Text

301 Exam Final Items (2 Year Delayed Release) (1)
	ML21256A293
Person / Time
Site:	Browns Ferry
Issue date:	09/13/2021
From:	; Tennessee Valley Authority
To:	; NRC/RGN-II
References
	Download: ML21256A293 (435)
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ML21256A293 Appendix D Scenario Outline Form ES-D1 Facility: BFN Scenario Number: NRC- 1 Op-Test Number: 21-04 Examiners: __________________________ Operators: SRO: _____________________

__________________________ ATC: _____________________

BOP: _____________________

Initial Conditions: 80% Reactor Power.

Turnover: Reactor Shutdown in progress. SRV 1-22 is INOPERABLE (ADS Valve). EHPM tagged for motor bearing inspection.

Critical Tasks:

1. With the Reactor at power and with a Primary System discharging into the Secondary Containment, manually SCRAM the Reactor before any area exceeds the Maximum Safe Temperature operating value by entering EOI-1, RPV Control.

2. With a Primary System discharging into the Secondary Containment, when two or more areas are greater than their Maximum Safe operating values for the same parameter, the Unit Operator initiates Emergency Depressurization as directed by the Nuclear Unit Senior Operator (NUSO).

Event Malfunction Event Event Description Number Number Type*

N-BOP

1. N/A Return Reactor Water Cleanup (RWCU) to Operation N-NUSO R-OATC

2. N/A Reduce Reactor Power to 75% using Core Flow R-NUSO C-BOP Reactor Building Closed Cooling Water (RBCCW) 3.#S XA-55-4C_13 C-NUSO Surge Tank Low Level SCHED C-OATC Reactor Water Cleanup (RWCU) Leak / One PCIV Fails 4.

RWCU TS-NUSO to Close

5. N/A TS-NUSO Core Spray Loop I Room Cooler EECW Leak C-OATC

6. ED08C 2C 4KV Unit Board Trip C-NUSO RC09 Un-isolable Reactor Core Isolation Cooling (RCIC)

7. FCV-71-2 M-ALL Steam Leak FCV-71-3 8.# TH23 M-ALL Fuel Damage C-BOP Reactor Feedwater Pumps (RFPTs) Trip / HPCI Fails to

9. FCV-73-16 C-NUSO Automatically Start and Inject

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor (TS)Technical Specification

Event on previous two NRC Exams

S Event on previous two NRC Exams Spare Scenario Unit 2 - Page 1 of 4

Appendix D Scenario Outline Form ES-D1 Events

1. The crew will return Reactor Water Cleanup (RWCU) to service in accordance with 2-OI-69, Reactor Water Cleanup System.

2. The crew will continue the Reactor shutdown and reduce Reactor Power using Core Flow to 75% in accordance with 2-OI-68, Reactor Recirculation System, and 2-GOI-100-12A, Unit Shutdown from Power Operation to Cold Shutdown and Reductions in Power During Power Operations.

3. The Reactor Building Closed Cooling Water (RBCCW) Surge Tank Low Level alarm will be received, requiring the crew to take action to fill the RBCCW Surge Tank in accordance with Alarm Response Procedure 2-ARP-9-4C, Window 13, RBCCW SURGE TANK LEVEL LOW.

4. A leak will develop in the Reactor Water Cleanup System (RWCU), requiring the crew to respond in accordance with 2-AOI-64-2A, Group 3 Reactor Water Cleanup Isolation. 2-FCV-69-1, RWCU INBOARD SUCTION ISOLATION VALVE, will fail to automatically close on an isolation signal, requiring manual action by the crew to isolate the RWCU System. The Nuclear Unit Senior Operator (NUSO) will address Technical Specification 3.6.1.3, Primary Containment Isolation Valves (PCIVs), Condition A.

5. The Reactor Building AUO will report that an Emergency Equipment Cooling Water (EECW) leak was found in the Loop I Core Spray Room Cooler and that the leak has been isolated, requiring the NUSO to address Technical Requirements Manual 3.5.3, Equipment Area Coolers, Condition A and Technical Specification 3.5.1, Emergency Core Cooling Systems - Operating, Condition F.

6. 4KV Unit Board 2C will trip, resulting in a loss of the following electrical loads:

2C Condensate Pump, 2C Condensate Booster Pump, 2C Raw Cooling Water (RCW) Pump, 2C Condenser Cooling Water (CCW) Pump, and 2A Control Rod Drive (CRD) Pump. Action will be required to restore CRD Flow in accordance with 2-AOI-85-3, CRD System Failure.

7. Reactor Core Isolation Cooling (RCIC) will develop an un-isolable leak, causing high temperatures and radiation levels in the Reactor Building. The SRO will respond in accordance with 2-EOI-3, Secondary Containment Control.

8. Fuel damage will occur when the Reactor SCRAMs, requiring the crew to Emergency Depressurize the Reactor due to two Area Radiation Levels exceeding their Maximum Safe Values in Secondary Containment.

9. Reactor Feedwater Pumps (RFPTs) will trip when the Reactor MODE SWITCH is placed in SHUTDOWN. When Reactor Water Level lowers to the initiation setpoint, the High Pressure Coolant Injection (HPCI) System will not automatically start, requiring the crew to take action to manually start HPCI for Reactor Water Level control.

Unit 2 - Page 2 of 4

Appendix D Scenario Outline Form ES-D1 The Scenario ends when the crew has Emergency Depressurized the Reactor and has control of Reactor Water level above the Top of Active Fuel ((-) 162 inches) using low pressure systems.

Critical Tasks 2

1. With the Reactor at power and with a Primary System discharging into the Secondary Containment, manually SCRAM the Reactor before any area exceeds the Maximum Safe Temperature operating value by entering EOI-1, RPV Control.

a. Safety Significance SCRAM reduces to decay heat the energy that the RPV may be discharging into the Secondary Containment.

b. Cues Procedural Compliance.

Secondary Containment Area Temperature, Level and/or radiation indication.

c. Measured by Observation - With a Primary System discharging into Secondary Containment a Reactor SCRAM is initiated before a Maximum Safe condition is reached.

d. Feedback Control Rod positions Reactor Power reduction

e. Critical Task Failure Criteria The operating crew fails to proceed without delay and in a controlled manner to initiate a Reactor SCRAM from the time it is announced that one Area Temperature is approaching the Maximum Safe value.

2. With a Primary System discharging into the Secondary Containment, when two or more areas are greater than their Maximum Safe operating values for the same parameter, the Unit Operator initiates Emergency Depressurization as directed by the Nuclear Unit Senior Operator.

a. Safety Significance Places the Primary System in the lowest possible energy state, rejects heat to the Suppression Pool in preference to outside the Primary Containment, and reduces driving head and flow of system discharging into the Secondary Containment.

b. Cues Procedural Compliance.

Secondary Containment Area Temperature, Level, or Radiation indication.

Unit 2 - Page 3 of 4

Appendix D Scenario Outline Form ES-D1

c. Measured by Observation - NUSO transitions to 2-EOI-C2, Emergency RPV Depressurization, and directs opening at least 6 MSRVs when two or more areas are greater than their Maximum Safe operating values for the same parameter.

d. Feedback Reactor Pressure Trend.

MSRV status indications.

e. Critical Task Failure Criteria The operating crew fails to proceed with without delay and in a controlled manner to initiate Emergency Depressurization when two Area Radiation Levels exceed Maximum Safe value.

Unit 2 - Page 4 of 4

Appendix D Scenario Outline Form ES-D1 Facility: BFN Scenario Number: NRC-1 Op-Test Number:21-04 Examiners: __________________________ Operators: SRO: _____________________

__________________________ ATC: _____________________

BOP: _____________________

Initial Conditions: 80% Reactor Power.

Turnover: Reactor Shutdown in progress. SRV 1-22 is INOPERABLE (ADS Valve). EHPM tagged for motor bearing inspection.

Critical Tasks:

1. With the Reactor at power and with a Primary System discharging into the Secondary Containment, manually SCRAM the Reactor before any area exceeds the Maximum Safe Temperature operating value by entering EOI-1, RPV Control.

2. With a Primary System discharging into the Secondary Containment, when two or more areas are greater than their Maximum Safe operating values for the same parameter, the Unit Operator initiates Emergency Depressurization as directed by the Nuclear Unit Senior Operator (NUSO).

Event Malfunction Event Event Description Number Number Type*

N-BOP

1. N/A Return Reactor Water Cleanup (RWCU) to Operation N-NUSO R-OATC

2. N/A Reduce Reactor Power to 75% using Core Flow R-NUSO C-BOP Reactor Building Closed Cooling Water (RBCCW) 3.#S XA-55-4C_13 C-NUSO Surge Tank Low Level SCHED C-OATC Reactor Water Cleanup (RWCU) Leak / One PCIV Fails 4.

RWCU TS-NUSO to Close

5. N/A TS-NUSO Core Spray Loop I Room Cooler EECW Leak C-OATC

6. ED08C 3C 4KV Unit Board Trip C-NUSO RC09 Un-isolable Reactor Core Isolation Cooling (RCIC)

7. FCV-71-2 M-ALL Steam Leak FCV-71-3 8.# TH23 M-ALL Fuel Damage C-BOP Reactor Feedwater Pumps (RFPTs) Trip / HPCI Fails to

9. FCV-73-16 C-NUSO Automatically Start and Inject

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor (TS)Technical Specification

Event on previous two NRC Exams

S Event on previous two NRC Exams Spare Scenario Unit 3 - Page 1 of 4

Appendix D Scenario Outline Form ES-D1 Events

1. The crew will return Reactor Water Cleanup (RWCU) to service in accordance with 3-OI-69, Reactor Water Cleanup System.

2. The crew will continue the Reactor shutdown and reduce Reactor Power using Core Flow to 75% in accordance with 3-OI-68, Reactor Recirculation System, and 3-GOI-100-12A, Unit Shutdown from Power Operation to Cold Shutdown and Reductions in Power During Power Operations.

3. The Reactor Building Closed Cooling Water (RBCCW) Surge Tank Low Level alarm will be received, requiring the crew to take action to fill the RBCCW Surge Tank in accordance with Alarm Response Procedure 3-ARP-9-4C, Window 13, RBCCW SURGE TANK LEVEL LOW.

4. A leak will develop in the Reactor Water Cleanup System (RWCU), requiring the crew to respond in accordance with 3-AOI-64-2A, Group 3 Reactor Water Cleanup Isolation. 3-FCV-69-1, RWCU INBOARD SUCTION ISOLATION VALVE, will fail to automatically close on an isolation signal, requiring manual action by the crew to isolate the RWCU System. The Nuclear Unit Senior Operator (NUSO) will address Technical Specification 3.6.1.3, Primary Containment Isolation Valves (PCIVs), Condition A.

5. The Reactor Building AUO will report that an Emergency Equipment Cooling Water (EECW) leak was found in the Loop I Core Spray Room Cooler and that the leak has been isolated, requiring the NUSO to address Technical Requirements Manual 3.5.3, Equipment Area Coolers, Condition A and Technical Specification 3.5.1, Emergency Core Cooling Systems - Operating, Condition F.

6. 4KV Unit Board 3C will trip, resulting in a loss of the following electrical loads:

3C Condensate Pump, 3C Condensate Booster Pump, 3C Raw Cooling Water (RCW) Pump, 3C Condenser Cooling Water (CCW) Pump, and 3A Control Rod Drive (CRD) Pump. Manual action will be required to restore CRD Flow in accordance with 3-AOI-85-3, CRD System Failure.

7. Reactor Core Isolation Cooling (RCIC) will develop an un-isolable leak, causing high temperatures and radiation levels in the Reactor Building. The SRO will respond in accordance with 3-EOI-3, Secondary Containment Control.

8. Fuel damage will occur when the Reactor SCRAMs, requiring the crew to Emergency Depressurize the Reactor due to two Area Radiation Levels exceeding their Maximum Safe Values in Secondary Containment.

9. Reactor Feedwater Pumps (RFPTs) will trip when the Reactor MODE SWITCH is placed in SHUTDOWN. When Reactor Water Level lowers to the initiation setpoint, the High Pressure Coolant Injection (HPCI) System will not automatically start, requiring the crew to take action to manually start HPCI for Reactor Water Level control.

Unit 3 - Page 2 of 4

Appendix D Scenario Outline Form ES-D1 The Scenario ends when the crew has Emergency Depressurized the Reactor and has control of Reactor Water level above the Top of Active Fuel ((-) 162 inches) using low pressure systems.

Critical Tasks 2

1. With the Reactor at power and with a Primary System discharging into the Secondary Containment, manually SCRAM the Reactor before any area exceeds the Maximum Safe Temperature operating value by entering EOI-1, RPV Control.

a. Safety Significance SCRAM reduces to decay heat the energy that the RPV may be discharging into the Secondary Containment.

b. Cues Procedural Compliance.

Secondary Containment Area Temperature, Level and/or radiation indication.

c. Measured by Observation - With a Primary System discharging into Secondary Containment a Reactor SCRAM is initiated before a Maximum Safe condition is reached.

d. Feedback Control Rod positions Reactor Power reduction

e. Critical Task Failure Criteria The operating crew fails to proceed without delay and in a controlled manner to initiate a Reactor SCRAM from the time it is announced that one Area Temperature is approaching the Maximum Safe value.

2. With a Primary System discharging into the Secondary Containment, when two or more areas are greater than their Maximum Safe operating values for the same parameter, the Unit Operator initiates Emergency Depressurization as directed by the Nuclear Unit Senior Operator.

a. Safety Significance Places the Primary System in the lowest possible energy state, rejects heat to the Suppression Pool in preference to outside the Primary Containment, and reduces driving head and flow of system discharging into the Secondary Containment.

b. Cues Procedural Compliance.

Secondary Containment Area Temperature, Level, or Radiation indication.

Unit 3 - Page 3 of 4

Appendix D Scenario Outline Form ES-D1

c. Measured by Observation - NUSO transitions to 3-EOI-C2, Emergency RPV Depressurization, and directs opening at least 6 MSRVs when two or more areas are greater than their Maximum Safe operating values for the same parameter.

d. Feedback Reactor Pressure Trend.

MSRV status indications.

e. Critical Task Failure Criteria The operating crew fails to proceed with without delay and in a controlled manner to initiate Emergency Depressurization when two Area Radiation Levels exceed Maximum Safe value.

Unit 3 - Page 4 of 4

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 1 Page 1 of 9 Event

Description:

Return Reactor Water Cleanup (RWCU) to Operation Time Position Applicants Actions or Behavior PRIOR to placing the simulator in RUN, start CPERF to record Driver critical parameters.

If the crew does not start Event 1, Return Reactor Water Cleanup (RWCU) to Operation, request that the Driver contact the Nuclear NRC Unit Senior Operator (NUSO) as the Shift Manager and direct the crew to return RWCU to operation.

If requested by the Chief Examiner, contact the NUSO as the Shift Manager and direct the crew to return RWCU to operation.

Driver If contacted by the crew as the Reactor Building Assistant Unit Operator (AUO) acknowledge any direction given.

If Panel 2-9-4B, Window 17, RWCU NON-REGENERATIVE HX DISCH NRC TEMP HIGH, is received, see page 9 of 57 for Alarm Response Procedure actions.

NUSO Directs the Balance of Plant Operator (BOP) to return RWCU to service.

2-OI-69, Reactor Water Cleanup System Section 5.1 RWCU Pump Startup NOTES

1) All controls and indications are located on Panel 2-9-4 unless noted otherwise.

2) RWCU is required to be operated within the following restrictions with Reactor Pressure 50 psig (MODES 2 or 3), or any time the unit BOP is in MODE 4, MODE 5, or de-fueled:

One pump in operation, pump can be operated to its maximum flow capacity Two pumps in operation, maximum flow limited to 100 gpm per pump (200 gpm total)

[4] ENSURE RESET the RWCU Group 3 Isolation using 2-HS-64-16A-S32, PCIS DIVISION I RESET and 2-HS-64-16A-S33 PCIS DIVISION II RESET, at Panel 2-9-4.

Page 1 of 57 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 1 Page 2 of 9 Event

Description:

Return Reactor Water Cleanup (RWCU) to Operation Time Position Applicants Actions or Behavior

[5] CHECK the following on Panel 2-LPNL-925-0003, Unit 2 Reactor Building Elevation 621:

[5.1] Demin 2A and/or 2B Holding Pumps are running BOP (2-HS-069-6015 and 2-HS-069-6005).

[5.2] Demin 2A and/or 2B Outlet Valves are closed (2-HS-069-0035 and/or 2-HS-069-0060).

If contacted as the Reactor Building AUO to perform Step [5],

Driver acknowledge the direction and report that the Holding Pumps are running with the Demineralizer Outlet Valves closed.

[6] N/A

[7] ENSURE 2-TIC-069-0010A, RWCU HEAT EXCHANGERS RBCCW BOP FLOW CONTROL, is in MANUAL, and FULL OPEN demand is on 2-TCV-70-49, RWCU NON-REGENERATIVE HEAT EXCHANGER OUTLET TCV.

When directed to place 2-TIC-069-0010A in manual, insert Event 11.

Driver Inform the crew that 2-TIC-069-0010A is in manual and is fully open.

[8] ENSURE CLOSED the following:

2-HC-69-15, RWCU BLOWDOWN PRESSURE CONTROL VALVE 2-HS-69-16, RWCU BLOWDOWN TO MAIN CONDENSER BOP 2-HS-69-17, RWCU BLOWDOWN TO RADWASTE

[9] ENSURE the DEFEAT/OPERATE SWITCH FOR 2-HC-069-0015 in the DEFEAT position, using 2-HS-069-0015A.

[10] N/A

[11] NOTIFY Chemistry that RWCU is being placed in service and to check the durability monitor.

Page 2 of 57 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 1 Page 3 of 9 Event

Description:

Return Reactor Water Cleanup (RWCU) to Operation Time Position Applicants Actions or Behavior When contacted as Chemistry acknowledge the direction to check Driver the durability monitor.

[12] ENSURE OPEN the following:

[12.1] 2-FCV-69-1, RWCU INBD SUCT ISOLATION VALVE.

[12.2] 2-FCV-69-2, RWCU OUTBD SUCT ISOLATION VALVE.

[12.3] 2-FCV-69-8, RWCU DEMIN BYPASS VALVE.

[13] OPEN 2-FCV-069-0012, RWCU RETURN ISOLATION VALVE by one of the two methods described below.

THROTTLE OPEN 2-FCV-069-0012, RWCU RETURN ISOLATION VALVE as follows:

PLACE 2-HS-69-12A in the OPEN position, THEN WHEN intermediate position (red and green light) is indicated, THEN RETURN 2-HS-69-12A to the NORM position FULLY OPEN RWCU RETURN ISOLATION VALVE, BOP 2-FCV-069-0012 as follows:

ENSURE 2-FCV-069-0012 is OPEN NOTES

1) Too high a flow on startup after isolation could cause 2-TIS-69-11 to actuate due to a high Non-Regenerative Heat Exchanger Outlet Temperature (2-XS-69-6, RWCU TEMP SELECT, Position 3, WATER TO RWCU DEMINS).

2) The RWCU Pump trips on low flow at 56 gpm, after a 30 second time delay. Failure to immediately raise flow to greater than 56 gpm in the following steps results in a pump trip.

[14] PLACE seal purge in operation to pump(s) to be placed in service.

(REFER TO Section 8.2)

Page 3 of 57 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 1 Page 4 of 9 Event

Description:

Return Reactor Water Cleanup (RWCU) to Operation Time Position Applicants Actions or Behavior When directed to place seal purge in service, acknowledge the Driver direction and inform the crew that seal purge has been placed in service in accordance with 2-OI-69, Section 8.2

[15] START RWCU PUMP 2A(2B) using 2-HS-69-4A(B)-A, AND RAISE flow, using 2-HS-69-12A, RWCU RETURN ISOLATION VALVE, to prevent low flow trip.

[16] IF two pump operation is desired, THEN START the second RWCU BOP PUMP 2B(2A) using 2-HS-69-4B(A)-A, AND RAISE flow using RWCU RETURN ISOLATION VALVE, 2-HS-69-12A, to prevent low flow trip.

[17] WHEN desired, THEN PLACE RWCU filter-demineralizers in service. (REFER TO Section 6.2) 2-OI-69, Reactor Water Cleanup System Section 6.2, Placing Filter-Demineralizers in Service NOTES

1) To prevent resin intrusion into the Reactor Vessel, the resin trap (Filter Effluent Strainer) on a vessel that was removed from service due to high D/P should be backwashed prior to returning the demin BOP vessel to service.

2) The flow rate in the mechanical seal areas can be to a maximum of 8.0 gpm for a 24 to 36 hour4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months period. Based on the layout design showing the flush coming in over the retainer at the back of the mechanical seal, the rise in flush gpm should NOT cause any operational problems with the mechanical seal when conditions are returned to normal.

Page 4 of 57 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 1 Page 5 of 9 Event

Description:

Return Reactor Water Cleanup (RWCU) to Operation Time Position Applicants Actions or Behavior CAUTION When initially placing a filter-demineralizer into service, it is desirable that most RWCU Discharge Flow be returned to the Main Condenser.

If the Reactor is pressurized, however, failure to follow temperature BOP restrictions could result in thermal shocking the Regenerative Heat Exchanger.

[1] REVIEW Precautions and Limitations in Section 3.0.

[2] - [10] Performed in the Field by an AUO.

When contacted as the Reactor Building AUO to prepare to roll in RWCU Demineralizer, acknowledge the direction and report that you are standing by with Steps complete Steps [2] through [10] of 2-OI-69, Reactor Water Cleanup System, Section 6.2 are complete.

When directed to place filter-demineralizers in service, insert Driver Event 1 to perform AUO actions to place demineralizers in service, and inform the crew that Demin Flow is rising.

Demineralizers will roll in over a 1-minute time frame - when complete inform the crew that RWCU filter-demineralizers have been placed in service.

If requested, Demineralizer flows are rising and are normal.

NOTE RWCU is required to be operated within the following restrictions with Reactor Pressure 50 psig (Modes 2 or 3), or any time the unit is in Mode 4, Mode 5, or defueled:

BOP One pump in operation, pump can be operated to its maximum flow capacity Two pumps in operation, maximum flow limited to 100 gpm per pump (200 gpm total)

Page 5 of 57 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 1 Page 6 of 9 Event

Description:

Return Reactor Water Cleanup (RWCU) to Operation Time Position Applicants Actions or Behavior

[11] PERFORM the following simultaneously:

BOP CLOSE 2-HS-69-8, RWCU DEMIN BYPASS VALVE on Panel 2-9-4 Verify that the crew is able to clear RWCU Demineralizer Alarm (Panel 2-9-4B, Window 24) - it will automatically reset on Event 1 Driver after 15 seconds. If the crew cannot get the alarm to reset, insert remote function AN01E to RESET again).

[12] RAISE flow through the Demin until the desired flow has been established.

[13] ENSURE 2-HS-069-6015(6005), DEMIN 2A(2B) HOLDING PUMP, in the AUTO position.

BOP

[14] CHECK that Holding Pump 2A(2B), on the Demin being placed in service, has STOPPED.

[15] CHECK 2-FCV-069-0035B(0060B), DEMIN 2A(2B) HOLDING PUMP DISCH VLV H, has CLOSED.

When directed to perform Steps [12], [13], and [14] acknowledge Driver the direction and inform the crew that Steps [12], [13], and [14] are complete.

[16] NOTIFY Chemistry that the filter-demineralizer is in service AND REQUEST a sample for conductivity and silica of the effluent.

[17] CHECK that the results of the filter-demineralizer effluent sample BOP taken by Chemistry are within the limits of CI-13.1, Chemistry Manual, for return to the Reactor Vessel.

[17.1] CHECK RWCU Flows on ICS per Section 8.16.

When contacted as Chemistry, acknowledge any information or Driver direction given.

Page 6 of 57 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 1 Page 7 of 9 Event

Description:

Return Reactor Water Cleanup (RWCU) to Operation Time Position Applicants Actions or Behavior Continuing 2-OI-69, Reactor Water Cleanup System Section 5.1 RWCU Pump Startup

[18] ADJUST 2-HS-69-12A, RWCU RETURN ISOLATION VALVE, using as required to maintain system parameters within limits specified in this procedure.

CAUTION BOP Failure to maintain RWCU Non-Regenerative Heat Exchanger tube side outlet temperature below 130 F will reduce resin efficiency and may result in resin damage.

[19] THROTTLE blowdown flow as required to maintain the following parameters (REFER TO Section 6.5):

Desired Reactor Water Level Non-regenerative Heat Exchanger Tube Outlet Temperature less than 130 F

[20] IF at Operations Management discretion it is desired to place 2-TIC-069-0010A, RWCU HEAT EXCHANGERS RBCCW FLOW BOP CONTROL in AUTO, THEN PERFORM the following:

[20.1] PLACE 2-TIC-069-0010A, RWCU HEAT EXCHANGERS RBCCW FLOW CONTROL, in AUTO (REFER TO Section 8.14).

If contacted by the crew to place 2-TIC-069-0010A in automatic Driver acknowledge the direction and inform the crew that 2-TIC-069-0010A has been placed in automatic.

Page 7 of 57 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 1 Page 8 of 9 Event

Description:

Return Reactor Water Cleanup (RWCU) to Operation Time Position Applicants Actions or Behavior NOTES Seal water to the RWCU Pumps has been observed to slightly lower after pump(s) are placed in service When the Reactor Vessel is at atmospheric pressure and RWCU Pump seal water is being supplied by CS&S system, RWCU Pump seal water flow may decrease to 0 gpm after the BOP RWCU Pump has started. See PRECAUTION P&L 3.6E

[21] ENSURE SEAL WATER TO RWCU PUMPS, at Panel 2-25-314, is within 1.8 to 2.0 gpm (REFER TO Section 8.2).

2-FI-085-0075, RWCU PUMP 2A PURGE WATER FLOW INDICATOR 2-FI-085-0077, RWCU PUMP 2B SEAL WATER When contacted as the Reactor Building AUO to perform Step [21],

Driver inform the crew that seal water flow is 1.9 gpm.

The restoration of RWCU ICS point to the heat balance is not NRC required if automatically bypassed.

Page 8 of 57 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 1 Page 9 of 9 Event

Description:

Return Reactor Water Cleanup (RWCU) to Operation Time Position Applicants Actions or Behavior 2-9-ARP-4B, Alarm Response Procedure RWCU NON-REGENERATIVE HX DISCHARGE TEMPERATURE HIGH, Window 17 Operator Action:

A. CHECK 2-XS-69-6, RWCU NRHX Discharge Temperature, on Panel 2-9-4.

B. CHECK RBCCW System Temperature indication normal, Panel 2-9-4.

C. IF temperature continues to rise, THEN PERFORM the following, otherwise, MARK steps N/A:

BOP REDUCE system flow or reject flow as necessary to control temperature.

REFER TO 2-OI-69, Reactor Water Cleanup System D. DISPATCH personnel to check the following:

RWCU Heat Exchangers RBCCW Flow Controller (normally in auto with setpoint at approximately 110 F), located on Panel 25-2 Rx Bldg 593 2-TCV-70-49, RWCU NON-REGENERATIVE HEAT EXCHANGER OUTLET TCV operating properly (RBCCW to NRHX), located in RWCU HX room E. N/A If contacted by the crew to check equipment in Step D (see above),

acknowledge the direction and report the following as required:

Driver RWCU Heat Exchangers RBCCW Flow Controller is set at 110 F and is in automatic 2-TCV-70-49 is operating properly End of Event 1. Proceed to Event 2, Reduce Reactor Power to 75%

NRC using Core Flow.

Page 9 of 57 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 2 Page 1 of 4 Event

Description:

Reduce Reactor Power to 75% using Core Flow Time Position Applicants Actions or Behavior If the crew does not proceed to Event 2, Reduce Reactor Power to 75% using Core Flow, request that the Driver contact the NRC NUSO as the Shift Manager and direct the crew to reduce Reactor Power to 75%.

If requested by the Chief Examiner, contact the crew as the Shift Manager and direct the crew to continue with Step [2] of the Driver Reactivity Control Plan (RCP) for the Reactor Shutdown and reduce Reactor Power to 75% using Core Flow.

The crew may elect to conduct a re-focus reactivity brief prior to NRC lowering Reactor Power.

Directs the Operator at the Controls (OATC) to lower Reactor Power to 75% in accordance with Step [2] of RCP U2-2104NRC1 and in accordance with 2-OI-68, Reactor Recirculation System, NUSO 2-GOI-100-12A, Unit Shutdown from Power Operation to Cold Shutdown and Reductions in Power During Power Operations and 2-GOI-100-12, Power Maneuvering.

2-GOI-100-12A, Unit Shutdown from Power Operation to Cold Shutdown and Reductions in Power During Power Operations Section 5.3, Power Reduction 5.3.1 Reducing Reactor Power to 40%

OATC

[1] ENSURE the operators are using Attachment 9, Operations Down Power Monitoring.

[2] REDUCE Reactor Power by combination of Control Rod insertions and Core Flow changes, as recommended by Reactor Engineer.

Page 10 of 57 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 2 Page 2 of 4 Event

Description:

Reduce Reactor Power to 75% using Core Flow Time Position Applicants Actions or Behavior 2-GOI-100-12, Power Maneuvering Section 5.0, Instruction Steps OATC [7] REDUCE Reactor Power by combination of Control Rod insertions and Core Flow changes, as recommended by Reactor Engineer.

REFER TO 2-SR-3.1.3.5(A), Control Rod Coupling Integrity Check and 2-OI-68, Reactor Recirculation System.

2-OI-68, Reactor Recirculation System 3.0 Precautions and Limitations Section 3.5.3, Dual Pump Operation E. When raising (lowering) Reactor Power per the Reactivity Plan, the following guideline should be used to establish the 60 rpm mismatch between the Recirc Pumps.

1. When the first Recirc Pump reaches 1200 rpm (1300 rpm) or both Recirc Pumps are at 1200 rpm (1300 rpm) then individual controls will be used.

2. While following the Reactivity Plan establish the 60 rpm NRC mismatch using the individual controls for the leading Recirc Pump.

3. While maintaining the 60 rpm mismatch and using the Reactivity Plan, raise (lower) the Recirc Pump speeds using either the Master Controllers or Individual Controllers.

4. Once a Recirc Pump reaches 1300 rpm (1200 rpm) the 60 rpm mismatch is no longer required. While following the Reactivity Plan raise (lower) the lagging Recirc Pump using the individual controller match Recirc Pump speeds.

5. Once the Recirc Pumps are matched, then the speeds may be adjusted as required using the Master Controller or Individual Controllers while maintaining the requirements of Attachment 1.

