as follows:

• RCS Pressure Control Channel X is inoperable

• "PRESSURIZER PRESSURE A/PIA-102X AND A/PIA-102Y" indicators are inoperable

• PZR Backup Heaters are on

• PZR Heater Cutout is inoperable PLACE HC-103, Pressurizer Pressure Channel Selector Switch in CHAN Y ATCO position. [Step 4.11]

Manually CONTROL Pressurizer Heaters per PC-11, Pressure Control.

ATCO

[Step 4.12]

MAINTAIN RCS pressure per PC-12, RCS Pressure-Temperature Limits.

ATCO

[Step 4.13]

NOTE

1. Only one additional channel trip is needed to actuate the PORVs, even if the channel in trip is bypassed.

2. When RCS Heatup or Cooldown is in progress, the PORVs are the primary means of Low Temperature Overpressure Protection.

3. Closing the PORV block valves requires entry into Tech Spec 2.1.6.

CRS CONSIDER closing PORV Block Valves HCV-150 and HCV-151. [Step 4.14]

NRC Simulator Scenario 4 Outline Rev. Final As Run

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 4 Page 25 of 37 Event

Description:

Loss of Instrument Bus Time Position Applicants Actions or Behavior NOTE Upon loss of Instrument Bus A, ALL of the following instrumentation or equipment associated with the Core Heat Removal Safety Function are inoperable:

• "SUBCOOLED MARGIN MONITOR A-168"

• "RC LOOP TEMPERATURES LOOP 1A "T-COLD" A/TI-112C"

• "RC LOOP TEMPERATURES LOOP 1 "T-HOT" A/TI-112H"

• "RC LOOP TEMPERATURES LOOP 2A "T-COLD" A/TI-122C"

• "RC LOOP TEMPERATURES LOOP 2 "T-HOT" A/TI-122H"

• "SHTDN HT EXCH AC-4A OUTLET VALVE CNTRLR HCV-484" ATCO DETERMINE all RCPs are RUNNING. [Step 4.15]

NOTE Upon loss of Instrument Bus A, ALL of the following instrumentation or equipment associated with the RCS Heat Removal Safety Function is inoperable:

• "EMGY FW STOR TNK LEVEL LIA-1183"

• "STEAM GENERATOR PRESSURE S/G RC-2A A/PIC-902"

• "STEAM GENERATOR PRESSURE S/G RC-2B A/PIC-905"

• "AUX FW PUMP FW-6 SUCTION FLOW FIC-1368" BOPO DETERMIN Steam Generator NR levels steady at ~63%. [Step 4.16]

NOTE

• Upon loss of Instrument Bus A, RM-091A, which is associated with the Containment Integrity Safety Function is inoperable.

ATCO PERFORM the following to CONFIRM Containment Integrity: [Step 4.17]

• DETERMINE no unexpected rise in Containment Sump level.

[Step 4.17.a]

• DETERMINE no Containment Area Radiation Monitor alarms.

[Step 4.17.b]

• DETERMINE Radiation Monitors RM-051 / RM-052 / RM-062 NOT in alarm. [Step 4.17.c]

• DETERMINE SG Blowdown or Condenser off Gas Radiation Monitors RM-054A / RM-054B / RM-057 NOT in alarm. [Step 4.17.d]

• DETERMINE Containment conditions NORMAL. [Step 4.17.e]

• DETERMINE Containment pressure < 3 psig.

NRC Simulator Scenario 4 Outline Rev. Final As Run

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 4 Page 26 of 37 Event

Description:

Loss of Instrument Bus Time Position Applicants Actions or Behavior

• DETERMINE Containment temperature <120°F.

ATCO PLACE the following switches in TEST: [Step 4.18]

• HC-344/TEST, CNTMT SPRAY VLV HCV-344 TEST SWITCH

• HC-345/TEST, CNTMT SPRAY VLV HCV-345 TEST SWITCH NOTE Upon loss of Instrument Bus A, ALL of the following instrumentation or equipment associated with the Engineered Safety Features Systems is affected as follows:

• Safety Injection Tanks 6A and 6C level and pressure indicators are inoperable

• OPLS is in half-trip

• Sequencer S2-2 is inoperable

• PPLS is in a two-out-of-three logic mode

• SGLS is in a two-out-of-three logic mode CRS REFER to all the following Technical Specifications: [Step 4.19]

• 2.1.6, Pressurizer and Steam System Safety Valves

• 2.2, Chemical and Volume Control System

• 2.3, Emergency Core Cooling System

• 2.5, Steam and Feedwater Systems

• 2.7, Electrical Systems

• 2.15, Instrumentation and Control Systems

• 2.21, Post-Accident Monitoring Instrumentation CRS EVALUATE Technical Specification LCO 2.7, Electrical Systems

• LCO 2.7.(1).h - 120 VAC Instrument Bus A (Panel AI-40A).