Page 11 of 57 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 2 Page 3 of 4 Event

Description:

Reduce Reactor Power to 75% using Core Flow Time Position Applicants Actions or Behavior NRC Page 12 of 57 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 2 Page 4 of 4 Event

Description:

Reduce Reactor Power to 75% using Core Flow Time Position Applicants Actions or Behavior 2-OI-68, Reactor Recirculation System Section 6.2, Adjusting Recirc Flow NOTES

1) Thermal Limits are shown in 0-TI-248, Station Reactor Engineer and 2-SR-2, Instrument Checks and Observations.

2) Recirc Flow changes made during the later part of the operating cycle (Coastdown) could cause Core Flow values to approach or exceed the allowable values of the Increased Core Flow (ICF)

Region of the power to flow map. Instruments used to monitor pump speed and Core Flow should be identified in the Reactivity Control Plan. These values should be recorded prior to reducing Core Flow and used as a benchmark to reestablish the previous conditions when returning to power. Increased caution should be used when changes in Recirc Flow are made in this area.

OATC

[1] IF desired to control Recirc Pumps 2A and/or 2B speed with Recirc Individual Control, THEN PERFORM the following:

LOWER Recirc Pump 2A using 2-HS-96-17A(17B)(17C),

SLOW(MEDIUM)(FAST), (Otherwise N/A)

AND/OR LOWER Recirc Pump 2B using 2-HS-96-18A(18B)(18C),

SLOW(MEDIUM)(FAST). (Otherwise N/A)

[2] WHEN desired to control Recirc Pumps 2A and/or 2B speed with the RECIRC MASTER CONTROL, THEN ADJUST Recirc Pump Speed 2A & 2B using the following pushbuttons as required.

2-HS-96-33, LOWER SLOW 2-HS-96-34, LOWER MEDIUM 2-HS-96-35, LOWER FAST When satisfied with power reduction, end of Event 2. Request NRC that the Driver insert Event 3, Reactor Building Closed Cooling Water (RBCCW) Surge Tank Low Level.

Page 13 of 57 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 3 Page 1 of 1 Event

Description:

Reactor Building Closed Cooling Water (RBCCW) Surge Tank Low Level Time Position Applicants Actions or Behavior When requested by the Chief Examiner, insert Event 3, Reactor Driver Building Closed Cooling Water (RBCCW) Surge Tank Low Level Acknowledges and reports the following alarm to the NUSO:

BOP RBCCW SURGE TANK LEVEL LOW, 2-9-4C, Window 13 Directs the Balance of Plant Operator (BOP) to respond in NUSO accordance with the appropriate Alarm Response Procedure.

Alarm Response Procedure, 2-ARP-9-4C RBCCW SURGE TANK LEVEL LOW, Window 13 A. ADD water to the RBCCW Surge Tank for approximately one minute or until low level alarm resets using the following:

2-FCV-70-1, RBCCW SYS SURGE TANK FILL VALVE (Panel 2-9-4) OR BOP 2-HCV-2-1369, FCV-70-1 BYPASS VALVE (locally)

B. IF alarm does NOT reset, THEN CHECK the tank locally.

C. IF unable to maintain RBCCW Surge Tank level, THEN REFER TO 2-AOI-70-1, Loss of Reactor Building Closed Cooling Water.

D. IF necessary to add water more than once per shift, THEN CHECK Drywell floor drain system for excessive operation AND INSPECT system outside the Drywell for leakage.

The RBCCW Surge Tank Low Level alarm can be cleared 15 NRC seconds after the fill valve is opened.

Opens 2-FCV-70-1, RBCCW SYS SURGE TANK FILL VALVE for BOP approximately one minute and checks that RBCCW SURGE TANK LEVEL LOW 2-9-4C, Window 13 clears.

If contacted as the Reactor Building AUO to check for leaks or check RBCCW Surge Tank Level locally, acknowledge the Driver direction. Wait 2 minutes and report that Surge Tank Level is normal and there are no leaks.

End of Event 3. Request that the Driver insert Event 4, Reactor NRC Water Cleanup (RWCU) Leak / One PCIV Fails to Close.

Page 14 of 57 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 4 Page 1 of 12 Event

Description:

Reactor Water Cleanup (RWCU) Leak / One PCIV Fails to Close Time Position Applicants Actions or Behavior When requested by the Chief Examiner, insert Event 4, Reactor Driver Water Cleanup (RWCU) Leak / One PCIV Fails to Close.

2-FCV-69-2, RWCU OUTBOARD SUCTION ISOLATION VALVE, NRC will not automatically close on the Primary Containment System Isolation (PCIS) signal, but can be closed manually.

Acknowledges and reports the following alarms:

RWCU LEAK DETECTION TEMP HIGH, 2-9-3D, Window 17 OATC/

BOP RWCU ISOL LOGIC CHANNEL A TEMP HIGH, 2-9-5B, Window 32 RWCU ISOL LOGIC CHANNEL B TEMP HIGH, 2-9-5B, Window 33 Directs the BOP to respond in accordance with applicable Alarm NUSO Response Procedures.

2-ARP-9-3D, Alarm Response Procedure RWCU LEAK DETECTION TEMP HIGH, Window 17 Operator Action:

BOP A. IF this alarm is received in conjunction with RWCU ISOL LOGIC CHANNEL A TEMP HIGH [2-XA-55-5B, Window 32] and RWCU ISOL LOGIC CHANNEL B TEMP HIGH [2-XA-55-5B, Window 33],

THEN EXIT this procedure and GO TO 2-ARP-9-5B. Otherwise, CONTINUE in this procedure.

Exits 2-ARP-9-3D, Alarm Response Procedure, and enters BOP 2-ARP-9-5B, Alarm Response Procedure.

Page 15 of 57 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 4 Page 2 of 12 Event

Description:

Reactor Water Cleanup (RWCU) Leak / One PCIV Fails to Close Time Position Applicants Actions or Behavior 2-ARP-9-5B, Alarm Response Procedure RWCU ISOL LOGIC CHANNEL A TEMP HIGH, Window 32 Operator Action:

A. CHECK alarm by checking:

1. Analog Trip Units (ATUs) on Panel 2-9-83 and 2-9-85.

2. RWCU LEAK DETECTION TEMP HIGH annunciator in alarm BOP (2-XA-55-3D, Window 17).

3. Area temperature indications on 2-TR-69-29, LEAK DETECTION SYSTEM TEMPERATURE, on Panel 2-9-22.

4. Area Radiation Monitors (ARMs) 2-RR-90-1, 2-RR-90-50B on Panel 2-9-2 and 2-RR-90-250 on Panel 1-9-44.

5. ICS HPTURB & RWCU mimics for the 834 and 835 temperature loops.

B. IF leak is suspected, THEN MANUALLY ISOLATE RWCU.

If contacted as Unit 1 Operator to check Area Radiation Monitors Driver or Radiation Recorders, acknowledge the request.

Determines that 2-FCV-69-2, RWCU OUTBOARD SUCTION BOP ISOLATION VALVE, failed to automatically close and manually closes 2-FCV-69-2. Informs the NUSO.

C. IF RWCU automatically isolates, THEN REFER TO 2-AOI-64-2A, Group 3 Reactor Water Cleanup Isolation.

BOP D. IF TIS-69-835A(C) indicates greater than 131 F, THEN ENTER 2-EOI-3, Secondary Containment Control.

E. REFER TO Tech Spec Table 3.3.6.1-1, Primary Containment Isolation Instrumentation.

NUSO F. EVALUATE equipment associated with this alarm to determine compensatory actions required to maintain REP function. REFER TO NPG-SPP-18.3.5, Equipment Important to Emergency Response.

Page 16 of 57 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 4 Page 3 of 12 Event

Description:

Reactor Water Cleanup (RWCU) Leak / One PCIV Fails to Close Time Position Applicants Actions or Behavior As required by 2-ARP-9-5B, Window 32 and 2-ARP-9-5B, Window 33, references NPG-SPP-18.3.5, Equipment Important to NUSO Emergency Response. No actions are required in accordance with this procedure.

NRC Technical Specifications are covered starting on page 21 of 57.

2-ARP-9-5B, Alarm Response Procedure RWCU ISOL LOGIC CHANNEL B TEMP HIGH, Window 33 Operator Action:

A. CHECK alarm by checking:

1. ATUs on Panel 2-9-84 and 2-9-86.

2. RWCU LEAK DETECTION TEMP HIGH annunciator in ALARM (2-XA-55-3D, Window 17).

3. Area temperature indications on 2-TR-69-29, LEAK DETECTION SYSTEM TEMPERATURE, on Panel 2-9-22.

BOP 4. ARMs 2-RR-90-1, 2-RR-90-50B on Panel 2-9-2 and 2-RR-90-250 on Panel 1-9-44.

5. ICS HPTURB & RWCU mimics for the 834 and 835 temperature loops.

B. IF a leak is suspected, THEN MANUALLY ISOLATE RWCU.

C. IF RWCU automatically isolates, THEN REFER TO 2-AOI-64-2A, Group 3 Reactor Water Cleanup Isolation.

D. IF TIS-69-835B(D) indicates greater than 131 F, THEN ENTER 2-EOI-3, Secondary Containment Control.

E. REFER TO Technical Specification Table 3.3.6.1-1, Primary Containment Isolation Instrumentation.

No actions are required in accordance with Technical NRC Specification 3.3.6.1.

Page 17 of 57 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 4 Page 4 of 12 Event

Description:

Reactor Water Cleanup (RWCU) Leak / One PCIV Fails to Close Time Position Applicants Actions or Behavior 2-AOI-64-2A, Group 3 Reactor Water Cleanup Isolation Immediate Actions

[1] PERFORM the following:

ENSURE CLOSED 2-FCV-69-1, RWCU INBD SUCTION ISOLATION VALVE ENSURE CLOSED 2-FCV-69-2, RWCU OUTBD SUCTION ISOLATION VALVE BOP ENSURE CLOSED 2-FCV-69-12, RWCU RETURN ISOLATION VALVE ENSURE TRIPPED Reactor Water Cleanup Recirc Pumps 2A and 2B Subsequent Actions

[1] IF any EOI entry condition is met, THEN ENTER appropriate EOI(s).

The NUSO may enter 2-EOI-3, Secondary Containment Control, if NRC Area Temperature or Radiation exceeds the Maximum Normal value. See page 23 of 57 for 2-EOI-3 actions.

If contacted as an AUO to check ATUs, acknowledge the direction. Wait 3 minutes and report that ATUs 2-TIS-69-835A-D Driver indicate 160 degrees and lowering.

If contacted as Work Control or Maintenance, acknowledge any direction or reports given.

[2] CHECK the following to confirm high area temperature condition exists:

BOP 2-TR-69-29, LEAK DETECTION SYSTEM TEMPERATURE, (Panel 2-9-22)

ATUs in Auxiliary Instrument Room Page 18 of 57 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 4 Page 5 of 12 Event

Description:

Reactor Water Cleanup (RWCU) Leak / One PCIV Fails to Close Time Position Applicants Actions or Behavior

[3] IF isolation is caused by high area temperature, THEN DETERMINE if a line break exists by:

RWCU ARMs 2-RI-90-9A, 13A, and 14A Visual Observation Rx Zone Exhaust Rad Monitors 2-RE-90-142A, 142B, 143A, and 143B

[4] PERFORM necessary Heat Balance adjustments. REFER TO 2-OI-69, Reactor Water Cleanup System.

[5] CHECK the following monitors for a rise in activity:

A. 2-RR-90-1, AREA RADIATION, Points 9, 13, and 14 (Panel 2-9-2).

B. AIR PARTICULATE MONITOR CONSOLE, 2-MON-90-50, 2-RM-90-55 and 57 (Panel 2-9-2).

C. Reactor Building, Turbine Building, and Refuel Zone Exh Rad BOP on 0-MON-90-361, CHEMISTRY CAM MONITOR CONTROLLER (Panel 1-9-2).

[6] IF it has been determined that leakage is the cause of the isolation, THEN NOTIFY RADCON of RWCU status.

[7] NOTIFY Chemistry that RWCU has been removed from service and to perform the following evaluations.

The need to begin sampling Reactor Water The need to remove the Durability Monitor from service

[8] IF the isolation cannot be reset, THEN PERFORM the following:

[8.1] ISOLATE seal water from the CRD System by closing 2-69-592(A) and 2-69-614(B) (R-12T, El. 593, Unit 2 Reactor Building).

[8.2] REFER TO 2-OI-68, Recirculation System, for Recirc System operating restrictions while RWCU is isolated.

If contacted as Radiation Protection or Chemistry acknowledge any directions or reports given.

Driver If contacted as Unit 1 to check Reactor Building, Turbine Building, and Refuel Zone Exhaust Radiation on 0-MON-90-361, CHEMISTRY CAM MONITOR CONTROLLER (Panel 1-9-2),

acknowledge the direction.

Page 19 of 57 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 4 Page 6 of 12 Event

Description:

Reactor Water Cleanup (RWCU) Leak / One PCIV Fails to Close Time Position Applicants Actions or Behavior

[9] EVALUATE Technical Requirements Manual 3.4.1, Coolant NUSO Chemistry, for limiting conditions for operation (Required Action).

Technical Requirements Manual 3.4.1, Coolant Chemistry.

NUSO TSR 3.4.1.1 is applicable SURVEILLANCE: FREQUENCY:

Monitor Reactor Coolant 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months when the continuous NUSO conductivity conductivity monitor is INOPERABLE and the Reactor is not in MODE 4 or 5 If contacted as Chemistry stating that Reactor Coolant sampling Driver is required every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months in accordance with TSR 3.4.1.1, acknowledge report.

Page 20 of 57 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 4 Page 7 of 12 Event

Description:

Reactor Water Cleanup (RWCU) Leak / One PCIV Fails to Close Time Position Applicants Actions or Behavior Technical Specification 3.6.1.3, Primary Containment Isolation Valves (PCIVs)

LCO 3.6.1.3 Each PCIV, except Reactor Building-to-Suppression Chamber Vacuum Breakers, shall be OPERABLE APPLICABILITY: MODES 1, 2, and 3 When associated instrumentation is required to be OPERABLE per LCO 3.3.6.1, "Primary Containment Isolation Instrumentation."

NOTES

1. Penetration flow paths except for 18 and 20 inch purge valve penetration flow paths may be un-isolated intermittently under administrative controls.

2. Separate Condition entry is allowed for each penetration flow NUSO path.

3. Enter applicable Conditions and Required Actions for systems made inoperable by PCIVs.

4. Enter applicable Conditions and Required Actions of LCO 3.6.1.1, "Primary Containment," when PCIV leakage results in exceeding overall containment leakage rate acceptance criteria in MODES 1, 2, and 3.

CONDITION:

NOTE: Only applicable to penetration flow paths with two PCIVs.

A. - One or more penetration flow paths with one PCIV inoperable except due to MSIV leakage not within limits.

Page 21 of 57 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 4 Page 8 of 12 Event

Description:

Reactor Water Cleanup (RWCU) Leak / One PCIV Fails to Close Time Position Applicants Actions or Behavior REQUIRED ACTION: COMPLETION TIME:

A.1 - Isolate the affected penetration A.1 - 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months except for Main flow path by use of at least one Steam Line closed and de-activated automatic valve, closed manual valve, blind flange, or check valve with flow through the valve secured AND NOTE: Isolation devices in High Radiation Areas may be verified by use of administrative means.

NUSO A.2 - Verify the affected penetration A.2 - Once per 31 days for flow path is isolated isolation devices outside Primary Containment AND Prior to entering MODE 2 or 3 from MODE 4, if Primary Containment was de-inerted while in MODE 4, if not performed within the previous 92 days, for isolation devices inside Primary Containment If RWCU Room Temperature exceeds the Maximum Normal 2-EOI-3, Secondary Containment Control NUSO Page 22 of 57 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 4 Page 9 of 12 Event

Description:

Reactor Water Cleanup (RWCU) Leak / One PCIV Fails to Close Time Position Applicants Actions or Behavior 2-EOI-3, Secondary Containment Control SC-1 IF THEN Reactor Zone Ventilation Exhaust Radiation level is above NO ACTION REQUIRED 72 mR/hr Refuel Zone Ventilation Exhaust Radiation level is above 72 NO ACTION REQUIRED mR/hr Reactor Zone Ventilation is NUSO isolated AND NO ACTION REQUIRED Reactor Zone Ventilation Exhaust Radiation level is below 72 mR/hr Refuel Zone Ventilation is isolated AND NO ACTION REQUIRED Refuel Zone Ventilation Exhaust Radiation level is below 72 mR/hr Page 23 of 57 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 4 Page 10 of 12 Event

Description:

Reactor Water Cleanup (RWCU) Leak / One PCIV Fails to Close Time Position Applicants Actions or Behavior NUSO Page 24 of 57 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 4 Page 11 of 12 Event

Description:

Reactor Water Cleanup (RWCU) Leak / One PCIV Fails to Close Time Position Applicants Actions or Behavior SC/T-1 IF Reactor Zone or Refuel Zone Ventilation Exhaust Radiation Level is below 72 mr/hr THEN operate available Reactor Zone or Refuel Zone Ventilation NUSO SC-3 ISOLATE all systems that are discharging into the area EXCEPT systems required:

For damage control OR To be operated by EOIs NOTE Tables SC-1 and SC-2 contain information that may be used to determine if a primary system is discharging into Secondary Containment (Emergency Depressurization will reduce discharge).

When the RWCU Leak has been isolated and Area Temperature and Radiation is below the Maximum Safe value, the NUSO may NRC contact the Shift Manager and recommend exiting 2-EOI-3, Secondary Containment Control, as an emergency no longer exists.

Page 25 of 57 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 4 Page 12 of 12 Event

Description:

Reactor Water Cleanup (RWCU) Leak / One PCIV Fails to Close Time Position Applicants Actions or Behavior If contacted as the Shift Manager by the NUSO to discuss exiting Driver 2-EOI-3, Secondary Containment Control, agree with any recommendation given.

End of Event 4. Request that the Driver insert Event 5, Core NRC Spray Loop I Room Cooler EECW Leak.

Page 26 of 57 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 5 Page 1 of 2 Event

Description:

Core Spray Loop I Room Cooler EECW Leak Time Position Applicants Actions or Behavior When requested by the Chief Examiner, insert Event 5, Core Spray Loop I Room Cooler EECW Leak. Contact the NUSO as the Reactor Building Assistant Unit Operator (AUO) and report that you discovered and isolated a water leak in the Core Spray Loop 1 Room Cooler. Report that the following valves were Driver closed to isolate the leak:

2-SHV-67-550, NW Core Spray Room Cooler Supply Shutoff 2-SHV-67-553, NW Core Spray Room Cooler Outlet If asked, the water appears to have stopped leaking.

If contacted as Work Control or Mechanical Maintenance, Driver acknowledge any direction concerning the Core Spray Loop I Room Cooler.

Technical Requirements Manual 3.5.3, Equipment Area Coolers LCO 3.5.3 The Equipment Area Cooler associated with each RHR Pump and the Equipment Area Cooler associated with each set of Core Spray Pumps (A and C or B and D) must be OPERABLE at all times when the pump or pumps served by that specific cooler is NUSO considered to be OPERABLE.

APPLICABILITY: Whenever the associated subsystem is required to be OPERABLE CONDITION:

A. - One or more Equipment Area Cooler inoperable REQUIRED ACTION: COMPLETION TIME:

A.1 - Declare the pump(s) served A.1 - Immediately NUSO by that cooler INOPERABLE (Refer to applicable Tech Spec and TRM LCOs)

Page 27 of 57 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 5 Page 2 of 2 Event

Description:

Core Spray Loop I Room Cooler EECW Leak Time Position Applicants Actions or Behavior Technical Specification 3.5.1, ECCS - Operating LCO 3.5.1 Each ECCS injection/spray subsystem and the Automatic Depressurization System (ADS) function of six safety/relief valves shall be OPERABLE APPLICABILITY: MODE 1, MODES 2 and 3, except High Pressure Coolant Injection (HPCI) and ADS valves are not required to NUSO be OPERABLE with Reactor Steam Dome Pressure 150 psig CONDITION:

A. - One low pressure ECCS injection/spray subsystem inoperable.

OR One low pressure coolant injection (LPCI) pump in both LPCI subsystems inoperable.

REQUIRED ACTION: COMPLETION TIME:

NUSO See Condition F See Condition F CONDITION:

F. - One ADS Valve inoperable NUSO AND Condition A entered REQUIRED ACTION: COMPLETION TIME:

F.1 - Restore ADS Valve to F.1 - 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months OPERABLE status NUSO OR F.2 - Restore Low Pressure ECCS F.2 - 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Injection / Spray subsystem to OPERABLE status End of Event 5. Request that the Driver insert Event 6, 2C 4KV NRC Unit Board Trip.

Page 28 of 57 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 6 Page 1 of 3 Event

Description:

2C 4KV Unit Board Trip Time Position Applicants Actions or Behavior When requested by the Chief Examiner, insert Event 6, 2C 4KV Driver Unit Board Trip.

Acknowledges and reports the following alarms:

4KV UNIT BOARD 2C UNDERVOLTAGE, 2-9-8B, Window 14 BOP CONDENSATE BOOSTER PUMP C AUX OIL PRESS LOW, 2-9-6A, Window 14 MOTOR TRIPOUT, 2-9-8C, Window 33 CREW Monitors Reactor Water Level.

Reports a loss of Control Rod Drive (CRD) System Flow due to 2A OATC CRD Pump being de-energized.

Directs the BOP to respond in accordance with applicable Alarm NUSO Response Procedures and direct the OATC to respond in accordance with 2-AOI-85-3, CRD System Failure.

2-AOI-85-3, CRD System Failure Immediate Actions

[1] IF operating CRD pump has failed AND standby CRD pump is available, THEN PERFORM the following at Panel 2-9-5: (Otherwise OATC N/A)

[1.1] PLACE 2-FIC-85-11, CRD SYSTEM FLOW CONTROL, in MAN at minimum setting.

[1.2] START associated standby CRD Pump using 2-HS-85-2A, CRD Pump 1B.

If contacted as Unit 1 concerning 1B CRD Pump being started for Unit 2, acknowledge any report given. If the crew requests to Driver use 1B CRD Pump for Unit 2, inform the crew that 1B CRD Pump is not needed for Unit 1.

[1.3] IF CRD Pump 1B was started, THEN OPEN 2-HS-85-8A, OATC CRD PUMP 1B DISCH TO U2.

Page 29 of 57 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 6 Page 2 of 3 Event

Description:

2C 4KV Unit Board Trip Time Position Applicants Actions or Behavior

[1.4] ADJUST 2-FIC-85-11, CRD SYSTEM FLOW CONTROL, to establish the following conditions:

2-PDI-85-18A, CRD COOLING WATER HEADER DP, between 10 psid and 20 psid OATC 2-FIC-85-11, CRD SYSTEM FLOW CONTROL, between 40 and 65 gpm

[1.5] BALANCE 2-FIC-85-11, CRD SYSTEM FLOW CONTROL, AND PLACE in AUTO or BALANCE.

2-ARP-9-8B, Alarm Response Procedure 4KV UNIT BOARD 2C UNDERVOLTAGE, Window 14 Operator Action:

A. CHECK Unit in stable condition by checking:

Condensate Pump C Condensate Booster Pump C RCW Pump C CCW Pump C CRD Pump 2A B. IF undervoltage has occurred, THEN BOP

1. CLEAR disagreement lights on breakers.

2. REDUCE load as necessary to maintain stable operating conditions.

3. Condenser discharge may need to be throttled for two CCW Pump operation. REFER TO 2-OI-27, Condenser Circulating Water System.

C. CHECK Unit Bd C for abnormal conditions: relay targets, smoke, burned paint, etc.

D. REFER TO 0-OI-57A, Switchyard and 4160V AC Electrical System, to re-energize board.

E. REFER TO appropriate OI for recovery or realignment of equipment.

Page 30 of 57 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 6 Page 3 of 3 Event

Description:

2C 4KV Unit Board Trip Time Position Applicants Actions or Behavior If contacted as an AUO, Work Control, or Electrical Maintenance to investigate, acknowledge the direction. If directed to prepare Driver protected equipment tags acknowledge the direction.

Wait 2 minutes and report that 2C 4KV Unit Board has an overcurrent trip flag.

2-ARP-9-8C, Alarm Response Procedure MOTOR TRIPOUT, Window 33 Operator Action:

A. CHECK Control Room for white disagreement light illuminated for affected equipment.

B. CLEAR disagreement light.

BOP C. DISPATCH personnel to check:

Relays at associated electrical board Equipment for abnormal conditions Safe-stop locally reset, if necessary D. REFER TO 0-GOI-300-2, Electrical, if relay targets are present or for motor starting limits.

E. REFER TO appropriate OI for recovery or realignment of equipment 2-ARP-9-6A, Alarm Response Procedure CONDENSATE BOOSTER PUMP C AUX OIL PUMP PRESS LOW, Window 14 BOP Operator Action:

A. DISPATCH personnel to check Booster Pump Lube Oil system:

1. ENSURE running or start Aux Oil Pump.

2. CHECK for leaks.

3. CHECK oil level and temperature at reservoir.

If contacted as the Turbine Building AUO to start 2C Condensate Driver Booster Pump Aux Oil Pump, insert Event 16 and report that the Aux Oil Pump is running.

End of Event 6. Request that the Driver insert Event 7 NRC Un-isolable Reactor Core Isolation Cooling (RCIC) Steam Leak.

Page 31 of 57 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 7 Page 1 of 11 Event

Description:

Un-isolable Reactor Core Isolation Cooling (RCIC) Steam Leak Time Position Applicants Actions or Behavior When requested by the Chief Examiner, insert Event 7, Driver Un-isolable Reactor Core Isolation Cooling (RCIC) Steam Leak.

Event 8, Fuel Damage, and Event 9, Reactor Feedwater Pumps Trip / HPCI Fails to Automatically Start and Inject, are NRC automatically entered by the simulator setup. No action is required by the Driver to Insert Event 8 or Event 9. See page 43 of 57 for Event 8 actions and page 50 of 57 for Event 9 actions.

Acknowledges and reports the following alarms as they are received:

REACTOR BUILDING RADIATION HIGH, 2-9-3A, Window 22 BOP REACTOR BUILDING, TURBINE BUILDING, REFUEL ZONE EXHAUST RADIATION HIGH, 2-9-3A, Window 4 RCIC STEAM LINE LEAK DETECTION TEMPERATURE HIGH, 2-9-3D, Window 10 NRC See Event 8 (page 43 of 57) for actions for Radiation Alarms.

Directs the BOP to respond in accordance with the applicable Alarm NUSO Response Procedure.

2-9-ARP-3D, Alarm Response Procedure RCIC STEAM LINE LEAK DETECTION TEMPERATURE HIGH, Window 10 BOP A. CHECK RCIC Temperature elements on 2-TR-69-29, LEAK DETECTION SYSTEM TEMPERATURE recorder (points 9-12) on Panel 2-9-22.

BOP Checks Area Temperatures on Panel 2-9-22.

B. IF RCIC is NOT in service AND 2-FI-71-1A(B), RCIC STEAM BOP FLOW indicates flow, THEN ISOLATE RCIC AND VERIFY temperatures lowering.

Determines that RCIC failed to automatically isolate, and attempts to BOP manually isolate RCIC. Informs the NUSO that RCIC will not isolate.

C. IF high temperature is confirmed, THEN ENTER 2-EOI-3, BOP Secondary Containment Control.

Page 32 of 57 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 7 Page 2 of 11 Event

Description:

Un-isolable Reactor Core Isolation Cooling (RCIC) Steam Leak Time Position Applicants Actions or Behavior BOP Confirms that Area Temperatures are rising and informs the NUSO.

Enters 2-EOI-3, Secondary Containment Control. Directs the BOP to NUSO monitor Secondary Containment parameters.

D. CHECK 2-RI-90-26A, CS/RCIC ROOM EL 519 RX BLDG radiation indicator on Panel 2-9-11 and NOTIFY RADCON if rising radiation BOP levels are observed.

E. DISPATCH personnel to investigate.

If contacted as Radiation Protection that radiation levels are rising, acknowledge the report.

Driver If contacted as the Reactor Building AUO to investigate, acknowledge the direction.

F. REFER TO Tech Specs 3.3.6.1, Primary Containment Isolation NUSO Instrumentation and 3.5.3, RCIC System.

Technical Specification evaluation for this event is not required NRC and should not be used to evaluate the candidates Technical Specification competency.

G. EVALUATE equipment associated with this alarm to determine compensatory actions required to maintain REP function. REFER TO NPG-SPP-18.3.5, Equipment Important to Emergency Response.

It is not expected that the SRO reference NPG-SPP-18.3.5, NRC Equipment Important to Emergency Response, during this event.

Page 33 of 57 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 7 Page 3 of 11 Event

Description:

Un-isolable Reactor Core Isolation Cooling (RCIC) Steam Leak Time Position Applicants Actions or Behavior 2-EOI-3, Secondary Containment Control SC-1 IF THEN Reactor Zone Ventilation Exhaust Radiation level is above NO ACTION REQUIRED 72 mR/hr Refuel Zone Ventilation Exhaust Radiation level is above 72 NO ACTION REQUIRED mR/hr Reactor Zone Ventilation is isolated AND NO ACTION REQUIRED NUSO Reactor Zone Ventilation Exhaust Radiation level is below 72 mR/hr Refuel Zone Ventilation is isolated AND NO ACTION REQUIRED Refuel Zone Ventilation Exhaust Radiation level is below 72 mR/hr Page 34 of 57 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 7 Page 4 of 11 Event

Description:

Un-isolable Reactor Core Isolation Cooling (RCIC) Steam Leak Time Position Applicants Actions or Behavior NUSO Page 35 of 57 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 7 Page 5 of 11 Event

Description:

Un-isolable Reactor Core Isolation Cooling (RCIC) Steam Leak Time Position Applicants Actions or Behavior SC/T-1 IF Reactor Zone or Refuel Zone Ventilation Exhaust Radiation level is below 72 mR/hr THEN Operate available Reactor Zone or Refuel Zone Ventilation SC/T-2 WHEN ANY Area Temperature exceeds its Max Normal Temperature (Table SC-1)

NUSO Page 36 of 57 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 7 Page 6 of 11 Event

Description:

Un-isolable Reactor Core Isolation Cooling (RCIC) Steam Leak Time Position Applicants Actions or Behavior SC-3 ISOLATE all systems that are discharging into the area EXCEPT systems required:

For damage control OR To be operated by EOIs NOTE Tables SC-1 and SC-2 contain information that may be used to determine if a primary system is discharging into Secondary Containment (Emergency Depressurization NUSO will reduce discharge).

SC-4 SC-7 WHEN A Primary System is discharging into Secondary Containment Page 37 of 57 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 7 Page 7 of 11 Event

Description:

Un-isolable Reactor Core Isolation Cooling (RCIC) Steam Leak Time Position Applicants Actions or Behavior SC-8 BEFORE ANY Secondary Containment parameter reaches its Max Safe Value (Tables SC-1, SC-2, and SC-3)

NUSO Critical Task:

With the Reactor at power and with a Primary System discharging into the Secondary Containment, manually SCRAM the Reactor before any area exceeds the Maximum Safe Temperature operating value by entering EOI-1, RPV Control.

CREW Critical Task Failure Criteria:

The operating crew fails to proceed without delay and in a controlled manner to initiate a Reactor SCRAM from the time it is announced that one Area Temperature is approaching the Maximum Safe value.

Enters 2-EOI-1, RPV Control. Directs the crew to enter 2-AOI-100-1, NUSO Reactor SCRAM, and directs the OATC to insert a manual Reactor SCRAM.

Event 8, Fuel Damage, and Event 9, Reactor Feedwater Pumps NRC Trip / HPCI Fails to Automatically Initiate, are inserted when the Reactor MODE SWITCH is placed in SHUTDOWN.

OATC Inserts a manual Reactor SCRAM.

Page 38 of 57 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 7 Page 8 of 11 Event

Description:

Un-isolable Reactor Core Isolation Cooling (RCIC) Steam Leak Time Position Applicants Actions or Behavior 2-AOI-100-1, Reactor SCRAM Immediate Actions

[1] DEPRESS 2-HS-99-5A/S3A, REACTOR SCRAM A and 2-HS-99-5A/S3B, REACTOR SCRAM B, on Panel 2-9-5.

[2] PLACE 2-HS-99-5A/S1, REACTOR MODE SWITCH, in SHUTDOWN.

[3] N/A

[4] IF Reactor Power is 5% or BELOW, THEN: (otherwise MARK N/A)

OATC REPORT the following to the US:

Reactor Scram Mode Switch is in Shutdown All rods in or rods out Reactor Water Level and trend (recovering or lowering)

Reactor Pressure and trend MSIV position (Open or Closed)

Power level

[5] N/A Determines that all Reactor Feedwater Pumps (RFPTs) have tripped OATC and informs the NUSO (See Event 9).

2-EOI-1, RPV Control NUSO RC/L-1 ENSURE each as required:

PCIS isolations (Groups 1, 2, and 3)

ECCS RCIC Page 39 of 57 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 7 Page 9 of 11 Event

Description:

Un-isolable Reactor Core Isolation Cooling (RCIC) Steam Leak Time Position Applicants Actions or Behavior RC/L-2 IF THEN RPV Water Level can be restored and maintained above

(-)162 in. INHIBIT ADS AND The ADS timer has initiated Loss of available injection systems is anticipated OR Raising RPV Water Level above

(+)51 in. will facilitate use of NO ACTION REQUIRED shutdown cooling, steam-driven injection systems, or Alternate NUSO Injection Subsystems (Table L-2)

RC/L-3 RESTORE and MAINTAIN RPV Water Level between (+)2 in. and

(+)51 in. with ANY Preferred Injection Systems (Table L-1)

IF THEN RPV Water Level cannot be restored and maintained NO ACTION REQUIRED between (+)2 in. and (+)51 in.

RPV Water Level cannot be restored and maintained above NO ACTION REQUIRED

(-)162 in.

Page 40 of 57 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 7 Page 10 of 11 Event

Description:

Un-isolable Reactor Core Isolation Cooling (RCIC) Steam Leak Time Position Applicants Actions or Behavior Directs the OATC/BOP to maintain Reactor Water Level using HPCI NUSO in accordance with 2-EOI-Appendix-5D, Injection System Lineup HPCI.

2-EOI-Appendix-5D, Injection System Lineup HPCI actions are NRC covered in Event 9. See page 50 of 57.

NUSO RC/P-1 IF THEN A high Drywell Pressure ECCS NO ACTION REQUIRED signal exists (2.45 psig)

EMERGENCY RPV DEPRESSURIZATION is C2 Emergency RPV NUSO Depressurization REQUIRED or has been required Emergency RPV NO ACTION REQUIRED Depressurization is anticipated RC-P/2 IF ANY MSRV is cycling THEN NO ACTION REQUIRED Page 41 of 57 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 7 Page 11 of 11 Event

Description:

Un-isolable Reactor Core Isolation Cooling (RCIC) Steam Leak Time Position Applicants Actions or Behavior RC/P-3 IF THEN Suppression Pool Temperature and Level CANNOT be maintained in a safe area of NO ACTION REQUIRED Curve 3 at the existing RPV Pressure Suppression Pool Level CANNOT be maintained in the NO ACTION REQUIRED safe area of Curve 4 NO ACTION REQUIRED NUSO RC/P-4 STABILIZE RPV Pressure below 1073 psig using the Main Turbine Bypass Valves (APPX 8B)

OK to use ANY Alternate RPV Pressure Control Systems (Table P-1)

Crosstie CAD or MSRV carts to DW Control Air (APPX 8G, 20H) if necessary IF THEN DW Control Air is or becomes NO ACTION REQUIRED unavailable Directs the BOP to maintain Reactor Pressure using the Main Turbine NUSO Bypass Valves.

NUSO NRC End of Event 7. Continue to Event 8.

Page 42 of 57 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 8 Page 1 of 7 Event

Description:

Fuel Damage Time Position Applicants Actions or Behavior Event 8, Fuel Damage, is inserted when the Reactor MODE NRC SWITCH is placed in SHUTDOWN. No action is required by the driver to insert Event 8.

Acknowledges and reports the following alarms to the NUSO as they are received:

BOP REACTOR BUILDING RADIATION HIGH, 2-9-3A, Window 22 REACTOR BUILDING, TURBINE BUILDING, REFUEL ZONE EXHAUST RADIATION HIGH, 2-9-3A, Window 4 2-ARP-9-3A, Alarm Response Procedure REACTOR BUILDING RADIATION HIGH, Window 22 BOP Operator Action:

A. DETERMINE area with high radiation level on Panel 2-9-11. (Alarm on Panel 2-9-11 will automatically reset if radiation level lowers below setpoint.)

Monitors Radiation Levels on Panel 2-9-11. Keeps the NUSO BOP informed on instruments which are alarming and which are approaching Maximum Safe Values.

B. N/A BOP C. N/A D. NOTIFY Radiation Protection.

If contacted as Radiation Protection, acknowledge any reports or Driver direction given.

E. IF the TSC is NOT manned and a VALID radiological condition BOP exists, THEN USE public address system to evacuate area where high radiological conditions exist.

Makes a plant announcement to evacuate the Reactor Building due to BOP high radiation.

Page 43 of 57 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 8 Page 2 of 7 Event

Description:

Fuel Damage Time Position Applicants Actions or Behavior F. N/A G. MONITOR other parameters providing input to this annunciator frequently as these parameters will be masked from alarming while this alarm is sealed in.

H. IF a CREV initiation is received, THEN

1. CHECK CREV A(B) Flow is 2700 CFM, and 3300 CFM as indicated on 0-FI-031-7214(7213) within 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months of the CREV BOP initiation.

2. IF CREV A(B) Flow is NOT 2700 CFM, and 3300 CFM as indicated on 0-FI-031-7214(7213), THEN PERFORM the following: (Otherwise N/A)

a. STOP the operating CREV per 0-OI-31, Control Bay and Off-Gas Treatment Building Air Conditioning System.

b. START the standby CREV per 0-OI-31, Control Bay and Off-Gas Treatment Building Air Conditioning System.

If contacted as an AUO to monitor CREV operation, acknowledge Driver the direction.

I. N/A BOP J. For all radiation indicators except FUEL STORAGE POOL radiation indicator, 2-RI-90-30, ENTER 2-EOI-3, Secondary Containment Control Flowchart.

Re-enters 2-EOI-3, Secondary Containment Control (if not already NUSO entered on Secondary Containment Radiation).

K. N/A BOP L. EVALUATE equipment associated with this alarm to determine compensatory actions required to maintain REP function. REFER TO NPG-SPP-18.3.5, Equipment Important to Emergency Response.

It is not expected that the SRO reference NPG-SPP-18.3.5, NRC Equipment Important to Emergency Response, during this event.

Page 44 of 57 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 8 Page 3 of 7 Event

Description:

Fuel Damage Time Position Applicants Actions or Behavior SC/R-1 WHEN ANY Area Radiation Level exceeds its Max Normal Radiation Level (Table SC-2)

NUSO Page 45 of 57 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 8 Page 4 of 7 Event

Description:

Fuel Damage Time Position Applicants Actions or Behavior SC-3 ISOLATE all systems that are discharging into the area EXCEPT systems required:

For damage control OR To be operated by EOIs NOTE Tables SC-1 and SC-2 contain information that may be used to determine if a primary system is discharging into Secondary Containment (Emergency Depressurization NUSO will reduce discharge).

SC-4 SC-7 WHEN A Primary System is discharging into Secondary Containment Page 46 of 57 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 8 Page 5 of 7 Event

Description:

Fuel Damage Time Position Applicants Actions or Behavior SC-8 BEFORE ANY Secondary Containment parameter reaches its Max Safe Value (Tables SC-1, SC-2, and SC-3)

NUSO Monitors Area Radiation levels and informs the NUSO when two BOP areas are at Maximum Safe.

SC-9 WHEN Any Secondary Containment parameter exceeds its Max Safe value in two (2) or more areas for the same parameter (Tables SC-1, NUSO SC-2, SC-3)

SC-10 EMERGENCY DEPRESSURIZATION IS REQUIRED Updates the crew that Emergency Depressurization is required.

NUSO Enters 2-C-2, Emergency RPV Depressurization.

Page 47 of 57 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 8 Page 6 of 7 Event

Description:

Fuel Damage Time Position Applicants Actions or Behavior Critical Task:

With a Primary System discharging into the Secondary Containment when two or more areas are greater than their maximum safe operating values for the same parameter, the Balance of Plant Operator initiates Emergency Depressurization Crew as directed by the Nuclear Unit Senior Operator.

Critical Task Failure Criteria:

The operating crew fails to proceed with without delay and in a controlled manner to initiate Emergency Depressurization when two Area Radiation Levels exceed Maximum Safe value.

2-C-2, Emergency RPV Depressurization C2-1 IF THEN Reactor Water Level CANNOT NO ACTION REQUIRED be determined NUSO It is anticipated that Reactor depressurization will result in NO ACTION REQUIRED loss of injection required for Adequate Core Cooling Containment Water Level CANNOT be maintained below NO ACTION REQUIRED 44 feet C2-2 IF Drywell Pressure is above 2.45 PSIG NUSO THEN PREVENT injection from ONLY those Core Spray and LPCI pumps NOT required to assure Adequate Core Cooling (Appendix 4)

Page 48 of 57 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 8 Page 7 of 7 Event

Description:

Fuel Damage Time Position Applicants Actions or Behavior C2-3 EMERGENCY DEPRESSURIZE the Reactor IF Suppression Pool Water Level is above 5.5 feet THEN OPEN 6 MSRVs (ADS Valves preferred)

OK to exceed 100ºF/hr cooldown rate IF THEN NUSO Drywell Control Air is or NO ACTION REQUIRED becomes unavailable Less than 4 MSRVs can be opened AND NO ACTION REQUIRED Reactor Pressure is 80 PSIG or more above Suppression Chamber Pressure Opens 5 SRVs and one additional SRV (due to ADS Valve 1-22 being BOP out of service).

C2-4 WHEN NUSO Shutdown Cooling RPV Pressure interlock clears AND further cooldown is required End of Event 8. When the crew has Emergency Depressurized the Reactor and has control of Reactor Water Level above the Top of NRC Active Fuel ((-) 162 inches) using low pressure systems, end of Scenario.

Page 49 of 57 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 9 Page 1 of 3 Event

Description:

Reactor Feedwater Pumps (RFPTs) Trip / HPCI Fails to Automatically Start and Inject Time Position Applicants Actions or Behavior Event 9, Reactor Feedwater Pumps Trip / HPCI Fails to Automatically Start and Inject, is inserted when the Reactor NRC MODE SWITCH is placed in SHUTDOWN. No action is required by the Driver to insert Event 9.

OATC Reports that all Reactor Feedwater Pumps (RFPTs) have tripped.

When Reactor Water Level reaches the High Pressure Coolant Injection (HPCI) initiation setpoint (-45), determines that HPCI did not BOP automatically start and manually starts HPCI. Informs the NUSO of the actions required to start HPCI.

Directs the BOP to maintain Reactor Water Level using NUSO 2-EOI-Appendix-5D, Injection System Lineup HPCI.

2-EOI-Appendix-5D, Injection System Lineup HPCI

[1] IF Suppression Pool Level drops below 12.75 ft. during HPCI operation, THEN TRIP HPCI and CONTROL injection using other options.

[2] IF Suppression Pool level CANNOT be maintained below 5.25 in, THEN EXECUTE EOI Appendix 16E, Bypassing HPCI High Suppression Pool Water Level Suction Transfer Interlock, concurrently with this procedure to bypass HPCI High Suppression Pool Water Level Suction Transfer Interlock.

BOP [3] IF BOTH of the following exist:

High temperature exists in the HPCI area, AND SRO directs bypass of HPCI High Temperature Isolation Interlocks, THEN PERFORM the following:

[3.1] EXECUTE EOI Appendix 16L, Bypassing HPCI High Temperature Isolation, concurrently with this procedure.

[3.2] RESET auto isolation logic using HPCI AUTO-ISOL LOGIC A (B) RESET pushbuttons.

[4] VERIFY at least one SGTS train in operation.

Page 50 of 57 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 9 Page 2 of 3 Event

Description:

Reactor Feedwater Pumps (RFPTs) Trip / HPCI Fails to Automatically Start and Inject Time Position Applicants Actions or Behavior CAUTIONS

1) Operating HPCI Turbine below 2400 rpm may result in unstable system operation and equipment damage.

2) Operating HPCI Turbine with Suction Temperatures above 140 F may result in equipment damage.

[5] VERIFY 2-FIC-73-33, HPCI SYSTEM FLOW/CONTROL, is in one of the following configurations, as desired:

in AUTO and set for 5300 gpm for rapid injection in AUTO and set for approximately 2500 gpm for slower injection in MANUAL with output at approximately 50% for slower injection NOTE BOP HPCI Auxiliary Oil Pump will NOT start UNTIL 2-FCV-73-16, HPCI TURBINE STEAM SUPPLY VLV, starts to open.

[6] IF high Reactor Water Level trip logic is actuated, THEN

[6.1] DEPRESS HPCI TURBINE TRIP RX LVL HIGH RESET pushbutton.

[6.2] CHECK HPCI TURBINE TRIP LVL HIGH amber light has extinguished.

[7] PLACE 2-HS-73-47A, HPCI AUXILIARY OIL PUMP handswitch in START.

[8] PLACE 2-HS-73-10A, HPCI STEAM PACKING EXHAUSTER handswitch in START.

[9] OPEN 2-FCV-73-30, HPCI PUMP MIN FLOW VALVE.

[10] OPEN 2-FCV-73-44, HPCI PUMP INJECTION VALVE.

[11] OPEN 2-FCV-73-16, HPCI TURBINE STEAM SUPPLY VALVE, to start the HPCI Turbine.

Page 51 of 57 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 9 Page 3 of 3 Event

Description:

Reactor Feedwater Pumps (RFPTs) Trip / HPCI Fails to Automatically Start and Inject Time Position Applicants Actions or Behavior

[12] CHECK proper HPCI operation by observing the following:

A. HPCI Turbine speed accelerates.

B. 2-CKV-73-45, HPCI SYSTEM CHECK VLV, opens by observing 2-ZI-73-45A, DISC POSITION, red light illuminated.

C. HPCI flow to RPV stabilizes and is controlled automatically at the setpoint. (N/A if controller in manual).

D. 2-FCV-73-30, HPCI PUMP MIN FLOW VALVE, closes as flow exceeds approximately 1200 gpm.

CAUTION HPCI PUMP MIN FLOW VALVE, 2-FCV-073-0030, automatically BOP opens when system flow is at or below 900 gpm (lowering) only if a system initiation signal is present. Manually opening the Min Flow Valve may be required for pump min flow protection.

[13] ADJUST 2-FIC-73-33, HPCI SYSTEM FLOW/CONTROL, controller as necessary to control injection.

[14] VERIFY HPCI Auxiliary Oil Pump stops and the shaft-driven oil pump operates properly.

[15] WHEN HPCI Auxiliary Oil Pump stops, THEN PLACE HPCI AUXILIARY OIL PUMP handswitch in AUTO.

[16] N/A

[17] N/A End of Event 9. When the crew has Emergency Depressurized the Reactor and has control of Reactor Water Level above the NRC Top of Active Fuel ((-) 162 inches) using low pressure systems, end of Scenario.

Page 52 of 57 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Scenario Setup UNIT 2 IC 54 Exam IC 276 Procedure Revision Procedure Revision Procedure Revision OI-68 159 EOI-3 17 ARP 6A 34 OI-69 139 2-C-2 11 ARP 8B 17 GOI-12 48 APPX 5D 10 ARP 8C 19 GOI-12A 118 ARP 3A 55 TRM 3.4.1 21 AOI-64-2A 27 ARP 3D 34 TS 3.5.1 A294 AOI-85-3 26 ARP 4B 51 TS 3.5.3 0 AOI-100-1 116 ARP 4C 35 TS 3.6.1.3 A253 EOI-1 18 ARP 5B 31 Verify camera system is powered down (admin password = abcd1234)

Simulator Setup Start CPERF PRIOR to placing the Simulator in RUN Ensure Danger Tags are placed on SRV 1-22 and the Emergency High Pressure Makeup Pump 2104 NRC Scenario 1 UNIT 2.sch Schedule Files(s):

RWCU.sch Event Files(s): 2104 NRC Scenario 1 UNIT 2.evt Schedule File - 2104 NRC Scenario 1 Unit 2.sch Event Action Description 2104 NRC Scenario 1 Unit 2.evt Event File 1 Insert remote CU01 to 45.00000 RWCU DEMIN FILTER A FRC-69-35 in 60 1 Insert remote CU02 to 45.00000 RWCU DEMIN FILTER B FRC-69-60 in 60 1 Insert remote AN01E after 15 to CU LOCAL RESET (2-XA-55-4B W24)

RESET 11 Insert remote CU05 to MANUAL RWCU HX RBCCW FLOW CONTROL TIC-69-10A 21 Insert remote CU05 to AUTO RWCU HX RBCCW FLOW CONTROL TIC-69-10A Page 53 of 57 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Schedule File Event Action Description Insert malfunction 3 RBCCW SURGE TANK LEVEL LOW 2-LA-70-2B XA-55-4C_13 to ON Insert malfunction 13 XA-55-4C_13 after 10 to RBCCW SURGE TANK LEVEL LOW 2-LA-70-2B NORMAL delete in 1 Schedule F:\ Scenarios\Scenario 4

1\RWCU.sch 4KV UNIT BOARD 2C FAILURE (RELAY 86-316 AND 6 Insert malfunction ED08C 86-532) 16 Insert remote FW06 to START CONDENSATE BOOSTER PUMP C AUX OIL PUMP Insert remote AN01D to RESET 16 FW LOCAL RESET 121 (2-XA-55-6B W6) in 5 Insert malfunction RC09 to 7 RCIC STEAM LEAK INTO RCIC ROOM 35.00000 in 1800 Insert malfunction FCV-71-2 to MOTOR_OPERATED_VALVE RCIC STEAM LINE INBD FAIL_NOW ISOL VLV Insert override ZLOHS712A_1 to HS-71-2A-GREEN RCIC STEAM LINE INBD ISOLATION Off VLV Insert override ZLOHS712A_2 to HS-71-2A-Red* RCIC STEAM LINE INBD ISOLATION VLV On 17 Delete override ZLOHS712A_2 HS-71-2A-Red* RCIC STEAM LINE INBD ISOLATION VLV Insert malfunction FCV-71-3 to MOTOR_OPERATED_VALVE RCIC STEAM LINE OUTBD FAIL_NOW ISOL VLV Insert override ZLOHS713A_1 to HS-71-3A-GREEN RCIC STEAM LINE OUTBD ISOLATION Off VLV Insert override ZLOHS713A_2 to HS-71-3A-Red* RCIC STEAM LINE OUTBD ISOLATION On VLV Insert malfunction TH23 to 8 FUEL CLADDING DAMAGE 10.00000 in 900 RFPT 2A TRIP ON RFPT 2A BEARING LOW OIL 8 Insert malfunction FW14A PRESSURE (PS-3-123B)

Page 54 of 57 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Schedule File - 2104 NRC Scenario 1 Unit 2.sch Event Action Description RFPT 2B TRIP ON RFPT 2A BEARING LOW OIL 8 Insert malfunction FW14B PRESSURE (PS-3-149B)

RFPT 2C TRIP ON RFPT 2A BEARING LOW OIL 8 Insert malfunction FW14C PRESSURE (PS-3-174B)

Insert malfunction FCV-73-16 to MOTOR_OPERATED_VALVE HPCI TURBINE STEAM FAIL_NOW SUPPLY VLV Insert override ZLOHS7316A_1 HS-73-16A-Green* HPCI TURBINE STEAM SUPPLY VLV to On MOTOR_OPERATED_VALVE HPCI TURBINE STEAM Delete malfunction FCV-73-16 SUPPLY VLV Delete override ZLOHS7316A_1 HS-73-16A-Green* HPCI TURBINE STEAM SUPPLY VLV Insert malfunction RC09 to 27 RCIC STEAM LEAK INTO RCIC ROOM 18.00000 in 1800 on event 27 Insert malfunction RC09 to 8 RCIC STEAM LEAK INTO RCIC ROOM 100.00000 in 1800 on event 8 Schedule File - RWCU.sch Event Action Description Insert malfunction CU04 to RWCU SYSTEM SUCTION BREAK 25.00000 Insert malfunction FCV-69-2 to MOTOR_OPERATED_VALVE RWCU OUTBOARD FAIL_NOW ISOLATION VLV Insert override ZLOHS692A_1 to HS-69-2A-GREEN RWCU OUTBD SUCT ISOLATION Off VALVE Insert override ZLOHS692A_2 to HS-69-2A-Red* RWCU OUTBD SUCT ISOLATION VALVE On 14 MOTOR_OPERATED_VALVE RWCU OUTBOARD Delete malfunction FCV-69-2 ISOLATION VLV HS-69-2A-GREEN RWCU OUTBD SUCT ISOLATION 14 Delete override ZLOHS692A_1 VALVE 14 Delete override ZLOHS692A_2 HS-69-2A-Red* RWCU OUTBD SUCT ISOLATION VALVE Page 55 of 57 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Event File List Details Page 56 of 57 Unit 2

UNIT 2 SHIFT TURNOVER MEETING Today DAYS ON LINE Total Drywell Leakage Protected Equipment MODE 208 (gpm) None 1 PRA (EOOS) -GREEN 1.55 Rx Power 500Kv GRID - Qualified Floor Drain (gpm) 80% 161Kv Grid -Qualified 0.11 Equipment Drain MWe Last breaker closure (gpm) 4/10/19 4:31 1.44 Review logs Qualifications Review RCP/Rx Brief Review LCO/OWA Actions Walkdown Panels/Verify EOOS CR Reviews Complete Leadership and Team Effectiveness CHANGES IN LCOs SRV 1-22 is INOPERABLE (ADS Valve). Tech Spec 3.5.1, Condition E (Day 4 of 14 Day LCO)

LCOs OF 72 HOURS OR LESS SIGNIFICANT ITEMS DURING PREVIOUS SHIFT/RADIOLOGICAL CHANGES Reactor Shutdown. Maintain RFPTs, Condensate, and Condensate Booster Pumps running until Reactor Power is <70%

EHPM tagged for motor bearing inspection RWCU Non-Regenerative Heat Exchanger repairs completed, tags are clear, and RWCU has been filled and vented.

MAJOR EQUIPMENT CHANGES PLANNED FOR THIS SHIFT Return RWCU to service in accordance with 2-OI-69, Section 5.1, beginning at Step [4]. RPHP is in effect.

Continue the Reactor Shutdown in accordance with 2-GOI-100-12A, Section 5.3.1, Step [4].

Reduce Reactor Power to 75% using Core Flow, then wait for further guidance from RE.

OPERATOR WORK AROUNDS OWAs - 0 Burdens - 0 Challenges - 6 ODMIs/ACMPs ONEAs FIRE RISK SIGNIFICANT ITEMS OOS/FPLCO Actions Due FPRM Attachment A, FPLCO A.2.2.1, Unit 2 Emergency High Pressure Makeup Pump - Fire Watch established SCHEDULED ITEMS NOT COMPLETED Page 57 of 57 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 1 Page 1 of 8 Event

Description:

Return Reactor Water Cleanup (RWCU) to Operation Time Position Applicants Actions or Behavior PRIOR to placing the simulator in RUN, start CPERF to record Driver critical parameters.

If the crew does not start Event 1, Return Reactor Water Cleanup (RWCU) to Operation, request that the Driver contact the Nuclear NRC Unit Senior Operator (NUSO) as the Shift Manager and direct the crew to return RWCU to operation.

If requested by the Chief Examiner, contact the NUSO as the Shift Manager and direct the crew to return RWCU to operation.

Driver If contacted by the crew as the Reactor Building Assistant Unit Operator (AUO) acknowledge any direction given.

If Panel 3-9-4B, Window 17, RWCU NON-REGENERATIVE HX DISCH NRC TEMP HIGH, is received, see page 8 of 57 for Alarm Response Procedure actions.

NUSO Directs the Balance of Plant Operator (BOP) to return RWCU to service.

3-OI-69, Reactor Water Cleanup System Section 5.1 RWCU Pump Startup NOTES

1) All controls and indications are located on Panel 3-9-4 unless noted otherwise.

2) RWCU is required to be operated with the following restrictions with Reactor Pressure 50 psig (MODES 2 or 3) or any time the unit is in BOP MODE 4, MODE 5, or defueled:

One pump in operation, pump can be operated to its maximum flow capacity.

Two pumps operation, max flow limit of 100 gpm per pump (200 gpm total).

[4] ENSURE RESET the RWCU Group 3 Isolation using 3-HS-64-16A-S32, PCIS DIV I RESET and 3-HS-64-16A-S33, PCIS DIV II RESET at Panel 3-9-4.

Page 1 of 57 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 1 Page 2 of 8 Event

Description:

Return Reactor Water Cleanup (RWCU) to Operation Time Position Applicants Actions or Behavior

[5] CHECK the following on Panel 3-LPNL-925-0003, Unit 3 Reactor Building, Elevation.621:

[5.1] Demin 3A and/or 3B Holding Pumps are running BOP (3-HS-069-6015 and 3-HS-069-6005).

[5.2] Demin 3A and/or 3B Outlet Valves are closed (3-HS-069-0035 and/or 3-HS-069-0060).

If contacted as the Reactor Building AUO to perform Step [5],

Driver acknowledge the direction and report that the Holding Pumps are running with the Demineralizer Outlet Valves closed.

[6] N/A

[7] ENSURE 3-TIC-069-0010A, RWCU HEAT EXCHANGERS RBCCW BOP FLOW CONTROL is in MANUAL, and FULL OPEN demand is on 3-TCV-70-49, RWCU NON-REGENERATIVE HEAT EXCHANGER OUTLET TCV. (REFER TO Attachment 5)

When directed to place 3-TIC-069-0010A in manual, insert Event 11.

Driver Inform the crew that 3-TIC-069-0010A is in manual and is fully open.

[8] ENSURE CLOSED the following:

3-HC-69-15, RWCU BLOWDOWN PRESSURE CONTROL VALVE 3-HS-69-16A, RWCU BLOWDOWN TO MAIN CONDENSER BOP 3-HS-69-17A, RWCU BLOWDOWN TO RADWASTE

[9] ENSURE 3-HS-69-15A, DEFEAT/OPERATE (FOR 3-HC-69-15) in the DEFEAT position.

[10] N/A

[11] NOTIFY Chemistry that RWCU is being placed in service and to check the durability monitor.

Page 2 of 57 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 1 Page 3 of 8 Event

Description:

Return Reactor Water Cleanup (RWCU) to Operation Time Position Applicants Actions or Behavior When contacted as Chemistry acknowledge the direction to check Driver the durability monitor.

[12] ENSURE OPEN the following valves:

3-FCV-69-1, RWCU INBD SUCT ISOLATION VALVE 3-FCV-69-2, RWCU OUTBD SUCT ISOLATION VALVE 3-FCV-69-8, RWCU DEMIN BYPASS VALVE

[13] OPEN 3-FCV-069-0012, RWCU RETURN ISOLATION VALVE by one of the two methods described below.

PLACE 3-HS-69-12A in the OPEN position, THEN RETURN 3-HS-69-12A to the NORM position when intermediate position (red and green light) is indicated PLACE 3-HS-69-12A in the OPEN position, THEN RETURN 3-HS-69-12A to the NORM position when FULL OPEN BOP position (red light only) is indicated NOTES

1) Too high a flow on startup after isolation could cause 3-TIS-69-11 to actuate due to a high Non-Regenerative Heat Exchanger Outlet Temperature (3-XS-69-6, RWCU TEMP SELECT, Position 3, WATER TO RWCU DEMINS).

2) The RWCU Pump will trip on low flow at 56 gpm, after a 30 second time delay. Failure to immediately raise flow to greater than 56 gpm in the following steps will result in a pump trip.

14] PLACE seal purge in operation to pump(s) to be placed in service.

REFER TO Section 8.3.

Page 3 of 57 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 1 Page 4 of 8 Event

Description:

Return Reactor Water Cleanup (RWCU) to Operation Time Position Applicants Actions or Behavior When directed to place seal purge in service, acknowledge the Driver direction and inform the crew that seal purge has been placed in service in accordance with 3-OI-69, Section 8.3

[15] N/A

[16] START RWCU RECIRC PUMP 3A(3B) using control switch 3-HS-69-4A(4B)-A, AND RAISE flow, using RWCU RETURN ISOLATION VALVE, 3-HS-69-12A, to prevent low flow trip.