• CONDITION 2.7.(2).h - 120 VAC Instrument Bus A (Panel AI-40A) inoperable

• ACTION 2.7.(2).h - May remain inoperable for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> provided RPS and ESF instrument channels supplied by the remaining 3 buses are all OPERABLE.

REFER to Electrical Load Distribution Listing Manual for a list of components CREW powered from AI-40A. [Step 4.20]

NRC Simulator Scenario 4 Outline Rev. Final As Run

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 4 Page 27 of 37 Event

Description:

Loss of Instrument Bus Time Position Applicants Actions or Behavior Examiner Note: Instrument Bus IA-40A will remain deenergized for duration of scenario.

Booth Operator: When contacted, REPORT Electrical Maintenance investigating issue with Inverter A.

When cause of power loss has been determined and corrected, RESTORE

+15 min CRS AI-40A to normal per Attachment 1 or 12 of OI-EE-4, 120 Volt AC System Normal Operation. [Step 4.21]

When Technical Specifications have been addressed, PROCEED to Events 5, 6, and 7.

(Alternate Path: If this event was initiated as Alternate Path Event 8, Terminate the Scenario).

NRC Simulator Scenario 4 Outline Rev. Final As Run

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 5, 6, & 7 Page 28 of 37 Event

Description:

RCP Trip / Automatic Reactor Trip Failure / Instrument Air Compressors Trip / Bearing Cooling Water Pump Trip / Steam Line Break inside Containment Time Position Applicants Actions or Behavior Booth Operator: When directed, EXECUTE Events 5, 6, and 7.

- Reactor Coolant Pump RC-3A trip.

- Instrument Air Compressors CA-1B and CA-1C trip.

- Bearing Cooling Water Pump AC-9B trip.

- Steam Line Break inside Containment on RC-2A @ 0.65% severity and 5 minute ramp.

Indications Available:

CB-1,2,3,4/A6 - REACTOR COOLANT PUMP RC-3A BREAKER O/L OR TRIP Low Flow Trip Unit lights lit on all RPS Channels B/C/D ERF Computer System alarms for low RCS flow

+30 sec ATCO RECOGNIZE RPS Low Flow lights lit and MANUALLY trip Reactor.

CRS DIRECT performance of EOP-00, Standard Post Trip Actions.

Examiner Note: The following steps are from EOP-00, Standard Post Trip Actions.

ATCO VERIFY Reactivity Control: [Step 1]

• VERIFY ALL of the following: [Step 1.a]

• DETERMINE more than one Regulating or Shutdown CEA NOT inserted.

• [CA] If Reactor did NOT trip, ESTABLISH Reactivity Control by performing step a, b, c or d: [Step 1.1]

Manually Trip Reactor to meet Core Design Criteria of Lowering Reactor Power CRITICAL TASK and Negative Startup Rate to Verify Reactivity Control Established During STATEMENT ATWS Event Prior to Exiting EOP-00, Standard Post Trip Actions.

CRITICAL TASK ATCO * [CA] Manually TRIP Reactor at CB-4. [Step 1.1.a]

• VERIFY Reactor Power is LOWERING.

• VERIFY Startup Rate is NEGATIVE.

NRC Simulator Scenario 4 Outline Rev. Final As Run

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 5, 6, & 7 Page 29 of 37 Event

Description:

RCP Trip / Automatic Reactor Trip Failure / Instrument Air Compressors Trip / Bearing Cooling Water Pump Trip / Steam Line Break inside Containment Time Position Applicants Actions or Behavior Examiner Note: An Emergency Boration is performed once the cooldown is recognized.