BOP [17] IF two pump operation is desired, THEN START the second RWCU RECIRC PUMP 3B(3A) using control switch 3-HS-69-4B(4A)-A, AND RAISE flow using RWCU RETURN ISOLATION VALVE, 3-HS-69-12A, to prevent low flow trip.

[18] IF the RWCU filter-demineralizers are to be placed in service, THEN REFER TO Section 6.2.

3-OI-69, Reactor Water Cleanup System Section 6.2, Placing Filter-Demineralizers in Service CAUTION When initially placing a filter-demineralizer into service, it is desirable BOP that most RWCU Discharge Flow be returned to the Main Condenser.

If the Reactor is pressurized, however, failure to follow temperature restrictions could result in thermal shocking the Regenerative Heat Exchanger.

[1] REVIEW Precautions and Limitations in Section 3.0.

[2] - [10] Performed in the Field by an AUO.

Page 4 of 57 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 1 Page 5 of 8 Event

Description:

Return Reactor Water Cleanup (RWCU) to Operation Time Position Applicants Actions or Behavior When contacted as the Reactor Building AUO to prepare to roll in RWCU Demineralizer, acknowledge the direction and report that you are standing by with Steps complete Steps [2] through [10] of 3-OI-69, Reactor Water Cleanup System, Section 6.2 are complete.

When directed to place filter-demineralizers in service, insert Driver Event 1 to perform AUO actions to place demineralizers in service, and inform the crew that Demin Flow is rising.

Demineralizers will roll in over a 1-minute time frame - when complete inform the crew that RWCU filter-demineralizers have been placed in service.

If requested, Demineralizer flows are rising and are normal.

NOTE RWCU is required to be operated with the following restrictions with Reactor Pressure 50 psig (MODES 2 or 3) or any time the unit is in MODE 4, MODE 5, or defueled:

One pump in operation, pump can be operated to its maximum BOP flow capacity.

Two pumps in operation, maximum flow limited to 100 gpm per pump (200 gpm total)

[11] PERFORM the following simultaneously:

CLOSE 3-HS-69-8A, RWCU DEMIN BYPASS VALVE on Panel 3-9-4 Verify that the crew is able to clear RWCU Demineralizer Alarm (Panel 3-9-4B, Window 24) - it will automatically reset on Event 1 Driver after 15 seconds. If the crew cannot get the alarm to reset, insert remote function AN01E to RESET again).

[12] RAISE flow through the Demin until the desired flow has been established.

BOP

[13] ENSURE DEMIN 3A(3B) HOLDING PUMP, 3-HS-069-6015(6005),

in the AUTO position.

Page 5 of 57 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 1 Page 6 of 8 Event

Description:

Return Reactor Water Cleanup (RWCU) to Operation Time Position Applicants Actions or Behavior

[14] CHECK that Holding Pump 3A(3B), on the Demin being placed in service, has STOPPED.

BOP

[15] CHECK DEMIN 3A(3B) HOLDING PUMP DISCH VALVE H, 3-HS-069-0035B(0060B), has CLOSED.

When directed to perform Steps [12], [13], and [14] acknowledge Driver the direction and inform the crew that Steps [12], [13], and [14] are complete.

[16] NOTIFY Chemistry that the filter-demineralizer has been placed in BOP service and REQUEST a sample for conductivity and silica of the effluent.

When contacted as Chemistry, acknowledge any information or Driver direction given.

Continuing 3-OI-69, Reactor Water Cleanup System Section 5.1 RWCU Pump Startup

[19] ADJUST the RWCU RETURN ISOLATION VALVE, using 3-HS-69-12A, as required to obtain desired system flow.

CAUTIONS

1) Failure to maintain RWCU Non-Regenerative Heat Exchanger Tube Side Outlet Temperature below 130 F will reduce resin BOP efficiency and may result in resin damage.

2) Exercise care when making RWCU System Flow adjustments to values greater than 270 gpm to ensure temperature limits are not exceeded.

[20] THROTTLE blowdown flow as required to maintain the following parameters. (REFER TO Section 6.5).

Desired Reactor Water Level Non-Regenerative Heat Exchanger Tube Outlet Temperature less than 130 F Page 6 of 57 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 1 Page 7 of 8 Event

Description:

Return Reactor Water Cleanup (RWCU) to Operation Time Position Applicants Actions or Behavior

[21] IF at Operations Management discretion it is desired to place 3-TIC-069-0010A, RWCU HEAT EXCHANGERS RBCCW FLOW BOP CONTROL in AUTO, THEN PERFORM the following: (Otherwise N/A)

[21.1] PLACE 3-TIC-069-0010A, RWCU HEAT EXCHANGERS RBCCW FLOW CONTROL in AUTO. (REFER TO Section 8.15.)

If contacted by the crew to place 3-TIC-069-0010A in automatic Driver acknowledge the direction and inform the crew that 3-TIC-069-0010A has been placed in automatic.

NOTES

1) Seal water to the RWCU Pumps has been observed to slightly lower after pump(s) are placed in service.

2) When the Reactor Vessel is at atmospheric pressure and RWCU Pump seal water is being supplied by CS&S system, RWCU Pump seal water flow may decrease to 0 gpm after the RWCU Pump has BOP started. See PRECAUTION P&L 3.6E.

[22] ENSURE PURGE (SEAL) WATER TO RWCU PUMPS at Panel 3-25-314 (1.8 to 2.0 gpm). (REFER TO Section 8.3.)

3-FI-085-0075, RWCU PUMP 3A PURGE WATER FLOW INDICATOR 3-FI-085-0077, RWCU PUMP 3B SEAL WATER When contacted as the Reactor Building AUO to perform Step [22],

Driver inform the crew that seal water flow is 1.9 gpm.

The restoration of RWCU ICS point to the heat balance is not NRC required if automatically bypassed.

Page 7 of 57 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 1 Page 8 of 8 Event

Description:

Return Reactor Water Cleanup (RWCU) to Operation Time Position Applicants Actions or Behavior 3-9-ARP-4B, Alarm Response Procedure RWCU NON-REGENERATIVE HX DISCHARGE TEMPERATURE HIGH, Window 17 Operator Action:

A. CHECK RWCU NRHX Discharge Temperature, 3-XS-69-6, on Panel 3-9-4.

B. CHECK RBCCW System Temperature indication normal, Panel 3-9-4.

C. IF temperature continues to rise, THEN PERFORM the following, BOP otherwise, MARK steps N/A:

REDUCE system flow or reject flow as necessary to control temperature REFER TO 3-OI-69, Reactor Water Cleanup System D. DISPATCH personnel to CHECK the following:

RWCU Heat Exchangers RBCCW Flow Controller (normally in auto with setpoint at approx. 110 F), located on 3-LPNL-925-0002 Reactor Bldg 593 3-TCV-70-49 operating properly (RBCCW to NRHX), located in RWCU HX room If contacted by the crew to check equipment in Step D (see above),

acknowledge the direction and report the following as required:

Driver RWCU Heat Exchangers RBCCW Flow Controller is set at 110 F and is in automatic 3-TCV-70-49 is operating properly End of Event 1. Proceed to Event 2, Reduce Reactor Power to 75%

NRC using Core Flow.

Page 8 of 57 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 2 Page 1 of 5 Event

Description:

Reduce Reactor Power to 75% using Core Flow Time Position Applicants Actions or Behavior If the crew does not proceed to Event 2, Reduce Reactor Power to 75% using Core Flow, request that the Driver contact the NUSO as NRC the Shift Manager and direct the crew to reduce Reactor Power to 75%.

If requested by the Chief Examiner, contact the crew as the Shift Manager and direct the crew to continue with Step [2] of the Driver Reactivity Control Plan (RCP) for the Reactor Shutdown and reduce Reactor Power to 75% using Core Flow.

The crew may elect to conduct a re-focus reactivity brief prior to NRC lowering Reactor Power.

Directs the Operator at the Controls (OATC) to lower Reactor Power to 75% in accordance with Step [2] of RCP U3-2104NRC1 and in accordance with 3-OI-68, Reactor Recirculation System, NUSO 3-GOI-100-12A, Unit Shutdown from Power Operation to Cold Shutdown and Reductions in Power During Power Operations and 3-GOI-100-12, Power Maneuvering.

3-GOI-100-12A, Unit Shutdown from Power Operation to Cold Shutdown and Reductions in Power During Power Operations Section 5.3, Power Reduction OATC 5.3.1 Reducing Reactor Power to 40%

[1] ENSURE the operators are using Attachment 9, Operations Down Power Monitoring.

[2] REDUCE Reactor Power by a combination of Control Rod insertions and core flow changes, as recommended by Reactor Engineer.

Page 9 of 57 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 2 Page 2 of 5 Event

Description:

Reduce Reactor Power to 75% using Core Flow Time Position Applicants Actions or Behavior 3-GOI-100-12, Power Maneuvering Section 5.0, Instruction Steps OATC [7] REDUCE Reactor Power by a combination of Control Rod insertions and Core Flow changes, as recommended by Reactor Engineer.

REFER TO 3-SR-3.1.3.5(A) and 3-OI-68. (N/A if entering 3-GOI-100-12, Power Maneuvering, to recover from Recirc Pump Trip).

3-OI-68, Reactor Recirculation System 3.0 Precautions and Limitations Section 3.5.3, Dual Pump Operation E. When raising (lowering) Reactor Power per the Reactivity Plan, the following guideline should be used to establish the 60 rpm mismatch between the Recirc Pumps.

1. When the first Recirc Pump reaches 1200 rpm (1300 rpm) or both Recirc Pumps are at 1200 rpm (1300 rpm) then individual controls will be used.

NRC 2. While following the Reactivity Plan establish the 60 rpm mismatch using the individual controls for the leading Recirc Pump.

3. While maintaining the 60 rpm mismatch and using the Reactivity Plan, raise (lower) the Recirc Pump speeds using either the Master Controllers or Individual Controllers.

4. Once a Recirc Pump reaches 1300 rpm (1200 rpm) the 60 rpm mismatch is no longer required. While following the Reactivity Plan raise (lower) the lagging Recirc Pump using the individual controller match Recirc Pump speeds.

5. Once the Recirc Pumps are matched, then the speeds may be adjusted as required using the Master Controller or Individual Controllers while maintaining the requirements of Attachment 1.

Page 10 of 57 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 2 Page 3 of 5 Event

Description:

Reduce Reactor Power to 75% using Core Flow Time Position Applicants Actions or Behavior NRC Page 11 of 57 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 2 Page 4 of 5 Event

Description:

Reduce Reactor Power to 75% using Core Flow Time Position Applicants Actions or Behavior 3-OI-68, Reactor Recirculation System Section 6.2, Adjusting Recirc Flow NOTES

1) Thermal Limits are shown on 0-TI-248, Station Reactor Engineer and 3-SR-2, Instrument Checks and Observations.

2) Recirc Flow changes made during the later part of the operating cycle (Coastdown) could cause Core Flow values to approach or exceed the allowable values of the Increased Core Flow (ICF) Region of the Power to Flow Map. Instruments used to monitor pump speed and Core Flow should be identified in the Reactivity Control Plan.

These values should be recorded prior to reducing Core Flow and used as a benchmark to reestablish the previous conditions when returning to power. Increased caution should be used when changes in Recirc Flow are made in this area.

OATC [1] IF desired to control Recirc Pump 3A speed with Recirc Individual Control, THEN PERFORM the following; (Otherwise N/A)

[1.1] N/A

[1.2] Lower Recirc Pump 3A using 3-HS-96-17A(17B)(17C), SLOW (MEDIUM) (FAST). (Otherwise N/A)

[2] IF desired to control Recirc Pump 3B speed with Recirc Individual Control, THEN PERFORM the following; (Otherwise N/A)

[2.1] N/A

[2.2] Lower Recirc Pump 3B using 3-HS-96-18A(18B)(18C), SLOW (MEDIUM) (FAST). (Otherwise N/A)

[3] WHEN desired to control Recirc Pumps 3A and/or 3B speed with the RECIRC MASTER CONTROL, THEN ADJUST Recirc Pump speed 3A

& 3B using the following push buttons as required:

3-HS-96-33, LOWER SLOW 3-HS-96-34, LOWER MEDIUM 3-HS-96-35, LOWER FAST Page 12 of 57 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 2 Page 5 of 5 Event

Description:

Reduce Reactor Power to 75% using Core Flow Time Position Applicants Actions or Behavior When satisfied with power reduction, end of Event 2. Request that NRC the Driver insert Event 3, Reactor Building Closed Cooling Water (RBCCW) Surge Tank Low Level.

Page 13 of 57 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 3 Page 1 of 1 Event

Description:

Reactor Building Closed Cooling Water (RBCCW) Surge Tank Low Level Time Position Applicants Actions or Behavior When requested by the Chief Examiner, insert Event 3, Reactor Driver Building Closed Cooling Water (RBCCW) Surge Tank Low Level Acknowledges and reports the following alarm to the NUSO:

BOP RBCCW SURGE TANK LEVEL LOW, 3-9-4C, Window 13 Directs the Balance of Plant Operator (BOP) to respond in accordance NUSO with the appropriate Alarm Response Procedure.

Alarm Response Procedure, 3-ARP-9-4C RBCCW SURGE TANK LEVEL LOW, Window 13 A. ADD water to the RBCCW Surge Tank for approximately one minute or until low level alarm resets using the following:

3-FCV-70-1, RBCCW SYS SURGE TANK FILL VALVE (Panel 3-9-4) OR BOP 3-BYV-002-1369, FCV-70-1 BYPASS VALVE (locally)

B. IF alarm does NOT reset, THEN CHECK tank locally.

C. IF unable to maintain RBCCW Surge Tank level, THEN REFER TO 3-AOI-70-1, Loss of Reactor Building Closed Cooling Water.

D. IF necessary to add water more than once per shift, THEN CHECK Drywell floor drain system for excessive operation AND INSPECT system outside Drywell for leakage.

The RBCCW Surge Tank Low Level alarm can be cleared 15 NRC seconds after the fill valve is opened.

Opens 3-FCV-70-1, RBCCW SYS SURGE TANK FILL VALVE for BOP approximately one minute and checks that RBCCW SURGE TANK LEVEL LOW 3-9-4C, Window 13 clears.

If contacted as the Reactor Building AUO to check for leaks or check RBCCW Surge Tank Level locally, acknowledge the Driver direction. Wait 2 minutes and report that Surge Tank Level is normal and there are no leaks.

End of Event 3. Request that the Driver insert Event 4, Reactor NRC Water Cleanup (RWCU) Leak / One PCIV Fails to Close.

Page 14 of 57 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 4 Page 1 of 12 Event

Description:

Reactor Water Cleanup (RWCU) Leak / One PCIV Fails to Close Time Position Applicants Actions or Behavior When requested by the Chief Examiner, insert Event 4, Reactor Driver Water Cleanup (RWCU) Leak / One PCIV Fails to Close.

3-FCV-69-2, RWCU OUTBOARD SUCTION ISOLATION VALVE, will NRC not automatically close on the Primary Containment System Isolation (PCIS) signal, but can be closed manually.

Acknowledges and reports the following alarms:

RWCU LEAK DETECTION TEMP HIGH, 3-9-3D, Window 17 OATC/

BOP RWCU ISOL LOGIC CHANNEL A TEMP HIGH, 3-9-5B, Window 32 RWCU ISOL LOGIC CHANNEL B TEMP HIGH, 3-9-5B, Window 33 Directs the BOP to respond in accordance with applicable Alarm NUSO Response Procedures.

3-ARP-9-3D, Alarm Response Procedure RWCU LEAK DETECTION TEMP HIGH, Window 17 Operator Action:

BOP A. IF this alarm is received in conjunction with RWCU ISOL LOGIC CHANNEL A TEMP HIGH [3-XA-55-5B, window 32] and RWCU ISOL LOGIC CHANNEL B TEMP HIGH [3-XA-55-5B, window 33], THEN EXIT this procedure and GO TO 3-ARP-9-5B. Otherwise, CONTINUE in this procedure.

Exits 3-ARP-9-3D, Alarm Response Procedure, and enters 3-ARP-9-5B, BOP Alarm Response Procedure.

3-ARP-9-5B, Alarm Response Procedure RWCU ISOL LOGIC CHANNEL A TEMP HIGH, Window 32 BOP Operator Action:

A. CHECK alarm by checking:

1. ATUs on Panel 3-9-83 and 3-9-85.

Page 15 of 57 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 4 Page 2 of 12 Event

Description:

Reactor Water Cleanup (RWCU) Leak / One PCIV Fails to Close Time Position Applicants Actions or Behavior

2. RWCU LEAK DETECTION TEMP HIGH annunciator in alarm (3-XA-55-3D, Window 17).

3. Area temperature indication on 3-TR-69-29, LEAK DETECTION SYSTEM TEMPERATURE, on Panel 3-9-21.

4. Area Radiation Monitors (ARMs) 3-RR-90-1, 3-CONS-90-50A on BOP Panel 3-9-2 and 0-CONS-90-361 on Panel 1-9-2.

5. ICS HPTURB' & RWCU' mimics for the 834 and 835 temperature loops.

B. IF leak is suspected, THEN MANUALLY ISOLATE RWCU or if RWCU automatically isolates, REFER TO 3-AOI-64-2A, Group 3 Reactor Water Cleanup Isolation.

If contacted as Unit 1 Operator to check Area Radiation Monitors or Driver Radiation Recorders, acknowledge the request.

Determines that 3-FCV-69-2, RWCU OUTBOARD SUCTION BOP ISOLATION VALVE, failed to automatically isolate and manually closes 3-FCV-69-2. Informs the NUSO.

C. IF TIS-69-835A(C) indicates greater than 131 F, THEN ENTER 3-EOI-3, Secondary Containment Control.

D. REFER TO Tech. Spec. Table 3.3.6.1-1, Primary Containment BOP Isolation Instrumentation.

E. EVALUATE equipment associated with this alarm to determine compensatory actions required to maintain REP function. REFER TO NPG-SPP-18.3.5, Equipment Important to Emergency Response.

As required by 3-ARP-9-5B, Window 32 and 3-ARP-9-5B, Window 33, references NPG-SPP-18.3.5, Equipment Important to NUSO Emergency Response. No actions are required in accordance with this procedure.

NRC Technical Specifications are covered starting on page 21 of 57.

Page 16 of 57 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 4 Page 3 of 12 Event

Description:

Reactor Water Cleanup (RWCU) Leak / One PCIV Fails to Close Time Position Applicants Actions or Behavior 3-ARP-9-5B, Alarm Response Procedure RWCU ISOL LOGIC CHANNEL B TEMP HIGH, Window 33 Operator Action:

A. CHECK alarm by checking:

1. ATUs on Panel 3-9-84 and 3-9-86.

2. RWCU LEAK DETECTION TEMP HIGH annunciator in alarm (3-XA-55-3D, Window 17).

3. Area temperature indications on LEAK DETECTION SYSTEM TEMPERATURE, 3-TR-69-29, on Panel 3-9-21.

4. ARMs 3-RR-90-1, 3-CONS-90-50A on Panel 3-9-2 and BOP 0-MON-90-361 on Panel 1-9-2.

5. ICS HPTURB' & RWCU' mimics for the 834 and 835 temperature loops.

B. IF a leak is suspected, THEN MANUALLY ISOLATE RWCU or if RWCU automatically isolates, REFER TO 3-AOI-64-2A, Group 3 Reactor Water Cleanup Isolation.

C. IF TIS-69-835B(D) indicates greater than 131 F, THEN ENTER 3-EOI-3, Secondary Containment Control.

D. REFER TO Tech. Spec. Table 3.3.6.1-1, Primary Containment Isolation Instrumentation.

E. EVALUATE equipment associated with this alarm to determine compensatory actions required to maintain REP function. REFER TO NPG-SPP-18.3.5, Equipment Important to Emergency Response.

No actions are required in accordance with Technical Specification NRC 3.3.6.1.

3-AOI-64-2A, Group 3 Reactor Water Cleanup Isolation BOP Immediate Actions

[1] ENSURE automatic actions occur.

Page 17 of 57 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 4 Page 4 of 12 Event

Description:

Reactor Water Cleanup (RWCU) Leak / One PCIV Fails to Close Time Position Applicants Actions or Behavior Automatic Actions:

3-FCV-69-1, RWCU INBD SUCTION ISOLATION VALVE CLOSES 3-FCV-69-2, RWCU OUTBD SUCTION ISOLATION VALVE BOP CLOSES 3-FCV-69-12, RWCU RETURN ISOLATION VALVE CLOSES Reactor Water Cleanup Recirc Pumps 3A and 3B TRIP Subsequent Actions:

[1] IF any EOI entry condition is met, THEN ENTER appropriate EOI(s).

The NUSO may enter 3-EOI-3, Secondary Containment Control, if NRC Area Temperature or Radiation exceeds the Maximum Normal value. See page 23 of 57 for 3-EOI-3 actions.

If contacted as an AUO to check ATUs, acknowledge the direction.

Wait 3 minutes and report that ATUs 3-TIS-69-835A-D indicate 160 Driver degrees and lowering.

If contacted as Work Control or Maintenance, acknowledge any direction or reports given.

[2] CHECK the following to confirm high area temperature condition exists:

3-TR-69-29, LEAK DETECTION SYSTEM TEMPERATURE (Panel 3-9-21)

ATUs in Auxiliary Instrument Room

[3] IF isolation is caused by high area temperature, THEN DETERMINE BOP if a line break exists by:

RWCU ARMs 3-RI-90-9A, 13A, and 14A Visual Observation Rx Zone Exhaust Rad Monitors 3-RE-90-142A, 142B, 143A, and 143B

[4] PERFORM necessary Heat Balance adjustments. REFER TO 3-OI-69, Reactor Water Cleanup System.

Page 18 of 57 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 4 Page 5 of 12 Event

Description:

Reactor Water Cleanup (RWCU) Leak / One PCIV Fails to Close Time Position Applicants Actions or Behavior

[5] CHECK the following monitors for a rise in activity:

3-RR-90-1, AREA RADIATION, Points 9, 13, and 14 (Panel 3-9-2) 3-MON-90-50, AIR PARTICULATE MONITOR CONSOLE, 3-RM-90-55 and 57 (Panel 3-9-2)

RB, TB, and Refuel Zone Exhaust Rad on 0-MON-90-361, CHEMISTRY CAM, MONITOR CONTROLLER, (Panel 1-9-2)

[6] IF it has been determined that leakage is the cause of the isolation, THEN NOTIFY RADCON of RWCU status.

NOTE Based on recent On-Shift Analysis from the License Amendment, chemistry sampling may be delayed following plant events. In the case of a REP event, Chemistry will be required to assist with dose BOP assessments. Dose assessments will have precedence over sampling actions in REP events.

[7] NOTIFY Chemistry that RWCU has been removed from service for the following evaluations:

The need to begin sampling Reactor Water The need to remove the Durability Monitor from service

[8] IF the isolation cannot be reset, THEN PERFORM the following:

[8.1] ISOLATE the CRD System by closing the following seal water valves in the Unit 3 Reactor Building Elevation 593:

3-SHV-069-0592 (A pump) 3-SHV-069-0614 (B pump)

[8.2] REFER TO 3-OI-68, Reactor Recirculation System for Recirc System operating restrictions while RWCU is isolated.

If contacted as Radiation Protection or Chemistry acknowledge any directions or reports given.

Driver If contacted as Unit 1 to check Reactor Building, Turbine Building, and Refuel Zone Exhaust Radiation on 0-MON-90-361, CHEMISTRY CAM MONITOR CONTROLLER (Panel 1-9-2), acknowledge the direction.

Page 19 of 57 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 4 Page 6 of 12 Event

Description:

Reactor Water Cleanup (RWCU) Leak / One PCIV Fails to Close Time Position Applicants Actions or Behavior

[9] EVALUATE Technical Requirements Manual Section 3.4.1, Coolant NUSO Chemistry, for limiting conditions for operation.

Technical Requirements Manual 3.4.1, Coolant Chemistry.

NUSO TSR 3.4.1.1 is applicable SURVEILLANCE: FREQUENCY:

Monitor Reactor Coolant 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months when the continuous NUSO conductivity conductivity monitor is INOPERABLE and the Reactor is not in MODE 4 or 5 If contacted as Chemistry stating that Reactor Coolant sampling is Driver required every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months in accordance with TSR 3.4.1.1, acknowledge report.

Page 20 of 57 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 4 Page 7 of 12 Event

Description:

Reactor Water Cleanup (RWCU) Leak / One PCIV Fails to Close Time Position Applicants Actions or Behavior Technical Specification 3.6.1.3, Primary Containment Isolation Valves (PCIVs)

LCO 3.6.1.3 Each PCIV, except Reactor Building-to-Suppression Chamber Vacuum Breakers, shall be OPERABLE APPLICABILITY: MODES 1, 2, and 3 When associated instrumentation is required to be OPERABLE per LCO 3.3.6.1, "Primary Containment Isolation Instrumentation."

NOTES

1. Penetration flow paths except for 18 and 20 inch purge valve penetration flow paths may be un-isolated intermittently under administrative controls.

NUSO 2. Separate Condition entry is allowed for each penetration flow path.

3. Enter applicable Conditions and Required Actions for systems made inoperable by PCIVs.

4. Enter applicable Conditions and Required Actions of LCO 3.6.1.1, "Primary Containment," when PCIV leakage results in exceeding overall containment leakage rate acceptance criteria in MODES 1, 2, and 3.

CONDITION:

NOTE: Only applicable to penetration flow paths with two PCIVs.

A. - One or more penetration flow paths with one PCIV inoperable except due to MSIV leakage not within limits.

Page 21 of 57 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 4 Page 8 of 12 Event

Description:

Reactor Water Cleanup (RWCU) Leak / One PCIV Fails to Close Time Position Applicants Actions or Behavior REQUIRED ACTION: COMPLETION TIME:

A.1 - Isolate the affected A.1 - 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months except for Main penetration flow path by use of at Steam Line least one closed and de-activated automatic valve, closed manual valve, blind flange, or check valve with flow through the valve secured AND NOTE: Isolation devices in High NUSO Radiation Areas may be verified by use of administrative means.

A.2 - Once per 31 days for A.2 - Verify the affected isolation devices outside Primary penetration flow path is isolated Containment AND Prior to entering MODE 2 or 3 from MODE 4, if Primary Containment was de-inerted while in MODE 4, if not performed within the previous 92 days, for isolation devices inside Primary Containment If RWCU Room Temperature exceeds the Maximum Normal 3-EOI-3, Secondary Containment Control NUSO Page 22 of 57 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 4 Page 9 of 12 Event

Description:

Reactor Water Cleanup (RWCU) Leak / One PCIV Fails to Close Time Position Applicants Actions or Behavior 3-EOI-3, Secondary Containment Control SC-1 IF THEN Reactor Zone Ventilation Exhaust NO ACTION REQUIRED Radiation level is above 72 mR/hr Refuel Zone Ventilation Exhaust NO ACTION REQUIRED Radiation level is above 72 mR/hr NUSO Reactor Zone Ventilation is isolated AND NO ACTION REQUIRED Reactor Zone Ventilation Exhaust Radiation level is below 72 mR/hr Refuel Zone Ventilation is isolated AND NO ACTION REQUIRED Refuel Zone Ventilation Exhaust Radiation level is below 72 mR/hr Page 23 of 57 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 4 Page 10 of 12 Event

Description:

Reactor Water Cleanup (RWCU) Leak / One PCIV Fails to Close Time Position Applicants Actions or Behavior NUSO Page 24 of 57 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 4 Page 11 of 12 Event

Description:

Reactor Water Cleanup (RWCU) Leak / One PCIV Fails to Close Time Position Applicants Actions or Behavior SC/T-1 IF Reactor Zone or Refuel Zone Ventilation Exhaust Radiation Level is below 72 mr/hr THEN operate available Reactor Zone or Refuel Zone Ventilation NUSO SC-3 ISOLATE all systems that are discharging into the area EXCEPT systems required:

For damage control OR To be operated by EOIs NOTE Tables SC-1 and SC-2 contain information that may be used to determine if a primary system is discharging into Secondary Containment (Emergency Depressurization will reduce discharge).

When the RWCU Leak has been isolated and Area Temperature and Radiation is below the Maximum Safe value, the NUSO may contact NRC the Shift Manager and recommend exiting 3-EOI-3, Secondary Containment Control, as an emergency no longer exists.

Page 25 of 57 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 4 Page 12 of 12 Event

Description:

Reactor Water Cleanup (RWCU) Leak / One PCIV Fails to Close Time Position Applicants Actions or Behavior If contacted as the Shift Manager by the NUSO to discuss exiting Driver 3-EOI-3, Secondary Containment Control, agree with any recommendation given.

End of Event 4. Request that the Driver insert Event 5, Core Spray NRC Loop I Room Cooler EECW Leak.

Page 26 of 57 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 5 Page 1 of 2 Event

Description:

Core Spray Loop I Room Cooler EECW Leak Time Position Applicants Actions or Behavior When requested by the Chief Examiner, insert Event 5, Core Spray Loop I Room Cooler EECW Leak. Contact the NUSO as the Reactor Building Assistant Unit Operator (AUO) and report that you discovered and isolated a water leak in the Core Spray Loop 1 Room Cooler. Report that the following valves were closed to Driver isolate the leak:

3-SHV-67-550, NW Core Spray Room Cooler Supply Shutoff 3-SHV-67-553, NW Core Spray Room Cooler Outlet If asked, the water appears to have stopped leaking.

If contacted as Work Control or Mechanical Maintenance, Driver acknowledge any direction concerning the Core Spray Loop I Room Cooler.

Technical Requirements Manual 3.5.3, Equipment Area Coolers LCO 3.5.3 The Equipment Area Cooler associated with each RHR Pump and the Equipment Area Cooler associated with each set of Core Spray Pumps (A and C or B and D) must be OPERABLE at all times when the pump or pumps served by that specific cooler is considered to NUSO be OPERABLE.

APPLICABILITY: Whenever the associated subsystem is required to be OPERABLE CONDITION:

A. - One or more Equipment Area Cooler inoperable REQUIRED ACTION: COMPLETION TIME:

A.1 - Declare the pump(s) served A.1 - Immediately NUSO by that cooler INOPERABLE (Refer to applicable Tech Spec and TRM LCOs)

Page 27 of 57 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 5 Page 2 of 2 Event

Description:

Core Spray Loop I Room Cooler EECW Leak Time Position Applicants Actions or Behavior Technical Specification 3.5.1, ECCS - Operating LCO 3.5.1 Each ECCS injection/spray subsystem and the Automatic Depressurization System (ADS) function of six safety/relief valves shall be OPERABLE APPLICABILITY: MODE 1, MODES 2 and 3, except High Pressure Coolant Injection (HPCI) and ADS valves are not required to NUSO be OPERABLE with Reactor Steam Dome Pressure 150 psig CONDITION:

A. - One low pressure ECCS injection/spray subsystem inoperable.

OR One low pressure coolant injection (LPCI) pump in both LPCI subsystems inoperable.

REQUIRED ACTION: COMPLETION TIME:

NUSO See Condition F See Condition F CONDITION:

F. - One ADS Valve inoperable NUSO AND Condition A entered REQUIRED ACTION: COMPLETION TIME:

F.1 - Restore ADS Valve to F.1 - 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months OPERABLE status NUSO OR F.2 - Restore Low Pressure ECCS F.2 - 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Injection / Spray subsystem to OPERABLE status End of Event 5. Request that the Driver insert Event 6, 3C 4KV Unit NRC Board Trip.

Page 28 of 57 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 6 Page 1 of 3 Event

Description:

3C 4KV Unit Board Trip Time Position Applicants Actions or Behavior When requested by the Chief Examiner, insert Event 6, 3C 4KV Unit Driver Board Trip.

Acknowledges and reports the following alarms:

4KV UNIT BOARD 3C UNDERVOLTAGE, 3-9-8B, Window 14 BOP CONDENSATE BOOSTER PUMP C AUX OIL PRESSURE LOW, 3-9-6A, Window 14 MOTOR TRIPOUT, 3-9-8C, Window 33 CREW Monitors Reactor Water Level.

Reports a loss of Control Rod Drive (CRD) System Flow due to 3A CRD OATC Pump being de-energized.

Directs the BOP to respond in accordance with applicable Alarm NUSO Response Procedures and direct the OATC to respond in accordance with 3-AOI-85-3, CRD System Failure.

3-AOI-85-3, CRD System Failure Immediate Actions

[1] IF operating CRD PUMP has failed AND the standby CRD Pump is available, THEN PERFORM the following at Panel 3-9-5: (Otherwise N/A)

[1.1] PLACE 3-FIC-85-11, CRD SYSTEM FLOW CONTROL in MAN at minimum setting.

[1.2] START associated standby CRD Pump using the following:

OATC 3-HS-85-2A, CRD PUMP 3B

[1.3] ADJUST 3-FIC-85-11, CRD SYSTEM FLOW CONTROL, to establish the following conditions:

3-PDI-85-18A, CRD COOLING WATER HEADER DP, between 10 psid and 20 psid 3-FIC-85-11, CRD SYSTEM FLOW CONTROL, between 40 and 65 gpm

[1.4] BALANCE CRD SYSTEM FLOW CONTROL, 3-FIC-85-11, and PLACE in AUTO or BALANCE.

Page 29 of 57 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 6 Page 2 of 3 Event

Description:

3C 4KV Unit Board Trip Time Position Applicants Actions or Behavior 3-ARP-9-8B, Alarm Response Procedure 4KV UNIT BOARD 3C UNDERVOLTAGE, Window 14 Operator Action:

A. CHECK Unit in stable condition by checking:

Condensate Pump 3C Condensate Booster Pump 3C RCW Pump 3C CCW Pump 3C BOP CRD Pump 3A B. IF undervoltage has occurred, THEN

1. CLEAR disagreement lights on breakers.

2. REDUCE load as necessary to maintain stable operating conditions.

3. Condenser discharge may need to be throttled for two CCW pump operation. REFER TO 3-OI-27, Condenser Circulating Water System.

C. CHECK Unit Bd 3C for abnormal conditions: relay targets, smoke, burned paint, etc.

D. REFER TO 0-OI-57A, Switchyard and 4160V AC Electrical System, to re-energize board.

E. REFER TO appropriate OI for recovery or realignment of equipment.

If contacted as an AUO, Work Control, or Electrical Maintenance to investigate, acknowledge the direction. If directed to prepare Driver protected equipment tags acknowledge the direction.

Wait 2 minutes and report that 3C 4KV Unit Board has an overcurrent trip flag.

Page 30 of 57 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 6 Page 3 of 3 Event

Description:

3C 4KV Unit Board Trip Time Position Applicants Actions or Behavior 3-ARP-9-8C, Alarm Response Procedure MOTOR TRIPOUT, Window 33 Operator Action:

A. CHECK Control Room for white disagreement light illuminated for affected equipment.

B. CLEAR disagreement light.

BOP C. DISPATCH personnel to CHECK:

1. Relays at associated electrical board.

2. Equipment for abnormal conditions.

3. Safe-stop locally reset, if necessary.

D. REFER TO 0-GOI-300-2, Electrical, if relay targets are present or for motor starting limits.

E. REFER TO appropriate OI for recovery or realignment of equipment.

3-ARP-9-6A, Alarm Response Procedure CONDENSATE BOOSTER PUMP C AUX OIL PUMP PRESS LOW, Window 14 Operator Action:

BOP A. DISPATCH personnel to check booster pump lube oil system:

1. ENSURE running or start Aux Oil Pump.

2. CHECK for leaks.

3. CHECK oil level and temperature at reservoir.

4. ROTATE Cuno Filter.

If contacted as the Turbine Building AUO to start 3C Condensate Driver Booster Pump Aux Oil Pump, insert Event 16 and report that the Aux Oil Pump is running.

End of Event 6. Request that the Driver insert Event 7 NRC Un-isolable Reactor Core Isolation Cooling (RCIC) Steam Leak.

Page 31 of 57 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 7 Page 1 of 11 Event

Description:

Un-isolable Reactor Core Isolation Cooling (RCIC) Steam Leak Time Position Applicants Actions or Behavior When requested by the Chief Examiner, insert Event 7, Driver Un-isolable Reactor Core Isolation Cooling (RCIC) Steam Leak.

Event 8, Fuel Damage, and Event 9, Reactor Feedwater Pumps Trip / HPCI Fails to Automatically Start and Inject, are automatically NRC entered by the simulator setup. No action is required by the Driver to Insert Event 8 or Event 9. See page 43 of 57 for Event 8 actions and page 50 of 57 for Event 9 actions.

Acknowledges and reports the following alarms as they are received:

REACTOR BUILDING RADIATION HIGH, 3-9-3A, Window 22 BOP REACTOR BUILDING, TURBINE BUILDING, REFUEL ZONE EXHAUST RADIATION HIGH, 3-9-3A, Window 4 RCIC STEAM LINE LEAK DETECTION TEMPERATURE HIGH, 3-9-3D, Window 10 NRC See Event 8 (page 43 of 57) for actions for Radiation Alarms.

Directs the BOP to respond in accordance with the applicable Alarm NUSO Response Procedure.

3-9-ARP-3D, Alarm Response Procedure RCIC STEAM LINE LEAK DETECTION TEMPERATURE HIGH, Window 10 BOP Operator Action:

A. CHECK RCIC temperature elements on LEAK DETECTION SYSTEM TEMPERATURE recorder, 3-TR-69-29 on Panel 3-9-21.

BOP Checks Area Temperatures on Panel 3-9-22.

B. IF RCIC is NOT in service AND 3-FI-71-1A(B), RCIC STEAM FLOW BOP indicates flow, THEN ISOLATE RCIC and CHECK temperatures lowering.

Determines that RCIC failed to automatically isolate, and attempts to BOP manually isolate RCIC. Informs the NUSO that RCIC will not isolate.

C. IF high temperature is confirmed, THEN ENTER 3-EOI-3, Secondary BOP Containment Control.

Page 32 of 57 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 7 Page 2 of 11 Event

Description:

Un-isolable Reactor Core Isolation Cooling (RCIC) Steam Leak Time Position Applicants Actions or Behavior BOP Confirms that Area Temperatures are rising and informs the NUSO.

Enters 3-EOI-3, Secondary Containment Control. Directs the BOP to NUSO monitor Secondary Containment parameters.

D. CHECK CS/RCIC ROOM El 519 RX BLDG radiation indicator, 3-RI-90-26A on Panel 3-9-11 and NOTIFY RADCON if rising radiation BOP levels are observed.

E. DISPATCH personnel to investigate.

If contacted as Radiation Protection that radiation levels are rising, acknowledge the report.

Driver If contacted as the Reactor Building AUO to investigate, acknowledge the direction.

F. REFER TO Tech Specs 3.3.6.1, Primary Containment Isolation NUSO Instrumentation and 3.5.3, RCIC System.

Technical Specification evaluation for this event is not required NRC and should not be used to evaluate the candidates Technical Specification competency.

G. EVALUATE equipment associated with this alarm to determine compensatory actions required to maintain REP function. REFER TO NPG-SPP-18.3.5, Equipment Important to Emergency Response.

It is not expected that the SRO reference NPG-SPP-18.3.5, NRC Equipment Important to Emergency Response, during this event.

Page 33 of 57 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 7 Page 3 of 11 Event

Description:

Un-isolable Reactor Core Isolation Cooling (RCIC) Steam Leak Time Position Applicants Actions or Behavior 3-EOI-3, Secondary Containment Control SC-1 IF THEN Reactor Zone Ventilation Exhaust NO ACTION REQUIRED Radiation level is above 72 mR/hr Refuel Zone Ventilation Exhaust NO ACTION REQUIRED Radiation level is above 72 mR/hr Reactor Zone Ventilation is isolated AND NO ACTION REQUIRED NUSO Reactor Zone Ventilation Exhaust Radiation level is below 72 mR/hr Refuel Zone Ventilation is isolated AND NO ACTION REQUIRED Refuel Zone Ventilation Exhaust Radiation level is below 72 mR/hr Page 34 of 57 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 7 Page 4 of 11 Event

Description:

Un-isolable Reactor Core Isolation Cooling (RCIC) Steam Leak Time Position Applicants Actions or Behavior NUSO Page 35 of 57 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 7 Page 5 of 11 Event

Description:

Un-isolable Reactor Core Isolation Cooling (RCIC) Steam Leak Time Position Applicants Actions or Behavior SC/T-1 IF Reactor Zone or Refuel Zone Ventilation Exhaust Radiation level is below 72 mR/hr THEN Operate available Reactor Zone or Refuel Zone Ventilation SC/T-2 WHEN ANY Area Temperature exceeds its Max Normal Temperature (Table SC-1)

NUSO Page 36 of 57 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 7 Page 6 of 11 Event

Description:

Un-isolable Reactor Core Isolation Cooling (RCIC) Steam Leak Time Position Applicants Actions or Behavior SC-3 ISOLATE all systems that are discharging into the area EXCEPT systems required:

For damage control OR To be operated by EOIs NOTE Tables SC-1 and SC-2 contain information that may be used to determine if a primary system is discharging into Secondary Containment (Emergency Depressurization NUSO will reduce discharge).

SC-4 SC-7 WHEN A Primary System is discharging into Secondary Containment Page 37 of 57 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 7 Page 7 of 11 Event

Description:

Un-isolable Reactor Core Isolation Cooling (RCIC) Steam Leak Time Position Applicants Actions or Behavior SC-8 BEFORE ANY Secondary Containment parameter reaches its Max Safe Value (Tables SC-1, SC-2, and SC-3)

NUSO Critical Task:

With the Reactor at power and with a Primary System discharging into the Secondary Containment, manually SCRAM the Reactor before any area exceeds the Maximum Safe Temperature operating value by entering EOI-1, RPV Control.

CREW Critical Task Failure Criteria:

The operating crew fails to proceed without delay and in a controlled manner to initiate a Reactor SCRAM from the time it is announced that one Area Temperature is approaching the Maximum Safe value.

Enters 3-EOI-1, RPV Control. Directs the crew to enter 3-AOI-100-1, NUSO Reactor SCRAM, and directs the OATC to insert a manual Reactor SCRAM.

Event 8, Fuel Damage, and Event 9, Reactor Feedwater Pumps NRC Trip / HPCI Fails to Automatically Initiate, are inserted when the Reactor MODE SWITCH is placed in SHUTDOWN.

OATC Inserts a manual Reactor SCRAM.

Page 38 of 57 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 7 Page 8 of 11 Event

Description:

Un-isolable Reactor Core Isolation Cooling (RCIC) Steam Leak Time Position Applicants Actions or Behavior 3-AOI-100-1, Reactor SCRAM Immediate Actions

[1] DEPRESS 3-HS-99-5A/S3A, REACTOR SCRAM A and 3-HS-99-5A/S3B, REACTOR SCRAM B, on Panel 3-9-5.

[2] PLACE 3-HS-99-5A/S1, REACTOR MODE SWITCH, in SHUTDOWN.

[3] IF all Control Rods can NOT be verified fully inserted, THEN INITIATE ARI. (Otherwise MARK N/A).

[4] IF Reactor Power is 5% or BELOW, THEN: (Otherwise MARK N/A)

OATC REPORT the following to the UNIT SRO:

Reactor Scram Mode Switch is in Shutdown All rods in or rods out Reactor Water Level and trend (recovering or lowering)

Reactor Pressure and trend MSIV position (Open or Closed)

Power level

[5] N/A Determines that all Reactor Feedwater Pumps (RFPTs) have tripped OATC and informs the NUSO (See Event 9).

3-EOI-1, RPV Control NUSO RC/L-1 ENSURE each as required:

PCIS isolations (Groups 1, 2, and 3)

ECCS RCIC Page 39 of 57 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 7 Page 9 of 11 Event

Description:

Un-isolable Reactor Core Isolation Cooling (RCIC) Steam Leak Time Position Applicants Actions or Behavior RC/L-2 IF THEN RPV Water Level can be restored and maintained above

(-)162 in. INHIBIT ADS AND The ADS timer has initiated Loss of available injection systems is anticipated OR Raising RPV Water Level above

(+)51 in. will facilitate use of NO ACTION REQUIRED shutdown cooling, steam-driven injection systems, or Alternate NUSO Injection Subsystems (Table L-2)

RC/L-3 RESTORE and MAINTAIN RPV Water Level between (+)2 in. and

(+)51 in. with ANY Preferred Injection Systems (Table L-1)

IF THEN RPV Water Level cannot be restored and maintained NO ACTION REQUIRED between (+)2 in. and (+)51 in.

RPV Water Level cannot be restored and maintained above NO ACTION REQUIRED

(-)162 in.

Page 40 of 57 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 7 Page 10 of 11 Event

Description:

Un-isolable Reactor Core Isolation Cooling (RCIC) Steam Leak Time Position Applicants Actions or Behavior Directs the OATC/BOP to maintain Reactor Water Level using HPCI in NUSO accordance with 3-EOI-Appendix-5D, Injection System Lineup HPCI.

3- EOI-Appendix-5D, Injection System Lineup HPCI actions are NRC covered in Event 9. See page 50 of 57.

NUSO RC/P-1 IF THEN A high Drywell Pressure ECCS NO ACTION REQUIRED signal exists (2.45 psig)

EMERGENCY RPV DEPRESSURIZATION is C2 Emergency RPV NUSO REQUIRED or has been Depressurization required Emergency RPV NO ACTION REQUIRED Depressurization is anticipated RC-P/2 IF ANY MSRV is cycling THEN NO ACTION REQUIRED Page 41 of 57 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 7 Page 11 of 11 Event

Description:

Un-isolable Reactor Core Isolation Cooling (RCIC) Steam Leak Time Position Applicants Actions or Behavior RC/P-3 IF THEN Suppression Pool Temperature and Level CANNOT be maintained in a safe area of NO ACTION REQUIRED Curve 3 at the existing RPV Pressure Suppression Pool Level CANNOT be maintained in the NO ACTION REQUIRED safe area of Curve 4 NO ACTION REQUIRED NUSO RC/P-4 STABILIZE RPV Pressure below 1073 psig using the Main Turbine Bypass Valves (APPX 8B)

OK to use ANY Alternate RPV Pressure Control Systems (Table P-1)

Crosstie CAD or MSRV carts to DW Control Air (APPX 8G, 20H) if necessary IF THEN DW Control Air is or becomes NO ACTION REQUIRED unavailable Directs the BOP to maintain Reactor Pressure using the Main Turbine NUSO Bypass Valves.

NUSO NRC End of Event 7. Continue to Event 8.

Page 42 of 57 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 8 Page 1 of 7 Event

Description:

Fuel Damage Time Position Applicants Actions or Behavior Event 8, Fuel Damage, is inserted when the Reactor MODE SWITCH NRC is placed in SHUTDOWN. No action is required by the driver to insert Event 8.

Acknowledges and reports the following alarms to the NUSO as they are received:

BOP REACTOR BUILDING RADIATION HIGH, 3-9-3A, Window 22 REACTOR BUILDING, TURBINE BUILDING, REFUEL ZONE EXHAUST RADIATION HIGH, 3-9-3A, Window 4 3-ARP-9-3A, Alarm Response Procedure REACTOR BUILDING RADIATION HIGH, Window 22 BOP Operator Action:

A. DETERMINE area with high radiation level on Panel 3-9-11. (Alarm on Panel 3-9-11 will automatically reset if radiation level lowers below setpoint.)

Monitors Radiation Levels on Panel 3-9-11. Keeps the NUSO informed BOP on instruments which are alarming and which are approaching Maximum Safe Values.

B. N/A BOP C. NOTIFY Radiation Protection.

If contacted as Radiation Protection, acknowledge any reports or Driver direction given.

D. IF the TSC is NOT manned and a VALID radiological condition exists., THEN USE public address system to evacuate area where high BOP radiological conditions exist E. N/A Makes a plant announcement to evacuate the Reactor Building due to BOP high radiation.

F. MONITOR other parameters providing input to this annunciator BOP frequently as these parameters will be masked from alarming while this alarm is sealed in.

Page 43 of 57 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 8 Page 2 of 7 Event

Description:

Fuel Damage Time Position Applicants Actions or Behavior G. IF a CREV initiation is received, THEN

1. CHECK CREV A(B) Flow is 2700 CFM, and 3300 CFM as indicated on 0-FI-031-7214(7213) within 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months of the CREV initiation.

2. IF CREV A(B) Flow is NOT 2700 CFM, and 3300 CFM as BOP indicated on 0-FI-031-7214(7213), THEN PERFORM the following:

(Otherwise N/A)

a. STOP the operating CREV per 0-OI-31, Control Bay and Off-Gas Treatment Building Air Conditioning System.

b. START the standby CREV per 0-OI-31, Control Bay and Off-Gas Treatment Building Air Conditioning System.

If contacted as an AUO to monitor CREV operation, acknowledge Driver the direction.

H. N/A BOP I. ENTER 3-EOI-3, Secondary Containment Control.

Re-enters 3-EOI-3, Secondary Containment Control (if not already NUSO entered on Secondary Containment Radiation).

K. EVALUATE equipment associated with this alarm to determine BOP compensatory actions required to maintain REP function. REFER TO NPG-SPP-18.3.5, Equipment Important to Emergency Response.

It is not expected that the SRO reference NPG-SPP-18.3.5, NRC Equipment Important to Emergency Response, during this event.

Page 44 of 57 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 8 Page 3 of 7 Event

Description:

Fuel Damage Time Position Applicants Actions or Behavior SC/R-1 WHEN ANY Area Radiation Level exceeds its Max Normal Radiation Level (Table SC-2)

NUSO Page 45 of 57 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 8 Page 4 of 7 Event

Description:

Fuel Damage Time Position Applicants Actions or Behavior SC-3 ISOLATE all systems that are discharging into the area EXCEPT systems required:

For damage control OR To be operated by EOIs NOTE Tables SC-1 and SC-2 contain information that may be used to determine if a primary system is discharging into NUSO Secondary Containment (Emergency Depressurization will reduce discharge).

SC-4 SC-7 WHEN A Primary System is discharging into Secondary Containment Page 46 of 57 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 8 Page 5 of 7 Event

Description:

Fuel Damage Time Position Applicants Actions or Behavior SC-8 BEFORE ANY Secondary Containment parameter reaches its Max Safe Value (Tables SC-1, SC-2, and SC-3)

NUSO Monitors Area Radiation levels and informs the NUSO when two areas BOP are at Maximum Safe.

SC-9 WHEN Any Secondary Containment parameter exceeds its Max Safe value in two (2) or more areas for the same parameter (Tables SC-1, NUSO SC-2, SC-3)

SC-10 EMERGENCY DEPRESSURIZATION IS REQUIRED Updates the crew that Emergency Depressurization is required. Enters NUSO 3-C-2, Emergency RPV Depressurization.

Page 47 of 57 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 8 Page 6 of 7 Event

Description:

Fuel Damage Time Position Applicants Actions or Behavior Critical Task:

With a Primary System discharging into the Secondary Containment when two or more areas are greater than their maximum safe operating values for the same parameter, the Balance of Plant Operator initiates Emergency Depressurization Crew as directed by the Nuclear Unit Senior Operator.

Critical Task Failure Criteria:

The operating crew fails to proceed with without delay and in a controlled manner to initiate Emergency Depressurization when two Area Radiation Levels exceed Maximum Safe value.

3-C-2, Emergency RPV Depressurization C2-1 IF THEN Reactor Water Level CANNOT NO ACTION REQUIRED be determined NUSO It is anticipated that Reactor depressurization will result in NO ACTION REQUIRED loss of injection required for Adequate Core Cooling Containment Water Level CANNOT be maintained below NO ACTION REQUIRED 44 feet C2-2 IF Drywell Pressure is above 2.45 PSIG NUSO THEN PREVENT injection from ONLY those Core Spray and LPCI pumps NOT required to assure Adequate Core Cooling (Appendix 4)

Page 48 of 57 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 8 Page 7 of 7 Event

Description:

Fuel Damage Time Position Applicants Actions or Behavior C2-3 EMERGENCY DEPRESSURIZE the Reactor IF Suppression Pool Water Level is above 5.5 feet THEN OPEN 6 MSRVs (ADS Valves preferred)

OK to exceed 100ºF/hr cooldown rate IF THEN NUSO Drywell Control Air is or NO ACTION REQUIRED becomes unavailable Less than 4 MSRVs can be opened AND NO ACTION REQUIRED Reactor Pressure is 80 PSIG or more above Suppression Chamber Pressure Opens 5 SRVs and one additional SRV (due to ADS Valve 1-22 being BOP out of service).

C2-4 WHEN NUSO Shutdown Cooling RPV Pressure interlock clears AND further cooldown is required End of Event 8. When the crew has Emergency Depressurized the Reactor and has control of Reactor Water Level above the Top of NRC Active Fuel ((-) 162 inches) using low pressure systems, end of Scenario.

Page 49 of 57 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 9 Page 1 of 3 Event

Description:

Reactor Feedwater Pumps (RFPTs) Trip / HPCI Fails to Automatically Start and Inject Time Position Applicants Actions or Behavior Event 9, Reactor Feedwater Pumps Trip / HPCI Fails to Automatically Start and Inject, is inserted when the Reactor MODE NRC SWITCH is placed in SHUTDOWN. No action is required by the Driver to insert Event 9.

OATC Reports that all Reactor Feedwater Pumps (RFPTs) have tripped.

When Reactor Water Level reaches the High Pressure Coolant Injection (HPCI) initiation setpoint (-45), determines that HPCI did not BOP automatically start and manually starts HPCI. Informs the NUSO of the actions required to start HPCI.

Directs the OATC/BOP to maintain Reactor Water Level using NUSO 3-EOI-Appendix-5D, Injection System Lineup HPCI.

3-EOI-Appendix-5D, Injection System Lineup HPCI

[1] IF Suppression Pool level drops below 12.75 ft. during HPCI operation, THEN TRIP HPCI and CONTROL injection using other options.

[2] IF Suppression Pool level CANNOT be maintained below 5.25 in, THEN EXECUTE EOI Appendix 16E concurrently with this procedure to bypass HPCI High Suppression Pool Water Level Suction Transfer Interlock.

[3] IF BOTH of the following exist:

BOP High temperature exists in the HPCI area, AND SRO directs bypass of HPCI High Temperature Isolation Interlocks, THEN PERFORM the following:

[3.1] EXECUTE EOI Appendix 16L, Bypassing HPCI High Temperature Isolation concurrently with this procedure.

[3.2] RESET auto isolation logic using HPCI AUTO-ISOL LOGIC A (B) RESET pushbuttons.

[4] VERIFY at least one SGTS train in operation.

Page 50 of 57 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 9 Page 2 of 3 Event

Description:

Reactor Feedwater Pumps (RFPTs) Trip / HPCI Fails to Automatically Start and Inject Time Position Applicants Actions or Behavior CAUTIONS

1) Operating HPCI Turbine below 2400 rpm may result in unstable system operation and equipment damage.

2) Operating HPCI Turbine with suction temperatures above 140 F may result in equipment damage.

[5] VERIFY 3-FIC-73-33, HPCI SYSTEM FLOW/CONTROL, controller is in one of the following configurations, as desired:

in AUTO and set for 5300 gpm for rapid injection in AUTO and set for approximately 2500 gpm for slower injection in MANUAL with output at approximately 50% for slower injection NOTE BOP HPCI Auxiliary Oil Pump will NOT start UNTIL 3-FCV-73-16, HPCI TURBINE STEAM SUPPLY VALVE, starts to open.

[6] IF high Reactor Water Level Trip logic is actuated, THEN

[6.1] DEPRESS HPCI TURBINE TRIP RX LVL HIGH RESET pushbutton.

[6.2] CHECK HPCI TURBINE TRIP LVL HIGH amber light has extinguished.

[7] PLACE HPCI AUXILIARY OIL PUMP handswitch in START.

[8] PLACE HPCI STEAM PACKING EXHAUSTER handswitch in START.

[9] OPEN 3-FCV-73-30, HPCI PUMP MIN FLOW VALVE.

[10] OPEN 3-FCV-73-44, HPCI PUMP INJECTION VALVE.

[11] OPEN 3-FCV-73-16, HPCI TURBINE STEAM SUPPLY VALVE, to start HPCI Turbine.

Page 51 of 57 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-1 Event No.: 9 Page 3 of 3 Event

Description:

Reactor Feedwater Pumps (RFPTs) Trip / HPCI Fails to Automatically Start and Inject Time Position Applicants Actions or Behavior

[12] CHECK proper HPCI operation by observing the following:

A. HPCI Turbine speed accelerates.

B. 3-CKV-73-45, HPCI SYSTEM CHECK VALVE, opens by observing 3-ZI-73-45A, DISC POSITION, red light illuminated.

C. HPCI flow to RPV stabilizes and is controlled automatically at the setpoint. (N/A if controller in manual).

D. 3-FCV-73-30, HPCI PUMP MIN FLOW VALVE, closes as flow exceeds approximately 1200 gpm.

CAUTION 3-FCV-073-0030, HPCI PUMP MIN FLOW VALVE, automatically BOP opens when system flow is at or below 900 gpm (lowering) only if a system initiation signal is present. Manually opening the min flow valve may be required for pump min flow protection.

[13] ADJUST 3-FIC-73-33, HPCI SYSTEM FLOW/CONTROL, controller as necessary to control injection.

[14] VERIFY HPCI Auxiliary Oil Pump stops and the shaft-driven oil pump operates properly.

[15] WHEN HPCI Auxiliary Oil Pump stops, THEN PLACE HPCI AUXILIARY OIL PUMP handswitch in AUTO.

[16] N/A

[17] N/A End of Event 9. When the crew has Emergency Depressurized the Reactor and has control of Reactor Water Level above the Top of NRC Active Fuel ((-) 162 inches) using low pressure systems, end of Scenario.