• DETERMINE an uncontrolled RCS Cooldown in progress. [Step 1.b]

• [CA] PERFORM Emergency Boration with uncontrolled cooldown in progress. [Step 1.2]

• [CA] ENSURE both following valves CLOSED: [Step 1.2.a]

• [CA] FCV-269X, Demin Water Makeup Valve

• [CA] FCV-269Y, Boric Acid Makeup Valve

• [CA] OPEN all the following valves: [Step 1.2.b]

• [CA] HCV-268, Boric Acid Pump Header to Charging Pumps Isolation Valve

• [CA] HCV-265, CH-11A Gravity Feed Valve

• [CA] HCV-258, CH-11B Gravity Feed Valve

• [CA] START all the following pumps: [Step 1.2.c]

• [CA] Boric Acid Pump CH-4A

• [CA] Boric Acid Pump CH-4B

• [CA] Charging Pump CH-1A (inoperable)

• [CA] Charging Pump CH-1B

• [CA] Charging Pump CH-1C

• [CA] CLOSE LCV-218-2, VCT Outlet Valve. [Step 1.2.d]

• [CA] ENSURE the following valves CLOSED: [Step 1.2.e]

• [CA] LCV-218-3, Charging Pump Suction SIRWT Isolation Valve

• [CA] HCV-257, CH-4B Recirc Valve

• [CA] HCV-264, CH-4A Recirc Valve

• [CA] BORATE until adequate Shutdown Margin is established.

[Step 1.2.f]

CRS DETERMINE Reactivity Control criteria SATISFIED.

BOPO VERIFY Turbine Trip: [Step 2]

• VERIFY HP & LP Stop and Intercept Valves CLOSED.

NRC Simulator Scenario 4 Outline Rev. Final As Run

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 5, 6, & 7 Page 30 of 37 Event

Description:

RCP Trip / Automatic Reactor Trip Failure / Instrument Air Compressors Trip / Bearing Cooling Water Pump Trip / Steam Line Break inside Containment Time Position Applicants Actions or Behavior Examiner Note: The Generator Output Breakers are CLOSED due to back feeding.

BOPO ENSURE all Generator Breakers are tripped: [Step 3]

• DETERMINE Generator Output Breaker 3451-4 CLOSED.

• DETERMINE Generator Output Breaker 3451-5 CLOSED.

• DETERMINE Generator Field Breaker 41E/G1F tripped.

BOPO VERIFY 4160 V Safeguards Buses 1A3 and 1A4 are ENERGIZED. [Step 4]

CRS DETERMINE Maintenance of Vital Auxiliaries criteria SATISFIED.

DETERMINE Safety Injection Actuation Signal has NOT occurred and both BOPO Diesel Generators are STOPPED. [Step 5]

VERIFY 4160 V Non-Safeguards Buses 1A1 and 1A2 are ENERGIZED.

BOPO

[Step 6]

BOPO VERIFY 125 VDC Buses 1 and 2 are ENERGIZED. [Step 7]

BOPO VERIFY Instrument Air is AVAILABLE: [Step 8]

• DETERMINE Instrument Air pressure < 90 psig.

• DETERMINE Instrument Air Compressors NOT RUNNING.

• [CA] If Instrument Air pressure is < 90 psig, PERFORM the following to restore Instrument Air: [Step 8.1]

BOPO * [CA] START Bearing Water Pump AC-9A.

BOPO * [CA] START Air Compressor CA-1A.

ATCO VERIFY Component Cooling Water System operation NORMAL: [Step 9]

• DETERMINE at least one CCW pump RUNNING. [Step 9.a]

• DETERMINE CCW Pump discharge pressure 60 psig. [Step 9.b]

NRC Simulator Scenario 4 Outline Rev. Final As Run

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 5, 6, & 7 Page 31 of 37 Event

Description:

RCP Trip / Automatic Reactor Trip Failure / Instrument Air Compressors Trip / Bearing Cooling Water Pump Trip / Steam Line Break inside Containment Time Position Applicants Actions or Behavior

• DETERMINE HCV-438A/B/C/D, CCW to RCP Coolers OPEN. [Step 9.c]

• DETERMINE at least one Raw Water Pump RUNNING. [Step 9.d]

CRS VERIFY RCS Inventory Control criteria satisfied: [Step 10]

• DETERMINE PZR level between 30% and 70% and NOT TRENDING to ATCO between 45% and 60%.

• [CA] RESTORE Inventory Control by manually controlling Charging and Letdown. [Step 10.1.a]

• DETERMINE RCS subcooling > 20°F.

CRS DETERMINE RCS Inventory Control criteria NOT SATISFIED.

CRS VERIFY RCS Pressure Control criteria satisfied: [Step 11]

ATCO

• DETERMINE RCS pressure between 1800 psia and 2300 psia.

• DETERMINE RCS pressure NOT TRENDING between 2050 psia and 2150 psia.

• [CA] MANUALLY CONTROL PZR Heaters and Spray to restore RCS pressure.

• DETERMINE PORVs are CLOSED.

CRS DETERMINE RCS Pressure Control criteria NOT SATISFIED.