Page 52 of 57 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Scenario Setup UNIT 3 IC 54 Exam IC 251 Procedure Revision Procedure Revision Procedure Revision OI-68 99 EOI-3 12 ARP 6A 27 OI-69 109 3-C-2 13 ARP 8B 24 GOI-12 49 APPX 5D 8 ARP 8C 20 GOI-12A 72 ARP 3A 57 TRM 3.4.1 21 AOI-64-2A 12 ARP 3D 32 TS 3.5.1 229 AOI-85-3 13 ARP 4B 53 TS 3.5.3 244 AOI-100-1 74 ARP 4C 42 TS 3.6.1.3 212 EOI-1 13 ARP 5B 32 Verify camera system is powered down (admin password = abcd1234)

Simulator Setup Start CPERF PRIOR to placing the Simulator in RUN Ensure Danger Tags are placed on SRV 1-22 and the Emergency High Pressure Makeup Pump 2104 NRC Scenario 1 UNIT 3.sch Schedule Files(s):

RWCU.sch Event Files(s): 2104 NRC Scenario 1 UNIT 3.evt Schedule File: 2104 NRC Scenario 1 UNIT 3.sch Event Action Description 2104 NRC Scenario 1 Unit 3.evt Event File 1 Insert remote CU01 to 45.00000 in 60 RWCU DEMIN FILTER A FRC-69-35 1 Insert remote CU02 to 45.00000 in 60 RWCU DEMIN FILTER B FRC-69-60 1 Insert remote AN01E after 15 to RESET CU LOCAL RESET (3-XA-55-4B W24) 11 Insert remote CU05 to MANUAL RWCU HX RBCCW FLOW CONTROL TIC-69-10A 21 Insert remote CU05 to AUTO RWCU HX RBCCW FLOW CONTROL TIC-69-10A Page 53 of 57 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Schedule File: 2104 NRC Scenario 1 UNIT 3.sch Event Action Description 3 Insert malfunction XA-55-4C_13 to ON RBCCW SURGE TANK LEVEL LOW 3-LA-70-2B Insert malfunction XA-55-4C_13 after 13 RBCCW SURGE TANK LEVEL LOW 3-LA-70-2B 10 to NORMAL delete in 1 Schedule F:\Scenarios\Scenario 4 RWCU Leak 1\RWCU.sch 4KV UNIT BOARD 3C FAILURE (RELAY 86-316 6 Insert malfunction ED08C AND 86-532)

CONDENSATE BOOSTER PUMP C AUX OIL 16 Insert remote FW06 to START PUMP 16 Insert remote AN01D to RESET in 5 on FW LOCAL RESET 121 (3-XA-55-6B W6)

Insert malfunction RC09 to 35.00000 7 RCIC STEAM LEAK INTO RCIC ROOM in 1800 Insert malfunction FCV-71-2 to MOTOR_OPERATED_VALVE RCIC STEAM LINE FAIL_NOW INBD ISOL VALVE Insert override ZLOHS712A_1 to Off HS-71-2A RCIC STEAM LINE INBD ISOL VALVE Insert override ZLOHS712A_2 to On HS-71-2A RCIC STEAM LINE INBD ISOL VALVE 17 Delete override ZLOHS712A_2 HS-71-2A RCIC STEAM LINE INBD ISOL VALVE Insert malfunction FCV-71-3 to MOTOR_OPERATED_VALVE RCIC STEAM LINE FAIL_NOW OUTBD ISOL VALVE HS-71-3A RCIC STEAM LINE OUTBD ISOL Insert override ZLOHS713A_1 to Off VALVE HS-71-3A RCIC STEAM LINE OUTBD ISOL Insert override ZLOHS713A_2 to On VALVE Insert malfunction TH23 to 10.00000 in 8 FUEL CLADDING DAMAGE 900 Page 54 of 57 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Schedule File: 2104 NRC Scenario 1 UNIT 3.sch Event Action Description RFPT 3A TRIP ON RFPT 3A BEARING LOW OIL 8 Insert malfunction FW14A PRESSURE (PS-3-123B)

RFPT 3B TRIP ON RFPT 3B BEARING LOW OIL 8 Insert malfunction FW14B PRESSURE (PS-3-149B)

RFPT 3C TRIP ON RFPT 3C BEARING LOW OIL 8 Insert malfunction FW14C PRESSURE (PS-3-174B)

Insert malfunction FCV-73-16 to MOTOR_OPERATED_VALVE HPCI TURBINE FAIL_NOW STEAM SUPPLY VALVE HS-73-16A HPCI TURBINE STEAM SUPPLY Insert override ZLOHS7316A_1 to On VALVE MOTOR_OPERATED_VALVE HPCI TURBINE 19 Delete malfunction FCV-73-16 STEAM SUPPLY VALVE HS-73-16A HPCI TURBINE STEAM SUPPLY 19 Delete override ZLOHS7316A_1 VALVE Insert malfunction RC09 to 18.00000 in 27 RCIC STEAM LEAK INTO RCIC ROOM 1800 on event 27 Insert malfunction RC09 to 100.00000 28 RCIC STEAM LEAK INTO RCIC ROOM in 1800 on event 8 Schedule File: RWCU.sch Event Action Description Insert malfunction CU04 to 25.00000 RWCU SYSTEM SUCTION BREAK Insert malfunction FCV-69-2 to MOTOR_OPERATED_VALVE RWCU OUTBOARD FAIL_NOW ISOLATION VLV HS-69-2A RWCU OUTBD SUCT ISOLATION Insert override ZLOHS692A_1 to Off VALVE HS-69-2A RWCU OUTBD SUCT ISOLATION Insert override ZLOHS692A_2 to On VALVE MOTOR_OPERATED_VALVE RWCU OUTBOARD 14 Delete malfunction FCV-69-2 ISOLATION VALVE HS-69-2A RWCU OUTBD SUCT ISOLATION 14 Delete override ZLOHS692A_1 VALVE HS-69-2A RWCU OUTBD SUCT ISOLATION 14 Delete override ZLOHS692A_2 VALVE Page 55 of 57 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Event File List Details Page 56 of 57 Unit 3

UNIT 3 SHIFT TURNOVER MEETING Today DAYS ON LINE Drywell Leakage (GPM) Protected Equipment MODE 1 234 None PRA (EOOS) -Green 1.81 Rx Power 500Kv GRID - Qualified Floor Drain (GPM) 80.0% 161Kv Grid -Qualified 0.27 MWe Last breaker closure Equipment Drain (GPM) 1012 3/15/19 5:41 1.54 Review logs Qualifications Review RCP/Rx Brief Review LCO/OWA Actions Walkdown Panels/Verify EOOS CR Reviews Complete Leadership and Team Effectiveness CHANGES IN LCOs SRV 1-22 is INOPERABLE - ADS Valve. Tech Spec 3.5.1.E, Condition E (Day 4 of 14 day LCO)

LCOs OF 72 HOURS OR LESS SIGNIFICANT ITEMS DURING PREVIOUS SHIFT/RADIOLOGICAL CHANGES Reactor Shutdown. Maintain RFPTs, Condensate, and Condensate Booster Pumps running until Reactor Power is <70%

EHPM tagged for bearing inspection.

RWCU Non-Regenerative Heat Exchanger repairs completed, tags are clear, and RWCU has been filled and vented.

MAJOR EQUIPMENT CHANGES PLANNED FOR THIS SHIFT Return RWCU to service in accordance with 3-OI-69, Section 5.1, beginning at Step [4]. RPHP is in effect.

Continue the Reactor Shutdown in accordance with 3-GOI-100-12A, Section 5.3.1, Step [4].

Reduce Reactor Power to 75% using Core Flow, then wait for further guidance from RE.

OPERATOR WORK AROUNDS OWAs - 0 Burdens - 0 Challenges - 6 ODMIs/ACMPs ONEAs FIRE RISK SIGNIFICANT ITEMS OOS/FPLCO Actions Due SCHEDULED ITEMS NOT COMPLETED Page 57 of 57 Unit 3

Appendix D Scenario Outline Form ES-D1 Facility: BFN Scenario Number: NRC-2 Op-Test Number:21-04 Examiners: __________________________ Operators: SRO: _____________________

__________________________ ATC: _____________________

BOP: _____________________

Initial Conditions: 100% Reactor Power.

Turnover: 100% Reactor Power. A3 EECW Pump is tagged for oil change. 2B EHC Pump is tagged for maintenance. 2A CCW Pump ready for restart.

Critical Tasks:

1. With a Reactor SCRAM required and the Reactor not shutdown, to prevent an uncontrolled Reactor depressurization and subsequent power excursion inhibit ADS or control Reactor Water Level such that no automatic ADS actuation occurs.

2. When Reactor Power is greater than 5% or unknown during an ATWS, STOP and PREVENT all injection into the Reactor except for RCIC, CRD, and SLC within 130 seconds of the loss of forced recirculation to prevent possible fuel damage.

Event Malfunction Event Event Description Number Number Type*

N-BOP

1. N/A Swap Recirc Drive Cooling Water Pumps N-NUSO C-BOP

2. SW03F EECW Pump Trip TS-NUSO C-OATC

3. NMAPRMGAIN(1) APRM 1 Fails Downscale C-NUSO C-OATC 4.# RD04R3023 Control Rod Drifts Out TS-NUSO C-BOP

5. OG04A 2A Steam Jet Air Ejector (SJAE) Trip C-NUSO R-OATC Clogged Traveling Screens / Lowering Condenser 6.# MC05 R-NUSO Vacuum RD09A RD09B Hydraulic Anticipated Transient Without SCRAM 7.# M-ALL RD17A (ATWS)

RD17B C-BOP

8. HS-47-1A 2A EHC Pump Trip C-NUSO C-BOP

9. SL01A SLC Pump Trip C-NUSO

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor (TS)Technical Specification

Event on previous two NRC Exams

S Event on previous two NRC Exams Spare Scenario Unit 2 - Page 1 of 3

Appendix D Scenario Outline Form ES-D1 Events

1. The crew will swap Recirc Drive Cooling Water Pumps in accordance with 2-OI-68, Reactor Recirculation System, Section 6.3.

2. D3 Emergency Equipment Cooling Water (EECW) Pump will trip. The crew will start another pump to support the South EECW Header in accordance with 0-OI-67, Emergency Equipment Cooling Water, Section 5.3. Depending on the course of action taken to restore EECW Flow, the Nuclear Unit Senior Operator (NUSO) will reference either Technical Specification 3.7.1, Residual Heat Removal Service Water (RHRSW) System, Condition A or Technical Specification 3.7.2, Emergency equipment Cooling Water (EECW) System and Ultimate Heat Sink (UHS), Condition A.

3. APRM 1 will fail downscale. The crew will respond in accordance with Alarm Response Procedures and 2-OI-92B, Average Power Range Monitoring to bypass the faulty instrument. The NUSO will review Technical Specification 3.3.1.1, RPS Instrumentation, Table 3.3.1.1.

4. A Control Rod will drift out. The crew will take actions to insert the Control Rod in accordance with 2-AOI-85-6, Rod Drift Out. The drifting Control Rod will latch into position 00 and the NUSO will address Technical Specification 3.1.3, Control Rod OPERABILITY, Condition C.

5. Steam Jet Air Ejector 2A will trip, requiring the crew to respond in accordance with Alarm Response Procedures and 2-OI-66, Off-Gas System to place the standby Steam Jet Air Ejector in service.

6. Due to eel grass intrusion, the intake traveling screens will become clogged, resulting in lowering Condenser Vacuum. The crew will respond to lowering Condenser Vacuum in accordance with 2-AOI-47-3, Loss of Condenser Vacuum. The crew will reduce Reactor Power in response to lowering Condenser Vacuum prior to inserting a manual Reactor SCRAM.

7. When a manual Reactor SCRAM is attempted, Control Rods will not insert due to a Hydraulic Anticipated Transient Without SCRAM (ATWS). The crew will respond in accordance with 2-EOI-1A, ATWS RPV Control.

8. When a Manual Reactor SCRAM is inserted, 2A EHC Pump will trip, resulting in Main Turbine Bypass Valves failing CLOSED, requiring the crew to take action to control Reactor Pressure.

9. When the crew attempts to inject SLC for Reactor Power Control, the first SLC Pump will trip, requiring the Balance of Plant Operator (BOP) to start the alternate SLC Pump in accordance with 2-EOI Appendix-3A, SLC Injection.

The Scenario ends when the crew has inserted all Control Rods and has control of Reactor Water Level above the Top of Active Fuel ((-) 162 inches) using high pressure systems.

Unit 2 - Page 2 of 3

Appendix D Scenario Outline Form ES-D1 Critical Tasks 2

1. With a Reactor SCRAM required and the Reactor not shutdown, to prevent an uncontrolled Reactor depressurization and subsequent power excursion inhibit ADS or control Reactor Water Level such that no automatic ADS actuation occurs.

a. Safety Significance Precludes core damage due to an uncontrolled reactivity addition.

b. Cues Procedural Compliance.

c. Measured by ADS Logic inhibited prior to an automatic initiation.

d. Feedback Reactor Pressure trend.

Reactor Water Level trend.

ADS LOGIC BUS A/B INHIBITED (2-9-3C, Window 18/31) annunciator status.

e. Critical Task Failure Criteria ADS automatic initiation with Control Rods out.

2. When Reactor Power is greater than 5% or unknown during an ATWS, STOP and PREVENT all injection into the Reactor except for RCIC, CRD, and SLC within 130 seconds of the loss of forced recirculation to prevent possible fuel damage.

a. Safety Significance With thermal power being produced in the Reactor and actions to lower Reactor Power have not brought power out of the heating range, power oscillations and subsequent fuel damage may be generated.

b. Cues The Reactor is SCRAMMED and ATWS actions are taken by the OATC and Reactor Power is greater than 5% or unknown.

c. Measured by:

Reactor Power on APRM displays.

d. Feedback Lowering Reactor Water Level.

Lowering Reactor Power.

e. Critical Task Failure Criteria:

The operating crew fails to STOP and PREVENT injection and lower Reactor Water Level for Reactor Power control during an ATWS within 130 seconds after Recirculation Flow is stopped.

Unit 2 - Page 3 of 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-2 Event No.: 1 Page 1 of 2 Event

Description:

Swap Recirc Drive Cooling Water Pumps Time Position Applicants Actions or Behavior Prior to placing the simulator in RUN, start CPERF to record critical Driver parameters.

If the crew does not start Event 1, Swap Recirc Drive Cooling Water Pumps, after assuming the shift, request that the Driver contact the NRC Nuclear Unit Senior Operator (NUSO) as the Shift Manager and direct the crew to swap Recirc Drive Cooling Water Pumps.

If requested by the Chief Examiner, contact the NUSO as the Shift Manager and direct the crew to swap Recirc Drive Cooling Water Driver Pumps.

If contacted as an Assistant Unit Operator (AUO), report that you are standing by, and are ready to support.

Directs the Balance of Plant Operator (BOP) to swap Recirc Drive NUSO Cooling Water Pumps in accordance with 2-OI-68, Reactor Recirculation System.

2-OI-68, Reactor Recirculation System Section 6.3, Swapping Recirc Drive Cooling Water Pumps NOTES

1) Perform these steps, as required, to swap the Recirc Drive Cooling Water Pumps.

2) Placing the standby pump in RUN will cause the running pump to shut down after ~2 seconds if the running pump is in AUTO.

BOP 3) The red light indication above the MCR handswitch only indicates that the motor starter has been energized. A successful pump start should be verified locally or by ICS flow indication.

4) ICS screen VFDPMPA(VFDPMPB) may be referred to observe Recirc Drive Cooling Water System parameters.

5) The time both Cooling Water Pumps are running should be minimized. The pump being placed in standby should be placed in AUTO as soon as possible after placing the lead pump in RUN.

Page 1 of 55 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-2 Event No.: 1 Page 2 of 2 Event

Description:

Swap Recirc Drive Cooling Water Pumps Time Position Applicants Actions or Behavior

[1] IF it is desired to place Recirc Drive Cooling Water Pump 2A2 in service, and place 2A1 pump in standby, THEN PERFORM the following: (Otherwise N/A):

[1.1] DEPRESS 2-HS-96-13, FAULT RESET, on Panel 2-9-4.

[1.2] PLACE in RUN 2-HS-68-2A2/A, RECIRC DRIVE 2A COOLING PUMP 2A2.

[1.3] CHECK RECIRC DRIVE 2A COOLING PUMP 2A2, STARTS.

[1.4] PLACE in AUTO 2-HS-68-2A1/A, RECIRC DRIVE 2A COOLING PUMP 2A1.

[1.5] CHECK RECIRC DRIVE 2A COOLING PUMP 2A1 STOPS.

[2] N/A BOP

[3] IF it is desired to place Recirc Drive Cooling Water Pump 2B2 in service, and place the B1 pump in standby, THEN PERFORM the following: (Otherwise N/A)

[3.1] DEPRESS 2-HS-96-14, FAULT RESET, on Panel 2-9-4.

[3.2] PLACE in RUN 2-HS-68-2B2/A, RECIRC DRIVE 2B COOLING PUMP 2B2.

[3.3] CHECK RECIRC DRIVE 2B COOLING PUMP 2B2 STARTS.

[3.4] PLACE in AUTO 2-HS-68-2B1/A, RECIRC DRIVE 2B COOLING PUMP 2B1.

[3.5] CHECK RECIRC DRIVE 2B COOLING PUMP 2B1 STOPS.

[4] N/A If contacted as an AUO to report the status, report that 2B1 and Driver 2B2 Recirc Drive Cooling Water Pumps are running normally.

Informs the Nuclear Unit Senior Operator (NUSO) that Recirc Drive BOP Cooling Water Pumps have been swapped.

End of Event 1. Request that the driver insert Event 2, D3 EECW NRC Pump Trip.

Page 2 of 55 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-2 Event No.: 2 Page 1 of 3 Event

Description:

EECW Pump Trip Time Position Applicants Actions or Behavior When requested by the Chief Examiner, insert Event 2, EECW Driver Pump Trip to trip D3 EECW Pump.

Acknowledges and reports the following alarms if received to the NUSO:

MOTOR TRIPOUT, 2-9-8C, Window 33 BOP EECW NORTH HEADER DG SECTION PRESSURE LOW, 2-9-20A, Window 21 Informs the NUSO that D3 EECW Pump has tripped.

Directs the BOP to respond in accordance with the appropriate Alarm NUSO Response Procedure.

Alarm Response Procedure, 2-ARP-9-8C MOTOR TRIPOUT, Window 33 Operator Action:

A. CHECK Control Room for white disagreement light illuminated for BOP affected equipment.

B. CLEAR disagreement light.

C. DISPATCH personnel to check:

Relays at associated electrical board Equipment for abnormal conditions Safe-stop locally reset, if necessary If contacted as the Outside NUSO, AUO, or Electrical Maintenance Driver to investigate the trip of D3 EECW Pump, acknowledge any direction given.

D. REFER TO 0-GOI-300-2, Electrical, if relay targets are present or for motor starting limits.

BOP E. REFER TO appropriate Operating Instruction for recovery or realignment of equipment.

Page 3 of 55 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-2 Event No.: 2 Page 2 of 3 Event

Description:

EECW Pump Trip Time Position Applicants Actions or Behavior Alarm Response Procedure, 2-ARP-9-20A EECW NORTH HEADER DG SECTION PRESSURE LOW, Window 21 Operator Action:

A. CHECK indications on Panel 2-9-20.

BOP 1. 0-PI-67-19/2, Unit 1-2 NORTH HEADER PRESSURE.

2. 0-FI-67-3A/2, EECW NORTH HEADER PUMP A FLOW.

3. 0-FI-67-9A/2, EECW NORTH HEADER PUMP C FLOW.

B. CHECK Panel 2-9-3 for status of North Header Pump(s) breaker lights and Pump Motor Amps normal.

C. NOTIFY UNIT SUPERVISOR, Unit 1 and Unit 3.

If contacted as the Unit 1 and/or Unit 3 NUSO, acknowledge any Driver information given.

D. START standby pump for affected header. REFER TO 0-OI-67, Emergency Equipment Cooling Water System.

E. DISPATCH personnel to check affected pump room and header for abnormal conditions.

BOP F. N/A G. N/A H. IF pump failure is cause of alarm, THEN REFER TO Technical Specification 3.7.2, Emergency Equipment Cooling Water (EECW)

System and Ultimate Heat Sink (UHS).

NUSO Directs the BOP to start B3 EECW Pump.

0-OI-67, Emergency Equipment Cooling Water System Precautions and Limitations C. The EECW System is aligned as follows:

BOP 1. At least one RHRSW Pump, assigned to the EECW System, should be running on each header to maintain the header charged at all times. If no pumps are running on a header and header pressure lowers to 0 psig, the header shall be declared inoperable and appropriate actions taken, as required by Technical Specifications.

Page 4 of 55 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-2 Event No.: 2 Page 3 of 3 Event

Description:

EECW Pump Trip Time Position Applicants Actions or Behavior BOP Starts B3 EECW Pump.

References Technical Specification 3.7.2, Emergency Equipment Cooling Water (EECW) System and Ultimate Heat Sink (UHS).

LCO 3.7.2: The EECW System with three pumps and UHS shall be OPERABLE.

NUSO APPLICABILITY: MODES 1, 2, and 3.

CONDITION:

A. One required EECW Pump INOPERABLE.

REQUIRED ACTION: COMPLETION TIME:

NUSO A.1 Restore the required EECW A.1 - 7 days Pump to OPERABLE status.

End of Event 2. Request that the Driver insert Event 3, APRM 1 NRC Fails Downscale.

Page 5 of 55 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-2 Event No.: 3 Page 1 of 3 Event

Description:

APRM 1 Fails Downscale Time Position Applicants Actions or Behavior When requested by the Chief Examiner, insert Event 3, APRM 1 Driver Fails Downscale.

Acknowledges and reports the following alarms:

OATC APRM DOWNSCALE / OPRM INOPERABLE, 2-9-5A, Window 4 CONTROL ROD WITHDRAWAL BLOCK, 2-9-5A, Window 7 Directs the Operator at the Controls (OATC) to respond in accordance NUSO with the appropriate Alarm Response Procedures.

Alarm Response Procedure, 2-ARP-9-5A APRM DOWNSCALE / OPRM INOP, Window 4 Operator Action:

A. DETERMINE which APRM/OPRM channel is downscale/inoperable.

B. IF APRM failed downscale, THEN BYPASS channel. REFER TO OATC 2-OI-92B, Average Power Range Monitoring.

C. N/A D. N/A E. REFER TO Technical Specification (Tech Spec) Tables 3.3.1.1-1, Reactor Protection System Instrumentation, and Technical Requirements Manual (TRM) Table 3.3.4-1, Control Rod Block Instrumentation.

OATC Recommends to the NUSO that APRM 1 be bypassed.

EXAMINER NOTE: The NUSO may reference Tech Specs and the NRC TRM. There are no required Tech Spec or TRM actions, but the NUSO may enter an Information Only LCO.

Directs the OATC to bypass APRM 1 in accordance with the appropriate NUSO Operating Instruction.

Page 6 of 55 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-2 Event No.: 3 Page 2 of 3 Event

Description:

APRM 1 Fails Downscale Time Position Applicants Actions or Behavior 2-OI-92B, Average Power Range Monitoring Section 6.0, System Operations NOTES

1) Only one APRM/OPRM can be bypassed at a time.

2) All operations are performed on Panel 2-9-5 unless specifically stated otherwise.

OATC 3) In order to prevent inadvertent Rod Withdrawal Block or Reactor SCRAM while operating APRM BYPASS Selector Switch, always ensure the previously bypassed channel returns to normal status by observing the BLUE bypassed lights on Panel 2-9-14 Voters are extinguished prior to selecting any other channel to be bypassed.

After bypassing a channel, the applicable BLUE BYPASSED status lights on Panel 2-9-14 Voters should be illuminated prior to testing, operating, or working on that channel.

Section 6.1, Bypassing APRM / OPRM Channel CAUTION NPG-SPP-10.4, Reactivity Management Program, requires approval of the Plant Manager or his designee prior to any planned operation with APRMs bypassed unless bypassing is specifically allowed within OATC approved procedures.

[1] REVIEW all Precautions and Limitations. REFER TO Section 3.0.

[2] PLACE 2-HS-92-7B/S3, APRM BYPASS, to desired channel to be bypassed.

[3] CHECK BLUE BYPASSED lights illuminated on Panel 2-9-14 Voters.

[4] CHECK white bypass light on Panel 2-9-5 is illuminated.

Page 7 of 55 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-2 Event No.: 3 Page 3 of 3 Event

Description:

APRM 1 Fails Downscale Time Position Applicants Actions or Behavior Alarm Response Procedure, 2-ARP-9-5A CONTROL ROD WITHDRAWAL BLOCK, Window 7 Operator Action:

A. DETERMINE initiating condition from corresponding rod withdrawal OATC block alarm(s) and REFER TO Operator Action for alarm(s).

B. N/A C. N/A D. N/A E. N/A End of Event 3. Request that the driver insert Event 4, Control Rod NRC Drifts Out.

Page 8 of 55 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-2 Event No.: 4 Page 1 of 5 Event

Description:

Control Rod Drifts Out Time Position Applicants Actions or Behavior When requested by the Chief Examiner, insert Event 4, Control Rod Driver Drift Out.

NRC Control Rod 30-11 will drift out.

Acknowledges and reports the following alarm to the NUSO:

CONTROL ROD DRIFT, 2-9-5A, Window 28 OATC CONTROL ROD WITHDRAWL BLOCK, 2-9-5A, Window 7 RBM DOWNSCALE, 2-9-5A, Window 31 Directs the OATC to respond in accordance with the appropriate Alarm NUSO Response and Abnormal Operating Procedures.

Alarm Response Procedure, 2-ARP-9-5A CONTROL ROD DRIFT, Window 28 Operator Action:

A. DETERMINE which rod is drifting from Full Core Display.

B. IF no Control Rod motion is observed, THEN RESET rod drift as follows.

OATC 1. PLACE 2-HS-85-3A-S7, ROD DRIFT ALARM TEST switch, in RESET and RELEASE.

2. RESET annunciator.

C. N/A D. IF rod drifting out, THEN REFER TO 2-AOI-85-6, Rod Drift Out and 2-AOI-85-7, Mispositioned Control Rod.

E. REFER TO Tech Spec 3.1.3, Control Rod Operability and 3.10.8, Shutdown Margin (SDM) Test - Refueling.

The Control Rod Drift condition will clear when the Control Rod is NRC driven to Position 0.

2-AOI-85-6, Rod Drift Out OATC Immediate Actions:

[1] N/A Page 9 of 55 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-2 Event No.: 4 Page 2 of 5 Event

Description:

Control Rod Drifts Out Time Position Applicants Actions or Behavior Subsequent Actions:

[1] N/A

[2] IF a Control Rod is moving from its intended position without operator actions, THEN SELECT the drifting Control Rod and INSERT to the FULL IN (00) position.

OATC

[3] IF a Control Rod Block occurs during rod insertion due to Rod Worth Minimizer, THEN BYPASS the RWM per step 4.2[1]. (Otherwise N/A)

[4] N/A

[5] NOTIFY the Reactor Engineer to Evaluate Core Thermal Limits and Preconditioning Limits for the current Control Rod pattern.

When contacted as the Reactor Engineer or an AUO acknowledge Driver any direction given.

[6] IF another Control Rod Drift occurs before Reactor Engineering completes the evaluation, THEN MANUALLY SCRAM Reactor and enter 2-AOI-100-1, Reactor SCRAM.

[7] N/A

[8] IF the Control Rod is latched into position 00, THEN REMOVE associated Hydraulic Control Unit (HCU) from service per 2-OI-85, Control Rod Drive System. (N/A if Control will not latch at "00")

[9] EVALUATE Tech Spec 3.1.3, Control Rod OPERABILITY.

[10] INITIATE Service Request/Work Order.

OATC [11] NOTIFY Reactor Engineer to perform the following:

EVALUATE condition of the Core to assure no resultant fuel damage has occurred, and EVALUATION of impact on Thermal Limits and PCIOMOR restraints. (N/A if SCRAM was initiated.)

DETERMINE if other Control Rods need to be repositioned in order to safely restore Core symmetry to prevent local fuel damage. (N/A if scram was initiated.)

[12] NOTIFY System Engineering to PERFORM 0-TI-20, Control Rod Drive System Testing and Troubleshooting, to determine problem with faulty Control Rod.

Page 10 of 55 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-2 Event No.: 4 Page 3 of 5 Event

Description:

Control Rod Drifts Out Time Position Applicants Actions or Behavior

[13] IF a manual SCRAM was not inserted and Reactor Startup or Shutdown is not in progress, THEN ENSURE 2-GOI-100-12, Power Maneuvering, has been entered if a power change occurred.

(Otherwise N/A)

OATC

[14] - [15] N/A

[16] NOTIFY the Reactor Engineer to EVALUATE impact on preconditioning envelope, prior to returning to normal power operation.

2-AOI-85-7, Mispositioned Control Rod (if entered)

Immediate Actions: None Subsequent Actions:

[1] N/A

[2] IF Control Rod is determined to be mispositioned, THEN NOTIFY the following:

Reactor Engineer (RE)

Shift Technical Advisor (STA)

Unit Supervisor OATC Shift Manager (SM)

Operations Superintendent

[3] N/A

[4] N/A

[5] CHECK the following radiation recorders for rise in radiation activity to determine if fuel damage occurred:

MAIN STEAM LINE RADIATION, 2-RR-90-135 (Panel 2-9-2)

OFFGAS RADIATION, 2-RR-90-266 (Panel 2-9-2)

OG PRETREATMENT RAD MON RTMR, 2-RM-90-157, on Panel 2-9-10 OFFGAS RAD MON RTMR, 2-RM-90-160, on Panel 2-9-10

[6] N/A Page 11 of 55 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-2 Event No.: 4 Page 4 of 5 Event

Description:

Control Rod Drifts Out Time Position Applicants Actions or Behavior

[7] INTIATE a Service Request/PER for Control Rod error or mispositioned Control Rod.

[8] N/A

[9] NOTIFY Reactor Engineer to perform the following when time OATC permits:

EVALUATE possible consequences DOCUMENT in Reactor Engineer log

[10] N/A Technical Specification 3.1.3, Control Rod OPERABILITY LCO 3.1.3 Each Control Rod shall be OPERABLE Applicability: Modes 1 and 2 NUSO ---------------------------------------------------------------------------------------------

NOTE: Separate Condition entry is allowed for each Control Rod.

CONDITION:

C. One or more Control Rods INOPERABLE for reasons other than Condition A or B.

REQUIRED ACTION: COMPLETION TIME:

C.1 --------------NOTE------------------ C.1 - 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months RWM may be bypassed as allowed by LCO 3.3.2.1, if required, to allow insertion of INOPERABLE Control Rod NUSO and continued operation.

Fully insert INOPERABLE Control Rod AND C.2 Disarm the associated CRD. C.2 - 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Page 12 of 55 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-2 Event No.: 4 Page 5 of 5 Event

Description:

Control Rod Drifts Out Time Position Applicants Actions or Behavior Tech Spec 3.10.8, Shutdown Margin (SDM) Test - Refueling is NRC not applicable to current plant conditions.

End of Event 4. Request that the Driver insert Event 5, 2A Steam NRC Jet Air Ejector (SJAE) Trip.

Page 13 of 55 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-2 Event No.: 5 Page 1 of 3 Event

Description:

2A Steam Jet Air Ejector (SJAE) Trip Time Position Applicants Actions or Behavior When requested by the Chief Examiner, insert Event 5, Steam Jet Driver Air Ejector (SJAE) Trip.

Acknowledges and reports the following alarms to the NUSO:

OFFGAS HOLDUP LINE INLET FLOW LOW, 2-9-53, Window 4 BOP H2 WATER CHEMISTRY ABNORMAL, 2-9-53, Window 10 H2 WATER CHEMISTRY SHUTDOWN, 2-9-53, Window 20 Alarm Response Procedure, 2-9-53 OFFGAS HOLDUP LINE INLET FLOW LOW, Window 4 A. CHECK holdup line flow on:

1. OFFGAS H2 CONTROL/OFFGAS TO 6 HR FLOW/ANALYZER, 2-XR-66-103 (CH 3-Narrow Range), Panel 2-9-53.

2. OFFGAS FLOW TO HOLDUP VOLUME, 2-FR-66-20, Panel 2-9-8.

BOP 3. Integrated Computer System, computer point 66-111A (D419).

B. ENSURE OPEN in-service Recombiner Inlet and Outlet Valves at Panel 2-9-53.

C. ENSURE OPEN, OFF-GAS SYSTEM ISOLATION VALVE, 2-FCV-66-28.

D. ENSURE that SJAE auto isolation has NOT occurred.

E. DISPATCH personnel to Panel 25-95 to verify the alarm, 2-FI-66-111A.

F. N/A Determines that the cause for the above alarms is the trip of 2A SJAE.

BOP Informs the NUSO.

Directs the BOP to place the Standby (3B) SJAE in service using the NUSO 2-OI-66, Off-Gas System, Attachment 6, Standby SJAE System Lineup Hard Card.

OATC Monitors Condenser Vacuum.

Page 14 of 55 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-2 Event No.: 5 Page 2 of 3 Event

Description:

2A Steam Jet Air Ejector (SJAE) Trip Time Position Applicants Actions or Behavior Driver If contacted as an AUO, acknowledge any direction given.

2-OI-66, Off-Gas System Attachment 6, Standby SJAE System Lineup Hard Card Operator Actions for SJAE 2B NOTES Radiation Protection should be notified prior to placing a SJAE in service. If time does not permit this due to plant conditions then notification should be made when possible.

2-HS-001-0375, SJAE TRAIN PERMISSIVE (located on 2-LPNL-925-0105, U2 TB, EL 586) should normally be in the position of the standby SJAE. If problems are encountered while placing a SJAE in service and time permits, operate this switch as required during the performance of this section.

BOP [1] ENSURE RESET Off-Gas isolation using 2-HS-90-155, OFFGAS OUTLET/DRAIN ISOLATION VALVES.

[2] ENSURE OPEN the following valves:

2-HS-66-15, SJAE 2B INLET VALVE 2-HS-1-156A, STEAM TO SJAE 2B

[3] ENSURE in AUTO/OPEN SJAE 2B OFFGAS OUTLET VALVE, 2-HS-66-18.

[4] PLACE 2-HS-1-152, SJAE 2B PRESS CONTROLLER, in CLOSE and then in OPEN.

[5] ENSURE OPEN the following valves (red light illuminated):

STEAM TO SJAE 2B STAGES 1,2, AND 3, 2-PCV-1-153/167 SJAE 2B INTERMEDIATE CONDENSER DRAIN,2-FCV-1-152

[6] MONITOR Hotwell Pressure as indicated on recorder CONDENSATE, 2-XR-002-0026 (Point 3), on Panel 2-9-6.

If notified as Radiation Protection that 2B SJAE is being placed Driver (or is) in service, acknowledge the report.

Page 15 of 55 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-2 Event No.: 5 Page 3 of 3 Event

Description:

2A Steam Jet Air Ejector (SJAE) Trip Time Position Applicants Actions or Behavior

[7] FOR the SJAE not being placed in service, ENSURE CLOSED the following valves:

SJAE 2A OG OUTLET VALVE, 2-HS-66-14 SJAE 2A PRESSURE CONTROLLER, 2-HS-1-150 BOP STEAM TO SJAE 2A, 2-HS-1-155A

[8] WHEN Condenser Vacuum has stabilized, THEN PROCEED to step 8.4[4] AND PERFORM the applicable steps to establish acceptable SJAE parameters to obtain stable and live pressure indications.

Informs the NUSO that 2B SJAE is in service. Monitors Condenser BOP Vacuum.

If contacted as the Work Control NUSO, Maintenance, or an AUO to Driver investigate the trip of 2A SJAE, acknowledge any direction given.

End of Event 5. Request that the driver insert Event 6, Clogged NRC Traveling Screens / Lowering Condenser Vacuum.

Page 16 of 55 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-2 Event No.: 6 Page 1 of 11 Event

Description:

Clogged Traveling Screens / Lowering Condenser Vacuum Time Position Applicants Actions or Behavior When requested by the Chief Examiner, insert Event 6, Clogged Driver Traveling Screens / Lowering Condenser Vacuum.

Acknowledges and reports the following alarm to the NUSO:

BOP TRAVELING SCEEN DP HIGH, 2-9-20A, Window 18 Directs the Balance of Plant Operator (BOP) to respond in accordance NUSO with the appropriate Alarm Response Procedure.

Alarm Response Procedure, 2-ARP-9-20A TRAVELING SCEEN DP HIGH, Window 18 Operator Action:

A. CHECK alarm on 2-PDI-27-1A, TRAVELING SCREEN DIFFERENTIAL WATER LEVEL, on Panel 2-9-20.

B. DISPATCH personnel to VERIFY the traveling screens are in service.

Refer to 2-OI-27A, Screen Wash System.

C. MONITOR Traveling Screens for carryover.

D. MONITOR Turbine Backpressure.

E. IF debris is being carried over, THEN BOP MONITOR 0-PDIS-067-0001(0005)(0008)(0011), EECW SUPPLY STRAINER DIFF PRESS, locally in RHRSW Pump Rooms MONITOR Waterbox D/P for indications of fouling (<160 H2O (does not apply to 2C2 waterbox) with 3 CCW pumps in service)

F. IF TRAVELING SCREEN DIFF WTR LVL, 2-PDI-27-1A, does NOT lower, THEN REFER TO 2-OI-27A, Screen Wash System REFER TO 0-AOI-27-1, Component Biofouling NOTIFY Mechanical Maintenance to SCRAPE the trash racks and/or operate Milfoil Harvester as needed Page 17 of 55 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-2 Event No.: 6 Page 2 of 11 Event

Description:

Clogged Traveling Screens / Lowering Condenser Vacuum Time Position Applicants Actions or Behavior G. IF Divers are required to clear the trash racks, THEN REMOVE the BOP Amertap system from service per 2-OI-27B, Amertap Condenser Tube Cleaning System.

0-AOI-27-1, Component Biofouling Immediate Actions: NONE Subsequent Actions:

NOTES

1) Procedure is written in a logical order but due to changing plant conditions and operator experience, steps may be performed in parallel or out of sequence as required.

2) The most common cause of degraded cooling water system performance is the fouling of intake screens for the CCW Pumps.

When intake screens begin to foul they are required to be placed in service as soon as possible using 1(2,3)-OI-27A, Screen Wash System. The timely response to this condition will keep screens from becoming over burdened with foreign material and collapsing.

3) If CCW Intake Screens cannot be cleaned with associated CCW BOP Pump running, the pump may have to be removed from service in order to clean screens. After screens are cleaned the affected CCW Pump may be returned to service (1(2,3)-OI-27, Condenser Circulating Water System) if desired.

4) Entry into this procedure requires evaluation of situation per EOOS Management NPG-SPP-09.11.1, Equipment Out of Service Management.

CAUTION Debris Filter may cycle repeatedly when total CCW flow is throttled excessively or due to heavy debris carryover from the intake. Debris Filter Flush Valve Motor is not rated for heavy repeated cycling. CCW flow should be maximized by throttling open available waterboxes until pump head is as low as possible (>20 H2O), Debris Filter should be run in manual and checked often. Expedite returning CCW System to normal three pump alignment with Debris Filter in AUTO.

Page 18 of 55 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-2 Event No.: 6 Page 3 of 11 Event

Description:

Clogged Traveling Screens / Lowering Condenser Vacuum Time Position Applicants Actions or Behavior BOP [1] CONTACT Maintenance to PERFORM attachment 7.

When contacted as Maintenance to perform Attachment 7, Driver acknowledge the direction.

[2] CHECK CCW Intake Screens for fouling.

BOP [3] IF CCW Intake Screens are fouled, THEN ENSURE in service per 2-OI-27A, Screen Wash System. (Otherwise N/A).

If contacted as the Outside NUSO, Maintenance, or an AUO to check intake screens for fouling, acknowledge the direction. Wait 2 minutes and report that the intake screens are becoming fouled Driver by Eel Grass.

Is asked if Eel Gras is coming over the intake screens, state that it is.

[3] IF CCW Intake Screens are fouled, THEN ENSURE in service per BOP 2-OI-27A, Screen Wash System. (Otherwise N/A).

[4] INITIATE Attachment 1, Continuous Action Summary.

0-AOI-27-1, Component Biofouling Attachment 1, Continuous Action Summary Action Summary

[1] IF at any time any of the following condition occurs:

Unexpected fouling indication of more than one river water supplied heat exchanger.

NUSO / THEN Action II is applicable, GO TO Step 4.2[9]. (Otherwise N/A)

BOP [2] IF at any time any of the following conditions occur:

Any indications of abnormal operation of the Circulating Water System OR Any removal of two or more Circulating Water Pumps from service THEN GO TO 1(2)(3)-OI-27, Circulating Water System (Otherwise N/A)

Page 19 of 55 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-2 Event No.: 6 Page 4 of 11 Event

Description:

Clogged Traveling Screens / Lowering Condenser Vacuum Time Position Applicants Actions or Behavior

[3] IF at any time a low reservoir level (<550) OR High Traveling Screen DP occur THEN: (Otherwise N/A)

MONITOR CCW Pumps for loss of suction/cavitation.

BOP [4] IF at any time the CCW Pumps indicate a loss of suction or cavitation THEN: (Otherwise N/A)

SHUTDOWN the CCW Pumps. REFER TO 2-OI-27, Circulating Water System.

As the malfunction ramps in, Condenser Vacuum will continue to NRC lower. The crew may enter 2-AOI-47-3, Loss of Condenser Vacuum before an alarm is received.

2-AOI-47-3, Loss of Condenser Vacuum Immediate Actions: NONE Subsequent Actions:

[1] IF ANY EOI entry condition is met, THEN ENTER the appropriate EOI(s).

CAUTION Operations outside of the allowable regions shown on the Recirculation System Operating Map could result in BOP thermal-hydraulic power oscillations and subsequent fuel damage.

REFER TO 2-GOI-100-12A, Unit Shutdown from Power Operation to Cold Shutdown and Reductions in Power During Power Operations, for required actions and monitoring to be performed during a power reduction.

[2] MONITOR Condenser Vacuum (Turb Exhaust) Margin to Trip using 2-XR-002-0026, CONDENSATE, Channel 7.

[3] IF Condenser Vacuum (Turb Exhaust) Margin to Trip as indicated on 2-XR-002-0026, CONDENSATE, approaches 0 inches Hg, with Reactor Power less than 26%, THEN TRIP the Main Turbine.

Page 20 of 55 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-2 Event No.: 6 Page 5 of 11 Event

Description:

Clogged Traveling Screens / Lowering Condenser Vacuum Time Position Applicants Actions or Behavior

[4] IF Condenser Vacuum is lost, THEN OPEN 2-DRV-043-1019, HOTWELL SAMPLE TO FLOOR DRAIN, (557@ T-10 C-Line) and BOP 2-DRV-043-1020, CONDENSATE DEMIN SAMPLE TO FLOOR DRAIN, (557@ T-6 G-Line), to establish flow through the sample lines.

If contacted as an AUO to perform any outside steps of this Driver procedure, acknowledge any direction given.

[5] REDUCE Reactor Power in an attempt to maintain Condenser BOP Vacuum.

See Event 6, Attachment 1 (page 23), Reactor Power Reduction NRC for Lowering Condenser Vacuum for Reactor Power reduction actions.

[6] ENSURE automatic actions.

[7] CHECK CCW Pumps for proper operation. IF available, THEN BOP START additional CCW Pumps.

[8] - [15] N/A NRC The crew may elect to NOT start 2A CCW Pump.

Directs the crew to monitor Condenser Vacuum and sets a trigger NUSO value of 1.0 to Trip Condenser Vacuum for a Reactor SCRAM.

In accordance with the ILT Simulator Expectations, the NUSO will set target values for Condenser Vacuum as follows:

Low Condenser Vacuum alarm (Panel 2-9-7B, Window 17) - Core Flow Runback NRC 1 Margin to Trip Condenser Vacuum - Reactor SCRAM However, the NUSO may conservatively direct these actions based on the rate of vacuum degradation and information from the field.

Acknowledges and reports the following alarm when received:

BOP CONDENSER A, B, OR C VACUUM LOW, 2-9-7B, Window 17 Page 21 of 55 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-2 Event No.: 6 Page 6 of 11 Event

Description:

Clogged Traveling Screens / Lowering Condenser Vacuum Time Position Applicants Actions or Behavior Alarm Response Procedure, 2-ARP-9-7B CONDENSER A, B, OR C VACUUM LOW, 2-9-7B, Window 17 Operator Action:

BOP A. CHECK alarm by checking Condenser Vacuum lowering, MWe lowering, and Exhaust Hood Temperature rise.

B. IF alarm is valid, THEN REFER TO 2-AOI-47-3, Loss of Condenser Vacuum.

Provides a Condenser Vacuum trigger value for the OATC to insert a NUSO manual Reactor SCRAM.

When the trigger value for Condenser Vacuum is reached or when OATC directed by the NUSO, inserts a manual Reactor SCRAM.

Condenser Vacuum will continue to lower, requiring the crew to insert a manual Reactor SCRAM.

NRC Two minutes after the crew inserts a manual SCRAM, the clogged Traveling Screen malfunction will be set to 50% to provide cooling water for the crew to place Suppression Pool Cooling in service if necessary.

Two minutes after the Reactor SCRAM report to the crew that Driver there has been some success in clearing Traveling Screens to allow some water flow.

When the crew inserts a manual Reactor SCRAM due to lowering Condenser Vacuum, the following Events are automatically inserted by Simulator Setup:

Event 7, Hydraulic Anticipated Transient Without SCRAM NRC (ATWS)

Event 8, 2A EHC Pump Trip Event 9, SLC Pump Trip No action is required by the Driver to insert Events 7, 8, or 9.

End of Event 6, proceed to Event 7.

Page 22 of 55 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-2 Event No.: 6 Page 7 of 11 Event 6, ATTACHMENT 1 Reactor Power Reduction for Lowering Condenser Vacuum for Reactor Power Time Position Applicants Actions or Behavior The crew may elect to reduce power by any method or combination NRC of runbacks in an attempt to maintain Condenser Vacuum.

Directs the OATC to reduce Reactor Power by using any one or any combination of the following:

Manually using Recirc Master Control pushbuttons on NUSO Panel 2-9-5 Upper Power Runback Mid Power Runback Core Flow Runback 2-OI-68, Reactor Recirculation System 3.0 Precautions and Limitations Section 3.5.3, Dual Pump Operation E. When raising (lowering) Reactor Power per the Reactivity Plan, the following guideline should be used to establish the 60 rpm mismatch between the Recirc Pumps.

1. When the first Recirc Pump reaches 1200 rpm (1300 rpm) or both Recirc Pumps are at 1200 rpm (1300 rpm) then individual controls will be used.

2. While following the Reactivity Plan establish the 60 rpm NRC mismatch using the individual controls for the leading Recirc Pump.

3. While maintaining the 60 rpm mismatch and using the Reactivity Plan, raise (lower) the Recirc Pump speeds using either the Master Controllers or Individual Controllers.

4. Once a Recirc Pump reaches 1300 rpm (1200 rpm) the 60 rpm mismatch is no longer required. While following the Reactivity Plan raise (lower) the lagging Recirc Pump using the individual controller match Recirc Pump speeds.

5. Once the Recirc Pumps are matched, then the speeds may be adjusted as required using the Master Controller or Individual Controllers while maintaining the requirements of Attachment 1.

Page 23 of 55 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-2 Event No.: 6 Page 8 of 11 Event 6, ATTACHMENT 1 Reactor Power Reduction for Lowering Condenser Vacuum for Reactor Power Time Position Applicants Actions or Behavior NRC Page 24 of 55 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-2 Event No.: 6 Page 9 of 11 Event 6, ATTACHMENT 1 Reactor Power Reduction for Lowering Condenser Vacuum for Reactor Power Time Position Applicants Actions or Behavior 2-OI-68, Reactor Recirculation System Section 6.2, Adjusting Recirc Flow NOTES

1) Thermal Limits are shown in 0-TI-248 and 2-SR-2, Instrument Checks and Observations.

2) Recirc Flow changes made during the later part of the operating cycle (Coastdown) could cause Core Flow values to approach or exceed the allowable values of the Increased Core Flow (ICF)

Region of the Power to Flow Map. Instruments used to monitor pump speed and Core Flow should be identified in the Reactivity Control Plan. These values should be recorded prior to reducing Core Flow and used as a benchmark to reestablish the previous conditions when returning to power. Increased caution should be used when changes in Recirc Flow are made in this area.

OATC

[1] IF desired to control Recirc Pumps 2A and/or 2B speed with Recirc Individual Control, THEN PERFORM the following:

LOWER Recirc Pump 2A using 2-HS-96-17A(17B)(17C),

SLOW(MEDIUM)(FAST), (Otherwise N/A)

AND/OR LOWER Recirc Pump 2B using 2-HS-96-18A(18B)(18C),

SLOW(MEDIUM)(FAST). (Otherwise N/A)

[2] WHEN desired to control Recirc Pumps 2A and/or 2B speed with the RECIRC MASTER CONTROL, THEN ADJUST Recirc Pump Speed 2A & 2B using the following pushbuttons as required.

2-HS-96-33, LOWER SLOW 2-HS-96-34, LOWER MEDIUM 2-HS-96-35, LOWER FAST Page 25 of 55 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-2 Event No.: 6 Page 10 of 11 Event 6, ATTACHMENT 1 Reactor Power Reduction for Lowering Condenser Vacuum for Reactor Power Time Position Applicants Actions or Behavior 2-OI-68, Reactor Recirculation System Section 8.12, Initiating Manual Runbacks NOTES

1) Manual runback controls are utilized when it becomes necessary to reduce Reactor Power and Core Flow during abnormal plant conditions.

2) This section is performed at Panel 2-9-5.

3) Depressing a manual runback pushbutton initiates a runback of both Recirc Pumps until the setpoint is reached. Depressing the pushbutton a second time stops the manual runback. The pushbutton can be depressed a third and fourth time to reinitiate and stop the manual runback. This pattern can be repeated until the applicable setpoint is reached.

OATC

4) Attachment 2 can be referred to for additional information on manual runback controls.

5) When initiating manual runbacks, the appropriate manual pushbutton must be depressed until the backlight is blinking, then the pushbutton can be released.

6) If 25 rpm mismatch in the lower direction exists between Speed Demand and Calculated Speed, the Manual Runback pushbuttons are disabled.

7) RECIRC PUMPS MID POWER RUNBACK is to be used any time a Condensate Pump trips and Reactor Power is greater than or equal to 90%.

[1] IF time permits, THEN REVIEW Precautions and Limitations.

(REFER TO Section 3.0).

Page 26 of 55 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-2 Event No.: 6 Page 11 of 11 Event 6, ATTACHMENT 1 Reactor Power Reduction for Lowering Condenser Vacuum for Reactor Power Time Position Applicants Actions or Behavior

[2] IF desired to reduce Reactor Power to approximately 90%, THEN (Otherwise N/A):

[2.1] DEPRESS 2-HS-68-42, RECIRC PUMPS UPPER POWER RUNBACK Pushbutton.

[2.2] CHECK the following:

Pushbutton backlight blinks until setpoint is reached Reactor Power lowers to approximately 90%

[3] IF desired to reduce Reactor Power to 66.3%, THEN (Otherwise N/A):

[3.1] DEPRESS 2-HS-68-43, RECIRC PUMPS MID POWER OATC RUNBACK pushbutton.

[3.2] CHECK the following:

Pushbutton backlight blinks until setpoint is reached Reactor Power lowers to 66.3%

[4] IF desired to reduce Core Flow to approximately 60%, THEN (Otherwise N/A):

[4.1] DEPRESS 2-HS-68-44, RECIRC PUMPS CORE FLOW RUNBACK Pushbutton.

[4.2] CHECK the following:

Pushbutton backlight blinks until setpoint is reached Core Flow lowers to approximately 60%

End of Reactor Power Reduction Actions. Continue with Event 6 NRC actions until the Reactor SCRAM is inserted.

Page 27 of 55 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-2 Event No.: 7 Page 1 of 13 Event

Description:

Hydraulic Anticipated Transient Without SCRAM (ATWS)

Time Position Applicants Actions or Behavior Event 7, Hydraulic Anticipated Transient Without SCRAM (ATWS) is NRC inserted on Simulator Setup. No action is required by the Driver to insert Event 7.

During Event 7, when contacted as the Outside NUSO acknowledge direction to perform EOI Appendices and enter events as necessary:

Event 17 EOI-Appendix-1D, Insert Control Rods Using Reactor Manual Control System (Close 2-FCV-85-586, CHARGING WATER ISOLATION)

Event 18 - Open 2-FCV-85-586, CHARGING WATER ISOLATION Driver Event 19 EOI-Appendix-1F, Manual SCRAM Event 20 EOI-Appendix-2, Defeating ARI Logic Trips Event 21 EOI-Appendix-8A, Bypassing Group RPV Low Low Low Level Isolation Interlocks Event 22 EOI-Appendix-8E, Bypassing Group 6 RPV Low Level and High Drywell Pressure Isolation Interlocks Once the event(s) requested have finished their time delay, report completion of the various EOI Appendices to the Control Room.

NUSO Enters 2-EOI-1A, ATWS RPV Control, and updates the crew.

ARC/L-1 ENSURE each as required:

NUSO PCIS isolations (Groups 1, 2, and 3)

ECCS RCIC ARC/L-2 INHIBIT ADS Page 28 of 55 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-2 Event No.: 7 Page 2 of 13 Event

Description:

Hydraulic Anticipated Transient Without SCRAM (ATWS)

Time Position Applicants Actions or Behavior Critical Task:

With a Reactor SCRAM required and the Reactor not shutdown, to prevent an uncontrolled Reactor depressurization and subsequent power excursion inhibit ADS or control Reactor Water Level such CREW that no automatic ADS actuation occurs.

Critical Task Failure Criteria:

ADS automatic initiation with Control Rods out.

ARC/L-3 IF ANY Main Steam Line is open THEN START defeating the following isolations:

NUSO MSIV Low Low Low RPV Water Level (2- EOI-Appendix 8A)

Reactor Building Ventilation Low RPV Water Level (2-EOI-Appendix-8E)

ARC/L-4 IF THEN Reactor Power is above 5% or unknown NUSO AND Reactor Water Level is above

(-) 50 inches ALL Level/Power conditions exist (Table Q-1)

NUSO Page 29 of 55 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-2 Event No.: 7 Page 3 of 13 Event

Description:

Hydraulic Anticipated Transient Without SCRAM (ATWS)

Time Position Applicants Actions or Behavior NUSO NOTE: The start time for the Critical Task below begins when NRC Recirc Pump Speeds are equal to zero.

Critical Task:

When Reactor Power is greater than 5% or unknown during an ATWS, STOP and PREVENT all injection into the Reactor except for RCIC, CRD, and SLC within 130 seconds of the loss of forced recirculation to prevent possible fuel damage.

Crew Critical Task Failure Criteria:

The operating crew fails to STOP and PREVENT injection and lower Reactor Water Level for Reactor Power control during an ATWS within 130 seconds after Recirculation Flow is stopped.

NUSO Page 30 of 55 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-2 Event No.: 7 Page 4 of 13 Event

Description:

Hydraulic Anticipated Transient Without SCRAM (ATWS)

Time Position Applicants Actions or Behavior NUSO NUSO NUSO Page 31 of 55 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-2 Event No.: 7 Page 5 of 13 Event

Description:

Hydraulic Anticipated Transient Without SCRAM (ATWS)

Time Position Applicants Actions or Behavior ARC/L-6 USE ANY Preferred ATWS Injection System (Table L-3) to maintain RPV Water Level between (-) 180 inches and:

Lowered level (if level was deliberately lowered in flowpath A)

OR

+51 inches (if level was NOT deliberately lowered)

Ok to use Core Spray (2-EOI-Appendix-6D or 6E) or NUSO Alternate Injection Subsystems (Table L-2) if previously required by flowpath E or C4A IF THEN Reactor Water Level CANNOT be restored and maintained above

(-) 180 inches NO ACTION REQUIRED AND Core Steam Flow remains below MCSF (Table L-5)

NUSO NUSO Page 32 of 55 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-2 Event No.: 7 Page 6 of 13 Event

Description:

Hydraulic Anticipated Transient Without SCRAM (ATWS)

Time Position Applicants Actions or Behavior NUSO NUSO NUSO Page 33 of 55 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-2 Event No.: 7 Page 7 of 13 Event

Description:

Hydraulic Anticipated Transient Without SCRAM (ATWS)

Time Position Applicants Actions or Behavior ARC/Q-1 IF THEN The Reactor is subcritical AND NO boron has been injected ARC/Q-2 ENSURE Reactor Mode Switch in SHUTDOWN NUSO ARC/Q-3 INITIATE ARI ARC/Q-4 IF tripping Recirc Pumps will cause loss of Main Turbine, RFPT, HPCI, or RCIC THEN ENSURE Recirc Runback (pump speed 480 RPM or less)

ARC/Q-5 IF Reactor Power is above 5% or unknown THEN TRIP Recirc Pumps NUSO In accordance with BFN-ODM-4.20, Strategies for Successful Transient Mitigation, Section 4.8.4.C, when EOI-1A, ATWS RPV Control, Step NUSO ARC/Q-8 is reached, IF Reactor Power is greater than APRM downscale, THEN INITIATE SLC.

Page 34 of 55 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-2 Event No.: 7 Page 8 of 13 Event

Description:

Hydraulic Anticipated Transient Without SCRAM (ATWS)

Time Position Applicants Actions or Behavior NUSO NRC See Event 9 on page 48 of 55.

ARC/Q-10

1. INITIATE SLC (2-EOI-Appendix-3A)

2. INHIBIT ADS IF THEN NUSO Boron CANNOT be injected using INJECT boron into RPV using SLC CRD (2-EOI-Appendix-3B)

SLC Tank Water Level drops to NO ACTION REQUIRED 0%

ARC/Q-11 NUSO ENSURE RWCU System Isolation NUSO Page 35 of 55 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-2 Event No.: 7 Page 9 of 13 Event

Description:

Hydraulic Anticipated Transient Without SCRAM (ATWS)

Time Position Applicants Actions or Behavior NUSO NUSO ARC/P-1 IF THEN A high Drywell Pressure ECCS NO ACTION REQUIRED NUSO signal exists (2.45 PSIG)

EMERGENCY DEPRESSURIZATION is required NO ACTION REQUIRED or has been required Page 36 of 55 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-2 Event No.: 7 Page 10 of 13 Event

Description:

Hydraulic Anticipated Transient Without SCRAM (ATWS)

Time Position Applicants Actions or Behavior ARC/P-2 IF ANY MSRV is cycling NUSO THEN MANUALLY OPEN MSRVs UNTIL RPV press drops to the pressure at which all Main Turbine Bypass Valves are fully open (APPX 11A)

ARC/P-3 IF THEN Suppression Pool Temperature and Water Level CANNOT be maintained in the safe area of NO ACTION REQUIRED Curve 3 at the existing RPV Pressure Suppression Pool Water Level NUSO CANNOT be maintained in the NO ACTION REQUIRED safe area of Curve 4 AND The Main Condenser is available NO ACTION REQUIRED AND There has been no indication of a steam line break Page 37 of 55 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-2 Event No.: 7 Page 11 of 13 Event

Description:

Hydraulic Anticipated Transient Without SCRAM (ATWS)

Time Position Applicants Actions or Behavior ARC/P-4 STABILIZE RPV Pressure below 1073 psig using the Main Turbine Bypass Valves (2-EOI-Appendix-8B)

Use Alternate RPV Pressure Control Systems (Table P-1), if necessary NUSO Crosstie CAD or MSRV carts to DW Control Air (APPX 8G, 20H) if necessary IF THEN Drywell Control Air is or becomes NO ACTION REQUIRED unavailable NUSO 2-EOI-Appendix-1F, Manual SCRAM

[1] VERIFY Reactor SCRAM and ARI reset.

[1.1] IF ARI CANNOT be reset, THEN EXECUTE EOI Appendix 2, Defeating ARI Logic Trips, concurrently with Step 1.0[1.2] of this procedure.

OATC

[1.2] IF Reactor SCRAM CANNOT be reset, THEN DISPATCH personnel to Unit 2 Auxiliary Instrument Room to defeat ALL RPS Logic trips.

[2] WHEN RPS Logic has been defeated, THEN RESET Reactor SCRAM.

[3] VERIFY OPEN SCRAM Discharge Volume Vent and Drain Valves.

Dispatches personnel to perform outside portions of 2-EOI-Appendix-1F, OATC Manual SCRAM.

Page 38 of 55 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-2 Event No.: 7 Page 12 of 13 Event

Description:

Hydraulic Anticipated Transient Without SCRAM (ATWS)

Time Position Applicants Actions or Behavior

[4] DRAIN SCRAM Discharge Volume (SDV) UNTIL the following annunciators clear:

OATC WEST CRD DISCHARGE VOLUME WATER LEVEL HIGH HALF SCRAM (Panel 2-9-4, 2-XA-55-4A, Window 1).

EAST CRD DISCHARGE VOLUME WATER LEVEL HIGH HALF SCRAM (Panel 2-9-4, 2-XA-55-4A, Window 29).

The accumulators will drain in approximately 7 minutes, and the NRC alarms at Panel 2-9-4, Windows 1 and 29, will clear. The OATC may then attempt a Reactor SCRAM.

[5] DISPATCH personnel to VERIFY OPEN 2-SHV-085-0586, CHARGING WATER ISOLATION.

NOTES

1) If EOI Appendix 2 has been executed, ARI initiation or reset will OATC NOT be possible or necessary in Step 1.0[6].

2) If Reactor Pressure is greater than 600 PSIG, NUSO may direct performance of step 1.0[6] prior to accumulators being fully recharged.

[6] WHEN CRD Accumulators are recharged, THEN INITIATE manual Reactor SCRAM and ARI.

Control Rods will insert the first time the OATC attempts a Reactor NRC SCRAM after the ATWS.

[7] CONTINUE to perform Steps 1.0[1] through 1.0[6] UNTIL ANY of the following exists:

ALL Control Rods are fully inserted, OR OATC NO inward movement of Control Rods is observed, OR NUSO directs otherwise.

END OF EOI APPENDIX 1F Page 39 of 55 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-2 Event No.: 7 Page 13 of 13 Event

Description:

Hydraulic Anticipated Transient Without SCRAM (ATWS)

Time Position Applicants Actions or Behavior 2--EOI-8A, Bypassing Group 1 RPV Low Low Low Level Isolation Interlocks

[1] BYPASS Group 1 RPV Low-Low-Low Level Isolation Interlocks as follows (Unit 2 Control Room, Panel 9-4):

[1.1] PLACE keylock switch 2-HS-064-0056A, GROUP 1 RPV LOW LEVEL BYPASS (SYS A1), in BYPASS.

[1.2] PLACE keylock switch 2-HS-064-0056B, GROUP 1 RPV LOW LEVEL BYPASS (SYS B1), in BYPASS.

[1.3] PLACE keylock switch 2-HS-064-0056C, GROUP 1 RPV LOW LEVEL BYPASS (SYS A2), in BYPASS.

[1.4] PLACE keylock switch 2-HS-064-0056D, GROUP 1 RPV LOW LEVEL BYPASS (SYS B2), in BYPASS.

OATC [1.5] ENSURE closed the following valves (Unit 2 Control Room, Panel 9-3):

2-FCV-43-13, RX RECIRC SAMPLE INBOARD ISOLATION VALVE 2-FCV-43-14, RX RECIRC SAMPLE OUTBOARD ISOLATION VALVE

[2] N/A

[3] NOTIFY Unit Operator to ensure closed the following valves (Unit 2 Control Room, Panel 9-3):

2-FCV-43-13, RX RECIRC SAMPLE INBOARD ISOLATION VALVE 2-FCV-43-14, RX RECIRC SAMPLE OUTBOARD ISOLATION VALVE END OF EOI APPENDIX 8A End of Event 7. When the crew has inserted all Control Rods and NRC has control of Reactor Water Level above the Top of Active Fuel

((-) 162 inches) using high pressure systems, end of Scenario.