CRS VERIFY Core Heat Removal criteria satisfied: [Step 12]

ATCO

• DETERMINE RCP NPSH requirements met.

• DETERMINE at least one RCP operating.

• DETERMINE Core T 10°F.

CRS DETERMINE Core Heat Removal criteria SATISFIED.

CRS VERIFY RCS Heat Removal criteria satisfied:

NRC Simulator Scenario 4 Outline Rev. Final As Run

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 5, 6, & 7 Page 32 of 37 Event

Description:

RCP Trip / Automatic Reactor Trip Failure / Instrument Air Compressors Trip / Bearing Cooling Water Pump Trip / Steam Line Break inside Containment Time Position Applicants Actions or Behavior VERIFY Main Feedwater is restoring SG levels to 35% to 80% NR and 73%

BOPO to 94% WR. [Step 13]

• DETERMINE FCV-1101 & FCV-1102 Feed Regulating Valves CLOSED.

[Step 13.a]

• DETERMINE HCV-1105 & HCV-1106 Feed Regulating Bypass Valves ramped to between 40% & 45% OPEN. [Step 13.b]

BOPO

• PLACE 43/FW switch in OFF. [Step 13.c]

• ENSURE no more than one Main Feedwater Pump RUNNING.

[Step 13.d]

• ENSURE no more than one Condensate Pump RUNNING. [Step 13.e]

• STOP running Heater Drain Pumps FW-5A, FW-5B, and/or FW-5C.

[Step 13.f]

BOPO

• ENSURE SG Blowdown Isolation Valves CLOSED. [Step 13.g]

• HCV-1387A & HCV-1387B

• HCV-1388A & HCV-1388B Booth Operator: When contacted, REPORT Air Assisted Main Steam Safety Valves CLOSED.

VERIFY Steam Dump and Bypass Valves controlling both of the following:

BOPO

[Step 14]

• DETERMINE RCS TCOLD NOT between 525°F and 535°F.

• DETERMINE Steam Generator RC-2A pressure < 700 psia.

• [CA] If TCOLD less than 525°F, PERFORM the following: [Step 14.1]

BOPO * [CA] CLOSE Steam Dump and Bypass Valves. [Step 14.1.a]

• [CA] VERIFY HCV-1040, Atmospheric Dump Valve CLOSED.

[Step 14.1.b]

• [CA] CHECK Air Assisted Main Steam Safety Valves CLOSED.

[Step 14.1.c]

• [CA] If Steam Generator pressure < 700 psia, ISOLATE Main Steam Header. [Step 14.1.d]

• [CA] CLOSE HCV-1041A, MSIV. [Step 14.1.d.1)]

• [CA] CLOSE HCV-1042A, MSIV. [Step 14.1.d.1)]

• [CA] VERIFY HCV-1041A, MSIV Bypass CLOSED.

[Step 14.1.d.2)]

NRC Simulator Scenario 4 Outline Rev. Final As Run

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 5, 6, & 7 Page 33 of 37 Event

Description:

RCP Trip / Automatic Reactor Trip Failure / Instrument Air Compressors Trip / Bearing Cooling Water Pump Trip / Steam Line Break inside Containment Time Position Applicants Actions or Behavior

• [CA] VERIFY CLOSE HCV-1042A, MSIV Bypass CLOSED.

[Step 14.1.d.2)]

• [CA] DETERMINE Steam Generator pressure > 500 psia.

[Step 14.1.e]

CRS DETERMINE RCS Heat Removal criteria NOT SATISFIED.

CRS VERIFY Containment Isolation criteria satisfied:

ATCO VERIFY Normal Containment conditions exist: [Step 15]

• DETERMINE rise in Containment Sump level in progress. [Step 15.a]

• DETERMINE Containment Area Radiation Monitors NOT in alarm.

[Step 15.b]

• DETERMINE Containment Ventilation Radiation Monitors NOT in alarm.

[Step 15.c]

• DETERMINE SG Blowdown and Condenser Off Gas Radiation Monitors ATCO NOT alarming. [Step 15.d]

• DETERMINE SG Blowdown and Condenser Off Gas Radiation Monitors ATCO NOT TRENDING to alarm. [Step 15.e]

ATCO

• VERIFY Containment conditions: [Step 15.f]

• DETERMINE Containment pressure > 3 psig.

• DETERMINE Containment temperature > 120°F.

• [CA] INITIATE Containment Cooling. [Step 15.f.1]

ATCO * [CA] ENSURE CCW flow to Containment Vent Fan coils.

• [CA] PLACE HCV-402B/D to OPEN.

• [CA] PLACE HCV-403B/D to OPEN.

• [CA] PLACE HCV-402A/C to OPEN.

• [CA] PLACE HCV-403A/C to OPEN.

ATCO * [CA] START all Containment Vent Fans.

• [CA] VERIFY Containment Vent Fans VA-3A & VA-3B RUNNING.

• [CA] START Containment Vent Fans VA-7C & VA-7D.

• [CA] DETERMINE Containment pressure < 5 psig. [Step 15.f.2]

NRC Simulator Scenario 4 Outline Rev. Final As Run

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 5, 6, & 7 Page 34 of 37 Event

Description:

RCP Trip / Automatic Reactor Trip Failure / Instrument Air Compressors Trip / Bearing Cooling Water Pump Trip / Steam Line Break inside Containment Time Position Applicants Actions or Behavior CRS DETERMINE Containment Integrity criteria NOT SATISFIED.

CRS DIAGNOSE event in progress: [Step 16]

• DETERMINE Reactivity Control requirements met.

• DETERMINE both DC buses energized.

• DETERMINE at least one Vital 4160 V Bus energized.

• DETERMINE at least one Non-Vital 4160 V Bus energized.

• DETERMINE at least one RCP running.

• DETERMINE adequate Feedwater flow to at least one SG.

• VERIFY Pressurizer pressure > 1800 psia with high subcooled margin, normal SG pressure, and no indications of primary to secondary leakage.

• If not, CONSIDER EOP-05, Uncontrolled Heat Extraction.

NOTE Certain events (i.e., LOCA, SGTR, UHE and Loss of All Feedwater) do not require offsite power in order to adequately, mitigate the effects of the accident.

For this reason, the LOCA, SGTR, UHE or Loss of All Feedwater procedure may be implemented even if a Loss of Offsite Power has also occurred.

• DETERMINE all Safety Function Acceptance Criteria NOT SATISFIED.

• DETERMINE single event in progress and TRANSITION to EOP-05, Uncontrolled Heat Extraction.

Examiner Note: The following steps are from EOP-05, Uncontrolled Heat Extraction.

CRS CONFIRM Standard Post Trip Actions have been performed. [Step 1]

CRS CONFIRM Uncontrolled Heat Extraction Diagnosis: [Step 2]

• VERIFY Safety Function Status Check Acceptance Criteria being satisfied. [Step 2.a]

• DETERMINE CIAS is NOT present and DIRECT Shift Chemist to SAMPLE both SGs for activity. [Step 2.c]

NRC Simulator Scenario 4 Outline Rev. Final As Run

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 5, 6, & 7 Page 35 of 37 Event

Description:

RCP Trip / Automatic Reactor Trip Failure / Instrument Air Compressors Trip / Bearing Cooling Water Pump Trip / Steam Line Break inside Containment Time Position Applicants Actions or Behavior CRS IMPLEMENT the Emergency Plan. [Step 3]

• Time: __________

CREW MONITOR the Floating Steps. [Step 4]

DETERMINE RCS pressure > 1600 psia, Containment pressure < 5 psig, CRS with Steam Generator > 500 psia. [Step 5]

CRS DETERMINE RCS pressure 1600 psia. [Step 6]

CRS DETERMINE Containment pressure < 5 psig. [Step 7]

CRS DETERMINE SIAS has NOT actuated. [Step 8]

ATCO VERIFY RCP operating parameters: [Step 9]

• DETERMINE RCP RC-3A TRIPPED and TCOLD < 500°F. [Step 9.a]

• DETERMINE RCS pressure ~1900 psia. [Step 9.b]

• DETERMINE RCPs subcooling > 20°F. [Step 9.c]

ATCO VERIFY normal CCW/RW System operation: [Step 10]

• DETERMINE at least 2 CCW Pumps are RUNNING. [Step 10.a]

• DETERMINE CCW Pump discharge pressure 60 psig. [Step 10.b]

• ENSURE at least two Raw Water Pumps operating. [Step 10.c]

ATCO

• START at least one Raw Water Pump.

• DETERMINE at least three RW/CCW Heat Exchangers in service.

[Step 10.d]

• DETERMINE all RCP cooler CCW Valves OPEN. [Step 10.e]

CRS DETERMINE affected SG is RC-2A and SG pressure is < 700 psia. [Step 11]

CRS DETERMINE Uncontrolled Heat Extraction has NOT been isolated. [Step 12]

NRC Simulator Scenario 4 Outline Rev. Final As Run

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 5, 6, & 7 Page 36 of 37 Event

Description:

RCP Trip / Automatic Reactor Trip Failure / Instrument Air Compressors Trip / Bearing Cooling Water Pump Trip / Steam Line Break inside Containment Time Position Applicants Actions or Behavior

• [CA] DETERMINE Emergency Boration already in progress. [Step 12.1]

BOPO DETERMINE SG RC-2A and SG RC-2B both > 500 psia. [Step 13]

BOPO DETERMINE Steam Generator RC-2A is most affected SG. [Step 14]

CRS DETERMINE Uncontrolled Heat Extraction has NOT been isolated. [Step 15]

IF RC-2A is most affected, ISOLATE RC-2A by performing HR-19, CRS Isolate/Restore Steam Generator A. [Step 16]

Examiner Note: The following steps are from HR-19, Isolate/Restore Steam Generator A.

CRITICAL TASK Isolate the Affected Steam Generator to Prevent Excess Plant Cooldown and STATEMENT Reactivity Additions Prior to Steam Generator Level = 0% Wide Range Level.

CRITICAL TASK BOPO PERFORM the following to isolate Steam Generator RC-2A: [Step 1]

• ENSURE all the following valves are CLOSED: [Step 1.a]

• VERIFY HCV-1041A, RC-2A MSIV CLOSED.

• VERIFY HCV-1041C, RC-2A MSIV Bypass Valve CLOSED.

• VERIFY MS-291, Air Assisted Main Steam Safety Valve CLOSED.

• VERIFY FCV-1101, RC-2A Feed Regulating Valve CLOSED.

• VERIFY HCV-1105, Feed Regulating Bypass Valve CLOSED.

BOPO

• VERIFY HCV-1386, RC-2A Feed Header Isolation Valve CLOSED.

• VERIFY HCV-1103, Feed Regulating Block Valve CLOSED.

• VERIFY HCV-1388A, Blowdown Isolation Valve CLOSED.

• VERIFY HCV-1388B, Blowdown Isolation Valve CLOSED.

• CLOSE HCV-1107A, AFW Isolation Valve.

• CLOSE HCV-1107B, AFW Isolation Valve.

• CONTACT Auxiliary Operator to CLOSE MS-298, Steam Valves HCV-1041A & 1042A Packing Leakoff Line Isolation Valve in Room 81.

[Step 1.b]

• If sampling is NOT in progress, CLOSE both Sample Valves: [Step 1.c]

• HCV-2506A, RC-2A Blowdown Sample Isolation Valve

• HCV-2506B, RC-2A Blowdown Sample Isolation Valve NRC Simulator Scenario 4 Outline Rev. Final As Run

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 5, 6, & 7 Page 37 of 37 Event

Description:

RCP Trip / Automatic Reactor Trip Failure / Instrument Air Compressors Trip / Bearing Cooling Water Pump Trip / Steam Line Break inside Containment Time Position Applicants Actions or Behavior

• PERFORM the following to CLOSE YCV-1045A: [Step 1.d]

• PLACE ISOLATION VALVE YCV-1045A OVERRIDE SW in BOPO OVERRIDE. [Step 1.d.1)]

• PLACE control switch for S/G RC-2A STM TO FW-10 HDR A BOPO ISOLATION VALVE YCV-1045A in CLOSE. [Step 1.d.2)]

NOTE Air accumulators will maintain the valve in a closed position for 30 minutes after a loss of Instrument Air.

• CONTACT Auxiliary Operator to HANDJACK YCV-1045A, MAIN STEAM LINE "A" TO AUX FEEDWATER PUMP FW-10 SUPPLY VALVE to CLOSE in Room 81. [Step 1.d.3)]

VERIFY RC-2A is most affected SG per Attachment HR-18, Most Affected CRS Steam Generator Determination. [Step 2]

Examiner Note: If a Reactor Trip was performed due to high CCW or RCP bearing temperatures and RC-2A is isolated, INITIATE Event 4 (Alternate Path Event 8), Loss of Instrument Bus AI-40A, and REFER to Page 21 of 37.

When Steam Generator RC-2A is isolated, TERMINATE the scenario.

NRC Simulator Scenario 4 Outline Rev. Final As Run

ES-301 Transient and Event Checklist Form ES-301-5 Facility: Fort Calhoun Station Date of Exam: 12/07/15 Operating Test No.: NRC A E SCENARIOS P V P E FCS #1 FCS #3 FCS #4 L N T MINIMUM(*)

I T CREW CREW CREW CREW O C

POSITION POSITION POSITION POSITION T A T A

N Y S A B S A B S A B S A B L

T P R T O R T O R T O R T O R I U E O C P O C P O C P O C P RX - - 0 1 1 0 NOR - 1 1 1 1 1 SRO-U1 I/C 2,3,4, 3,6 6 4 4 2 5

MAJ 6 5,7 3 2 2 1 TS 2,4 - 2 0 2 2 RX - - 0 1 1 0 NOR - 1 1 1 1 1 SRO-U2 I/C 2,3,4, 2,4 6 4 4 2 5

MAJ 6 7 2 2 2 1 TS 2,4 2,4 4 0 2 2 RX - 6 - 1 1 1 0 NOR - - 1 1 1 1 1 SRO-I1 I/C 2,3,4, 2,4,5 6 8 4 4 2 5

MAJ 6 7 7 3 2 2 1 TS 2,4 - - 2 0 2 2 RX - - - 0 1 1 0 NOR 1 6 1 3 1 1 1 SRO-I2 I/C 2,4,8, 3,4,8 2,4 9 4 4 2 9

MAJ 6 7 7 3 2 2 1 TS - - 2,4 2 0 2 2 RX - - 1 1 1 1 0 NOR - 6 - 1 1 1 1 SRO-I3 I/C 3,5,7 2,3,4, 2,4 9 4 4 2 5

MAJ 6 7 7 3 2 2 1 TS - 4,5 - 2 0 2 2 RX - - 0 1 1 0 NOR 1 1 2 1 1 1 SRO-I4 I/C 2,4,8, 2,3,4 7 4 4 2 9

MAJ 6 5,7 3 2 2 1 TS - 2,3,4 3 0 2 2 FCS 2015 NRC ES-301-5 Transient and Event Checklist Final As Run

ES-301 Transient and Event Checklist Form ES-301-5 Facility: Fort Calhoun Station Date of Exam: 12/07/15 Operating Test No.: NRC A E SCENARIOS P V P E FCS #1 FCS #3 FCS #4 L N T MINIMUM(*)

I T CREW CREW CREW CREW O C

POSITION POSITION POSITION POSITION T A T A

N Y S A B S A B S A B S A B R T O R T O R T O R T O L T P R I U E O C P O C P O C P O C P RX - 1 1 1 1 0 NOR - - 0 1 1 1 RO-1 I/C 3,5,7 2,4 5 4 4 2 MAJ 6 7 2 2 2 1 TS - - 0 0 2 2 RX - 1 1 1 1 0 NOR - - 0 1 1 1 RO-2 I/C 3,5,7 2,4 5 4 4 2 MAJ 6 5,7 3 2 2 1 TS - - 0 0 2 2 RX - - 0 1 1 0 NOR 1 1 2 1 1 1 RO-3 I/C 2,4,8, 6 5 4 4 2 9

MAJ 6 7 2 2 2 1 TS - - 0 0 2 2 Instructions:

1. Check the applicant level and enter the operating test number and Form ES-D-1 event numbers for each event type; TS are not applicable for RO applicants. ROs must serve in both the at-the-controls (ATC) and balance-of-plant (BOP) positions; Instant SROs (SRO-I) must serve in both the SRO and the ATC positions, including at least two instrument or component (I/C) malfunctions and one major transient, in the ATC position. If an SRO-1 additionally serves in the BOP position, one I/C malfunction can be credited toward the two I/C malfunctions required for the ATC position.

2. Reactivity manipulations may be conducted under normal or controlled abnormal conditions (refer to Section D.5.d) but must be significant per Section C.2.a of Appendix D. (*) Reactivity and normal evolutions may be replaced with additional instrument or component malfunctions on a 1-for-1 basis.

3. Whenever practical, both instrument and component malfunctions should be included; only those that require verifiable actions that provide insight to the applicants competence count toward the minimum requirements specified for the applicants license level in the right-hand columns.

4. For licensees that use the ATC operator primarily for monitoring plant parameters, the chief examiner may place SRO-I applicants in either the ATC or BOP position to best evaluate the SRO-I in manipulating plant controls.

FCS 2015 NRC ES-301-5 Transient and Event Checklist Final As Run

ES-301 Competencies Checklist Form ES-301-6 Facility: FCS Date of Examination: 12/07/15 Operating Test No. NRC 1/3/4 Applicants SROU-1 SROU-2 SROI-1 Competencies SCENARIO SCENARIO SCENARIO 1 3 4 1 3 4 1 3 4 Interpret/Diag-2,3,4,5, 3,5,6, 2,3,4,5, 2,3,4,5, 2,4,5, nose Events 6 7 6

- 2,4 6 7 6,7 and Conditions Comply With 2,3,4,5, 1,3,5, 2,3,4,5, 1,2,4, 2,3,4,5, 2,4,5, and Use 6 6,7 6 7 6 6,7 1,6,7 Procedures (1)

Operate 1,3,5, 2,4,5, Control Boards N/A -

6,7 N/A - N/A N/A 6,7 1,6,7 (2)

Communicate 1,2,3,4, 1,3,5, 1,2,3,4, 1,2,4, 1,2,3,4, 2,4,5, and 5,6 6,7 5,6 7 5,6 6,7 1,6,7 Interact Demonstrate 2,3,4,5, 2,3,4,5, 1,2,4, 2,3,4,5, Supervisory 6

- N/A 6

7 6 N/A N/A Ability (3)

Comply With and Use Tech. 2,4 - N/A 2,4 - 2,4 2,4 N/A N/A Specs. (3)

Notes:

(1) Includes Technical Specification compliance for an RO.

(2) Optional for an SRO-U.

(3) Only applicable to SROs.

FCS 2015 NRC ES-301-6 Competencies Checklist Final As Run

ES-301 Competencies Checklist Form ES-301-6 Facility: FCS Date of Examination: 12/07/15 Operating Test No. NRC 1/3/4 Applicants SROI-2 SROI-3 SROI-4 Competencies SCENARIO SCENARIO SCENARIO 1 3 4 1 3 4 1 3 4 Interpret/Diag-2,4,5,6, 3,5,7, 2,3,4, 2,4,5,6, nose Events 8,9 8 2,4 3,5,6,7 5,7 2,4,7 8,9

- 2,3,4,5 and Conditions Comply With 1,2,4,5, 1,3,5, 1,2,4, 2,3,4, 1,2,4, 1,2,4,5, 1,2,3,4, and Use 6,8,9 6,7,8 7 3,5,6,7 5,6,7 7 6,8,9 5,7 Procedures (1)

Operate 1,2,4,5, 1,3,5, 1,2,4, 1,2,4,5, Control Boards 6,8,9 6,7,8 N/A 3,5,6,7 N/A 7 6,8,9

- N/A (2)

Communicate 1,2,3,4, 1,3,5, 1,2,4, 1,2,3,4, 1,2,4, 1,2,3,4, 1,2,3,4, and 5,6,8,9 6,7,8 7 3,5,6,7 5,6,7 7 5,6,8,9 5,7 Interact Demonstrate 1,2,4, 1,2,3,4, 1,2,3,4, Supervisory N/A N/A 7

N/A 5,6,7 N/A N/A -

5,7 Ability (3)

Comply With and Use Tech. N/A N/A 2,4 N/A 4,5 N/A N/A - 2,3,4 Specs. (3)

Notes:

(1) Includes Technical Specification compliance for an RO.

(2) Optional for an SRO-U.

(3) Only applicable to SROs.

FCS 2015 NRC ES-301-6 Competencies Checklist Final As Run

ES-301 Competencies Checklist Form ES-301-6 Facility: FCS Date of Examination: 12/07/15 Operating Test No. NRC 1/3/4 Applicants RO-1 RO-2 RO-3 Competencies SCENARIO SCENARIO SCENARIO 1 3 4 1 3 4 1 3 4 Interpret/Diag-2,4,5,6, nose Events 3,5,6,7 - 2,4,7 3,5,6,7 - 2,4,5,7 8,9

- 6,7 and Conditions Comply With 1,2,4, 1,2,4, 1,2,4,5, and Use 3,5,6,7 -

7 3,5,6,7 -

5,7 6,8,9

- 1,6,7 Procedures (1)

Operate 1,2,4, 1,2,4, 1,2,4,5, Control Boards 3,5,6,7 -

7 3,5,6,7 -

5,7 6,8,9

- 1,6,7 (2)

Communicate 1,2,4, 1,2,4, 1,2,3,4, and 3,5,6,7 -

7 3,5,6,7 -

5,7 5,6,8,9

- 1,6,7 Interact Demonstrate Supervisory N/A - N/A N/A - N/A N/A - N/A Ability (3)

Comply With and Use Tech. N/A - N/A N/A - N/A N/A - N/A Specs. (3)

Notes:

(1) Includes Technical Specification compliance for an RO.

(2) Optional for an SRO-U.

(3) Only applicable to SROs.

FCS 2015 NRC ES-301-6 Competencies Checklist Final As Run