Page 40 of 55 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-2 Event No.: 8 Page 1 of 7 Event

Description:

2A EHC Pump Trip Time Position Applicants Actions or Behavior Event 8, EHC Pump Trip, is inserted on Simulator Setup. No action is required by the Driver to insert this event.

NRC Thirty (30) seconds after the MODE SWITCH is placed in RUN, 2A EHC pump will be stopped.

Acknowledges and reports the following alarms when received:

STANDBY EHC PUMP FAILED, 2-9-7B, Window 15 BOP EHC HYDRAULIC FLUID HEADER PRESSURE LOW, 2-9-7B, Window 1 Directs the BOP to respond in accordance with the appropriate Alarm NUSO Response Procedure(s).

Alarm Response Procedure, 2-ARP-9-7B STANDBY EHC PUMP FAILED, Window 15 Operator Action:

A. PERFORM the following on Panel 2-9-7:

1. CHECK alarm by checking 2-PI-47-7, EHC HEADER PRESSURE.

2. CHECK 2-HS-47-2A, EHC PUMP 2B and/or 2-HS-47-1A, EHC PUMP 2A running.

3. CHECK 2-EI-47-2, EHC PUMP 2B PUMP MTR CURRENT BOP and/or 2-EI-47-1, EHC PUMP 2A PUMP MTR CURRENT.

NOTE Lights extinguish at 1300 psig lowering and illuminate at 1500 psig rising.

4. CHECK lights above 2-HS-47-4A, EHC PUMP 2A TEST pushbutton and 2-HS-47-5A, EHC PUMP 2B TEST pushbutton.

B. DISPATCH personnel to pumping unit to check for abnormal conditions.

Page 41 of 55 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-2 Event No.: 8 Page 2 of 7 Event

Description:

2A EHC Pump Trip Time Position Applicants Actions or Behavior If contacted as the outside NUSO, Work Control, Maintenance, or Driver an AUO to investigate the cause for 2A EHC Pump Trip or to check for abnormal conditions, acknowledge the direction.

NOTE On EHC Hydraulic System failure accumulator and check valve arrangement will provide approximately one minute Bypass Valve BOP operation.

C. IF EHC Hydraulic System fails, THEN ENSURE Turbine trips at or below 1100 psig.

Alarm Response Procedure, 2-ARP-9-7B EHC HYDRAULIC FLUID HEADER PRESSURE LOW, Window 1 Operator Action:

A. N/A.

B. CHECK EHC HEADER PRESSURE indicator, 2-PI-47-7 between 1550 and 1650 psig.

BOP C. DISPATCH personnel to inspect EHC Pump unit.

NOTE On EHC Hydraulic System failure, accumulator and check valve arrangement will provide approximately one minute Bypass Valve operation.

D. IF EHC Hydraulic system fails, THEN ENSURE Turbine trips at or below 1100 psig.

Directs the BOP to maintain Reactor Pressure with Main Steam Relief NUSO Valves (MSRVs) using 2-EOI-Appendix-11A, Alternate RPV Pressure Control Systems MSRVs.

Page 42 of 55 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-2 Event No.: 8 Page 3 of 7 Event

Description:

2A EHC Pump Trip Time Position Applicants Actions or Behavior 2-EOI-Appendix-11A, Alternate RPV Pressure Control Systems MSRVs.

[1] N/A

[2] N/A

[3] OPEN MSRVs using the following sequence to control RPV pressure as directed by the SRO:

1 2-PCV-1-179 MN STM LINE A RELIEF VALVE 2 2-PCV-1-180 MN STM LINE D RELIEF VALVE 3 2-PCV-1-4 MN STM LINE A RELIEF VALVE 4 2-PCV-1-31 MN STM LINE C RELIEF VALVE 5 2-PCV-1-23 MN STM LINE B RELIEF VALVE 6 2-PCV-1-42 MN STM LINE D RELIEF VALVE BOP 7 2-PCV-1-30 MN STM LINE C RELIEF VALVE 8 2-PCV-1-19 MN STM LINE B RELIEF VALVE 9 2-PCV-1-5 MN STM LINE A RELIEF VALVE 10 2-PCV-1-41 MN STM LINE D RELIEF VALVE 11 2-PCV-1-22 MN STM LINE B RELIEF VALVE 12 2-PCV-1-18 MN STM LINE B RELIEF VALVE 13 2-PCV-1-34 MN STM LINE C RELIEF VALVE

[4] N/A

[5] N/A

[6] N/A END OF EOI APPENDIX 11A Acknowledges and reports the following alarm when received:

BOP SUPPRESSION POOL AVERAGE TEMPERATURE HIGH, 2-9-3E, Window 12 Page 43 of 55 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-2 Event No.: 8 Page 4 of 7 Event

Description:

2A EHC Pump Trip Time Position Applicants Actions or Behavior Directs the BOP to respond in accordance with the appropriate Alarm NUSO Response Procedure.

(If received)

Alarm Response Procedure, 2-ARP-9-3E SUPPRESSION POOL AVERAGE TEMPERATURE HIGH, Window 12 BOP Operator Action:

A. IF alarm is valid, THEN ENTER 2-EOI-2, Primary Containment Control.

NUSO Enters 2-EOI-2, Primary Containment Control.

NUSO Page 44 of 55 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-2 Event No.: 8 Page 5 of 7 Event

Description:

2A EHC Pump Trip Time Position Applicants Actions or Behavior SP/T-1 MONITOR and CONTROL Suppr Pool Temperature below 95°F using available Suppr Pool Cooling (APPX 17A)

SP/T-2 NUSO SP/T-3 OPERATE all available Suppression Pool Cooling using only RHR Pumps NOT required to assure adequate Core Cooling by continuous injection (APPX 17A)

Directs the BOP to place Suppression Pool Cooling in service in NUSO accordance with 2-EOI-Appendix-17A, RHR System Operation Suppression Pool Cooling.

2-EOI-Appendix-17A, RHR System Operation Suppression Pool Cooling BOP NOTE Placing a BYPASS SEL switch in BYPASS in Step 1.0[1] below prevents automatic opening of the affected RHR loop's Outboard Injection Valve. This makes LPCI Mode of that RHR Loop inoperable.

Page 45 of 55 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-2 Event No.: 8 Page 6 of 7 Event

Description:

2A EHC Pump Trip Time Position Applicants Actions or Behavior

[1] IF Adequate Core Cooling is assured OR directed to cool the Suppression Pool irrespective of Adequate Core Cooling, THEN BYPASS LPCI Injection Valve open interlock AS NECESSARY:

PLACE 2-HS-74-155A, LPCI SYSTEM I OUTBOARD INJECTION VALVE BYPASS SELECT in BYPASS PLACE 2-HS-74-155B, LPCI SYSTEM II OUTBOARD INJECTION VALVE BYPASS SELECT in BYPASS

[2] PLACE RHR SYSTEM I(II) in Suppression Pool Cooling as follows:

[2.1] ENSURE at least one RHRSW Pump supplying each EECW Header.

[2.2] ENSURE RHRSW Pump supplying desired RHR Heat Exchanger(s).

[2.3] THROTTLE the following in service RHRSW Outlet Valves to obtain required RHRSW Flow:

2-FCV-23-34, RHR HX 2A RHRSW OUTLET VALVE (Required flow is 1700 to 4500 gpm)

BOP 2-FCV-23-46, RHR HX 2B RHRSW OUTLET VALVE (Required flow is 1350 to 4500 gpm for B1 pump)

(Required flow is 1700 to 4500 gpm for B2 pump) 2-FCV-23-40, RHR HX 2C RHRSW OUTLET VALVE (Required flow is 1700 to 4500 gpm) 2-FCV-23-52, RHR HX 2D RHRSW OUTLET VALVE (Required flow is 1700 to 4500 gpm)

[2.4] IF Directed by SRO, THEN PLACE 2-XS-74-122(130), RHR SYS I(II) LPCI 2/3 CORE HEIGHT OVERRIDE, in MANUAL OVERRIDE.

[2.5] IF LPCI Initiation signal exists, THEN MOMENTARILY PLACE 2-XS-74-121(129), RHR SYS I (II) CONTAINMENT SPRAY/COOLING VALVE SELECT, in SELECT.

[2.6] IF 2-FCV-74-53(67), RHR SYS I(II) LPCI INBOARD INJECTION VALVE, is OPEN, THEN ENSURE CLOSED 2-FCV-74-52(66), RHR SYSTEM I(II) LPCI OUTBOARD INJECTION VALVE.

Page 46 of 55 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-2 Event No.: 8 Page 7 of 7 Event

Description:

2A EHC Pump Trip Time Position Applicants Actions or Behavior

[2.7] OPEN 2-FCV-74-57(71), RHR SYS I(II) SUPPRESSION CHAMBER/POOL ISOLATION VALVE.

[2.8] ENSURE desired RHR Pump(s) for Suppression Pool Cooling are operating.

CAUTION RHR System Flows below 7,000 gpm or above 10,000 gpm for one pump operation may result in excessive vibration and equipment damage.

[2.9] THROTTLE OPEN 2-FCV-74-59(73), RHR SYS I(II)

BOP SUPPRESSION POOL COOLING/TEST VALVE, to maintain EITHER of the following as indicated on 2-FI-74-50(64), RHR SYS I(II) FLOW:

Between 7,000 and 10,000 gpm for one pump operation OR At or below 13,000 gpm for two pump operation

[2.10] ENSURE CLOSED 2-FCV-74-7(30), RHR SYS I(II) MIN FLOW VALVE.

[2.11] MONITOR RHR Pump NPSH using Attachment 1.

[2.12] NOTIFY Chemistry that RHRSW is aligned to in service RHR Heat Exchangers.

When contacted as Chemistry, acknowledge any information Driver given.

[2.13] IF Additional Suppression Pool Cooling Flow is necessary, THEN PLACE additional RHR and RHRSW Pumps in service using BOP Steps 1.0[2.2] through 1.0[2.12].

[3] N/A END OF EOI APPENDIX 17A End of Event 8. Once the crew has inserted all Control Rods and NRC has control of Reactor Water Level above the Top of Active Fuel

((-) 162 inches) using high pressure systems, end of Scenario.

Page 47 of 55 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-2 Event No.: 9 Page 1 of 2 Event

Description:

SLC Pump Trip Time Position Applicants Actions or Behavior Event 9, SLC Pump Trip, is inserted on Simulator Setup. No action NRC is required by the Driver to insert this event.

NOTE: The first SLC Pump that is started will trip.

2-EOI-Appendix-3A, SLC Injection

[1] UNLOCK and PLACE 2-HS-63-6A, SLC PUMP 2A/2B, control switch in START-A or START-B position.

[2] CHECK SLC System for injection by observing the following:

Selected pump starts, as indicated by red light illuminated above pump control switch Squib valves fire, as indicated by SQUIB VALVE A and B CONTINUITY blue lights extinguished BOP SLC SQUIB VALVE CONTINUITY LOST Annunciator in alarm on Panel 9-5 (2-XA-55-5B, Window 20) 2-PI-63-7A, SLC PUMP DISCH PRESS, indicates above RPV Pressure System flow, as indicated by 2-IL-63-11, SLC FLOW, red light illuminated on Panel 9-5 SLC INJECTION FLOW TO REACTOR Annunciator in alarm on Panel 9-5 (2-XA-55-5B, Window 14).

[3] IF proper system operation CANNOT be verified, THEN RETURN to Step 1.0[1] and START other SLC pump.

Determines that the first SLC Pump that was started trips, and starts the BOP alternate SLC Pump.

[4] VERIFY RWCU isolation by observing the following:

RWCU Pumps 2A and 2B tripped 2-FCV-69-1, RWCU INBOARD SUCT ISOLATION VALVE BOP closed 2-FCV-69-2, RWCU OUTBOARD SUCT ISOLATION VALVE closed 2-FCV-69-12, RWCU RETURN ISOLATION VALVE closed Page 48 of 55 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-2 Event No.: 9 Page 2 of 2 Event

Description:

SLC Pump Trip Time Position Applicants Actions or Behavior

[5] VERIFY ADS inhibited.

[6] MONITOR Reactor Power for downward trend.

[7] MONITOR 2-LI-63-1A, SLC STORAGE TANK LEVEL, and CHECK that level is dropping approximately 1% per minute.

[8] WHEN EITHER of the following exists:

SLC Tank Level drops to 0%,

OR BOP As directed by SRO, THEN STOP SLC Pump 2A or 2B.

[9] NOTIFY Chemistry to mix additional solution to compensate for dilution as directed by the SRO.

[10] WHEN directed by the SRO to perform system flush, THEN REFER to OI-63, Standby Liquid Control System, Section 8.1, for system flush.

END OF 2-EOI-APPENDIX-3A Driver If contacted as Chemistry, acknowledge any direction given.

End of Event 9. Once the crew has inserted all Control Rods and NRC has control of Reactor Water Level above the Top of Active Fuel

((-) 162 inches) using high pressure systems, end of Scenario.

Page 49 of 55 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Scenario Setup UNIT 2 IC 28 Exam IC 277 Procedure Revision Procedure Revision Procedure Revision OI-68 161 ARP-3E 31 APPX-17A 18 OI-67 126 ARP-5A 60 APPX-1F 6 OI-92B 43 ARP-7B 36 APPX-8A 5 GOI-100-12A 118 ARP-8C 19 0-AOI-27-1 12 ARP-20A 30 AOI-47-3 22 EOI-1A 2 AOI-85-6 21 EOI-2 16 ODM 4.20 7 EOI-3 6 Verify camera system is powered down (admin password = abcd1234)

Start CPERF PRIOR to placing the Simulator in RUN Verify EECW Pump Alarm borders are properly arranged on Panels Simulator Setup 2-9-23A / B / C / D.

Hang Danger Tags on A3 EECW Pump and 2B EHC Pump.

Hang Protected Equipment Tag on 2A EHC Pump.

Schedule Files(s): 2104 NRC Scenario 2 UNIT 2.sch Event Files(s): 2104 NRC Scenario 2 UNIT 2.evt Schedule File: 2104 NRC Scenario 2 UNIT 2.sch Event Action Description 2104 NRC Scenario 2 UNIT 2.evt 2 Insert malfunction SW03M RHRSW PUMP D3 TRIP 3 set nmaprmgain(1)=0.000005 Fails APRM 1 Low 4 Insert malfunction RD04R3011 DRIFT ANY CONTROL ROD OUT 30-11 14 Delete malfunction RD04R3011 DRIFT ANY CONTROL ROD OUT 30-11 5 Insert malfunction OG04A OFFGAS SJAE ISOLATION (RELAY R5A)

Insert malfunction MC05 to 100.00000 in INTAKE PUMPING STATION TRAVELING SCREENS 6

600 CLOG Page 50 of 55 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Schedule File: 2104 NRC Scenario 2 UNIT 2.sch Event Action Description Insert malfunction MC05 to 50.00000 in INTAKE PUMPING STATION TRAVELING SCREENS 7 120 CLOG 26 Insert malfunction MC02A after 60 CCW PUMP 2A MOTOR TRIP Insert malfunction RD09A after 2 to ATWS (HYDRAULIC LOCK) EAST 55.00000 Insert malfunction RD09B after 2 to ATWS (HYDRAULIC LOCK) WEST 55.00000 Insert malfunction RD06ALLSTICK STICK ALL CONTROL RODS EAST SDV LEVEL SWITCH FAILS LOW Insert malfunction RD17A (LS-85-45E,F,G,H,K,M)

WEST SDV LEVEL SWITCH FAILS LOW Insert malfunction RD17B (LS-85-45A,B,C,D,J,L)

Insert override HS-47-1A after 30 to 7 EHC HYD FLUID PUMP 2A STOP Insert malfunction PMP-47-1 to 52_BREAKER EHC HYD FLUID PUMP A FAIL_CCOIL Insert malfunction SL01A SLC PUMP 2A TRIP Insert malfunction SL01B SLC PUMP 2B TRIP 9 Delete malfunction SL01A SLC PUMP 2A TRIP 10 Delete malfunction SL01B SLC PUMP 2B TRIP 17 Insert remote RD06 to CLOSE CRD CHARGING WATER VALVE FCV-2-85-586 18 Insert remote RD06 to OPEN CRD CHARGING WATER VALVE FCV-2-85-586 Schedule F:\Scenarios\ Scenario 2\App.

19 1F.sch Schedule 20 F:\2104\NRC\Scenarios\U2\Scenario 2\App. 2.sch Schedule F:\Scenarios\ Scenario 2\App.

21 8A.sch Schedule F:\Scenarios\ Scenario 2\App.

22 8E.sch Page 51 of 55 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Schedule File: 2104 NRC Scenario 2 UNIT 2.sch Event Action Description EAST SDV LEVEL SWITCH FAILS LOW 7 Delete malfunction RD17A (LS-85-45E,F,G,H,K,M)

WEST SDV LEVEL SWITCH FAILS LOW 7 Delete malfunction RD17B (LS-85-45A,B,C,D,J,L) 27 Delete malfunction RD09A ATWS (HYDRAULIC LOCK) EAST 27 Delete malfunction RD09B ATWS (HYDRAULIC LOCK) WEST 27 Delete malfunction RD06ALLSTICK STICK ALL CONTROL RODS Schedule File: App. 1F.sch Event Action Description Insert remote RP13A to BYP after 300 DEFEAT CHANNEL A1 AUTO SCRAM Insert remote RP13B to BYP after 300 DEFEAT CHANNEL B1 AUTO SCRAM Insert remote RP13C to BYP after 300 DEFEAT CHANNEL A2 AUTO SCRAM Insert remote RP13D to BYP after 300 DEFEAT CHANNEL B2 AUTO SCRAM Schedule File: App. 2.sch Event Action Description OPERATE LOCAL ATWS MODE SWITCH Insert remote RP12A to TEST after 300 2-HS-68-118A OPERATE LOCAL ATWS MODE SWITCH Insert remote RP12B to TEST after 300 2-HS-68-118B Page 52 of 55 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Schedule File: App. 8A.sch Event Action Description Insert remote RP06A to BYP after 300 BYP MSIV GP 1 LO LVL -APPNDX 8A-16A-K1A Insert remote RP06B to BYP after 300 BYP MSIV GP 1 LO LVL -APPNDX 8A-16A-K1B Insert remote RP06C to BYP after 300 BYP MSIV GP 1 LO LVL -APPNDX 8A-16A-K1 Insert remote RP06D to BYP after 300 BYP MSIV GP 1 LO LVL -APPNDX 8A-16A-K1D Schedule File: App. 8E.sch Event Action Description BYPASS GROUP 6 INBOARD LOW LVL/HI DW Insert remote RP14A to BYP after 300 PRESS INTLKS BYPASS GROUP 6 OUTBOARD LOW LVL/HI DW Insert remote RP14B to BYP after 300 PRESS INTLKS Page 53 of 55 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Event File: 2104 NRC Scenario 2 UNIT 2.evt List Details Page 54 of 55 Unit 2

UNIT 2 SHIFT TURNOVER MEETING Today DAYS ON LINE Total Drywell Leakage Protected Equipment MODE 227 (gpm) 2A EHC Pump 1 PRA (EOOS) -GREEN 1.55 Rx Power 500Kv GRID - Qualified Floor Drain (gpm) 100% 161Kv Grid -Qualified 0.11 Equipment Drain MWe Last breaker closure (gpm) 1295 10/01/20 4:31 1.44 Review logs Qualifications Review RCP/Rx Brief Review LCO/OWA Actions Walkdown Panels/Verify EOOS CR Reviews Complete Leadership and Team Effectiveness CHANGES IN LCOs A3 EECW Pump is tagged for oil change (information only LCO).

LCOs OF 72 HOURS OR LESS SIGNIFICANT ITEMS DURING PREVIOUS SHIFT/RADIOLOGICAL CHANGES A3 EECW Pump is tagged for oil change 2B EHC Pump tagged for discharge filter replacement.

2A CCW Pump repairs are complete, tags are cleared. Ready to re-start when Maintenance is ready.

MAJOR EQUIPMENT CHANGES PLANNED FOR THIS SHIFT Continue to support A3 EECW and 2B EHC Pump maintenance.

Alternate both Recirc Drive Cooling Water Pumps.

OPERATOR WORK AROUNDS OWAs - 1* Burdens - 0 Challenges - 7 ODMIs/ACMPs ONEAs FIRE RISK SIGNIFICANT ITEMS OOS/FPLCO Actions Due SCHEDULED ITEMS NOT COMPLETED Page 55 of 55 Unit 2

Appendix D Scenario Outline Form ES-D1 Facility: BFN Scenario Number: NRC-3 Op-Test Number:21-04 Examiners: __________________________ Operators: SRO: _____________________

__________________________ ATC: _____________________

BOP: _____________________

Initial Conditions: ~2.2 % Reactor Power.

Turnover: Reactor Startup in progress. Raise Reactor Power and place the Reactor MODE SWITCH in RUN. A thunderstorm watch has just been issued for Limestone County, AL. H IRM is bypassed due to noise.

Critical Tasks:

1. When Suppression Pool Water Level cannot be maintained within the safe area of Curve 4, SRV Tail Pipe Level Limit, the crew will insert a manual Reactor SCRAM as directed by the Nuclear Unit Senior Operator (NUSO).

2. When Suppression Pool Water Level and RPV Pressure cannot be restored and maintained within the safe area of Curve 4, the NUSO determines that Emergency Depressurization is required. The Operator initiates Emergency Depressurization as directed by the NUSO.

Event Malfunction Event Event Description Number Number Type*

N-BOP

1. N/A Transfer Seal Steam to Main Steam N-NUSO R-OATC

2. N/A Raise Reactor Power using Control Rods R-NUSO C-OATC

3. XS-92-7/42G IRM Failure C-NUSO

4. RD08R2227 TS-NUSO Control Rod Accumulator INOPERABLE SCHEDULE C-BOP Failure of 3A Stack Dilution Fan, Standby Fan Fails to 5.

STACK C-NUSO Automatically Start TS-NUSO 3EA Emergency Diesel Generator (EDG) Logic Breaker

6. DG01B Tripped SCHEDULE High Suppression Pool Water Level /

7. M-ALL TORUS Emergency Depressurization RD06R3019 C-OATC

8. Two Control Rods Fail to Insert RD06R2615 C-NUSO ED10A C-BOP 9.# 480V Shutdown Board Trip ED10B C-NUSO

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor (TS)Technical Specification

Event on previous two NRC Exams

S Event on previous two NRC Exams Scenario (Spare)

Unit 3 - Page 1 of 4

Appendix D Scenario Outline Form ES-D1 Events

1. The crew will transfer Seal Steam from Auxiliary Steam to Main Steam in accordance with 3-OI-47C, Seal Steam System.

2. The crew will continue the Reactor startup by withdrawing Control Rods in accordance with 3-OI-85, Control Rod Drive System, and 3-GOI-100-1A, Unit Startup and Power Operation.

3. During Control Rod withdrawal, the IRM G Range Switch will fail in position 8, requiring the crew to bypass IRM G in accordance with 3-OI-92A, Intermediate Range Monitors. The Nuclear Unit Senior Operator (NUSO) will reference Technical Specification 3.3.1.1, RPS Instrumentation, Table 3.3.1.1.

4. Control Rod 22-27 Accumulator Pressure will lower below 940 psig, requiring the NUSO to address Technical Specification 3.1.5, Control Rod SCRAM Accumulators, Condition A.

5. Stack Dilution Fan 3A will fail and the standby Stack Dilution Fan will fail to automatically start. The crew will take action to manually restore Stack Dilution Flow by starting 3B Stack Dilution Fan in accordance with 3-OI-66, Off-Gas System, Section 5.1.

6. The 3EA Emergency Diesel Generator (EDG) Logic Breaker will trip, causing an annunciation in the Control Room and disabling 3EA EDG. The NUSO will address Technical Specification 3.8.1, AC Sources - Operating, Condition B.

7. The crew will respond to a rising Suppression Pool Water Level in accordance with Emergency Operating Instruction 3-EOI-2, Primary Containment Control.

Suppression Pool Water Level will not be able to be maintained in the safe area of Curve 4, SRV Tail Pipe Level Limit, requiring the crew to Emergency Depressurize the Reactor in accordance with 3-C-2, Emergency RPV Depressurization.

8. When the crew inserts a manual Reactor SCRAM due to high Suppression Pool Water Level, two Control Rods will fail to insert, requiring the crew to take actions to insert the Control Rods in accordance with 3-AOI-100-1, Reactor SCRAM.

9. When the crew is Emergency Depressurizing the Reactor due to rising Suppression Pool Water Level, 480V Shutdown Power to the injection valves on the loop the crew selects to inject water into the core will be lost, requiring action to choose another source for Reactor Water injection The Scenario ends when the crew has Emergency Depressurized the Reactor and has control of Reactor Water Level above the Top of Active Fuel (TAF, (-) 162 inches) using low pressure systems.

Unit 3 - Page 2 of 4

Appendix D Scenario Outline Form ES-D1 Critical Tasks: 2

1. When Suppression Pool Water Level cannot be maintained within the safe area of Curve 4, SRV Tail Pipe Level Limit, the crew will insert a manual Reactor SCRAM as directed by the Nuclear Unit Senior Operator (NUSO).

a. Safety Significance Prevent failure of Primary Containment from over pressurization.

b. Cues Procedural Compliance.

Suppression Pool Level indication.

c. Measured by Observation - Both RPS SCRAM switches are depressed.

d. Feedback Control Rods insert to their full in position.

e. Critical Task Failure Criteria The operating crew fails to proceed without delay and in a controlled manner to insert a manual SCRAM before Suppression Pool Water Level enters the Action Required area of Curve 4, SRV Tail Pipe Level Limit.

2. When Suppression Pool Water Level and RPV Pressure cannot be restored and maintained within the safe area of Curve 4, the NUSO determines that Emergency Depressurization is required. The Operator initiates Emergency Depressurization as directed by the Nuclear Unit Senior Operator.

a. Safety Significance Prevent failure of Primary Containment from over pressurization.

b. Cues Procedural Compliance.

RPV Pressure indication.

Suppression Pool Level indication.

c. Measured by Observation - the Nuclear Unit Senior Operator determines (as indicated by announcement or observable transition to 3-C-2, RPV Emergency Depressurization), Emergency Depressurization is required at or before Suppression Pool Water Level and RPV Pressure cannot be restored and maintained within the safe area of Curve 4.

AND Observation - the Nuclear Unit Senior Operator directs the Operator to open 6 ADS valves Unit 3 - Page 3 of 4

Appendix D Scenario Outline Form ES-D1

d. Feedback Suppression Pool Water Level trend.

RPV Pressure trend.

MSRV status indication.

e. Critical Task Failure Criteria The operating crew fails to proceed without delay and in a controlled manner to initiate Emergency Depressurization from the time that Suppression Pool Water Level and RPV Pressure cannot be restored and maintained within the safe area of Curve 4.

Unit 3 - Page 4 of 4

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-3 Event No.: 1 Page 1 of 2 Event

Description:

Transfer Seal Steam to Main Steam Time Position Applicants Actions or Behavior Prior to placing the simulator in RUN, start CPERF to record critical Driver parameters.

If the crew does not proceed to Event 1, request that the Driver NRC insert Event 1.

If contacted by the Chief Examiner to insert Event 1, contact the Nuclear Unit Supervisor Operator (NUSO) as the Shift Manager and Driver direct the crew to transfer Seal Steam from Auxiliary Steam to Main Steam.

Directs the Balance of the Plant Operator (BOP) to transfer Seal Steam NUSO from Auxiliary Steam to Main Steam in accordance with 3-OI-47C, Seal Steam System, Section 6.1.

3-OI-47C, Seal Steam System Section 6.1, Shifting Supply from Auxiliary Steam to Main Steam NOTES

1) Section 6.1 is entered with the Seal Steam supply on Auxiliary Steam.

2) Section 6.1 is performed at Panel 3-9-7 in the Control Room unless otherwise specified.

3) To seal the Turbine at startup with less than 250 psig Main Steam Pressure, or with worn packings, the 3-FCV-1-145, STEAM SEAL BOP REGULATOR BYPASS VALVE, is required to be adjusted to supplement 3-PCV-1-147, STEAM SEAL FEED VALVE, to obtain the needed flow.

4) The 3-PCV-1-147, STEAM SEAL FEED VALVE, is designed to handle the Steam Seal Header requirements when Main Steam Pressure exceeds 250 psig.

[1] BEFORE placing Seal Steam System on Nuclear Steam, PERFORM the following:

[1.1] NOTIFY Radiation Protection that an RPHP is in effect for the impending action to place Seal Steam System on nuclear steam.

RECORD time Radiation Protection notified in the Narrative Log.

Page 1 of 39 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-3 Event No.: 1 Page 2 of 2 Event

Description:

Transfer Seal Steam to Main Steam Time Position Applicants Actions or Behavior NRC An RPHP was provided to the crew at turnover.

[2] CHECK Reactor Pressure is greater than 200 psig.

[3] ENSURE OPEN 3-PCV-1-147, STEAM TO STEAM SEAL PRESS REGULATOR by placing 3-HS-1-147, STEAM SEAL REGULATOR, in AUTO.

[4] OPEN 3-FCV-1-146, MAIN STEAM SUPPLY TO STEAM SEAL VALVE.

BOP [5] CLOSE 3-FCV-1-154, AUX BOILER SUPPLY TO STEAM SEAL VALVE.

[6] CHECK Steam Seal Header Pressure, as indicated on 3-PI-1-148A, STEAM SEAL HEADER PRESSURE, is between 2 1/2 and 9 psig.

[7] CLOSE 3-SHV-012-0638, TURBINE SEAL STM VALVE. (Turbine Building Elevation 586', T16 J-Line near the EHC Unit behind Panel 25-111)

When directed as the Turbine Building AUO to close 3-12-638, Driver TURBINE SEAL STM VALVE, acknowledge the direction and inform the crew that 3-12-638 is closed.

If contacted as the Boiler AUO that steam load is being removed, Driver acknowledge report.

[8] ENSURE CLOSED 3-FCV-001-0149, STEAM SEAL UNLOADING MANUAL BYPASS VALVE.

CAUTION Throttling 3-BYV-001-0145, STEAM SEAL REG BYPASS VALVE, with Main Steam Pressure above 250 psig could result in excessive vibration of the Steam Seal Header.

BOP

[9] THROTTLE 3-FCV-1-145, STEAM SEAL REGULATOR BYPASS VALVE, to keep Steam Seal Header Pressure, as indicated on 3-PI-1-148A, STEAM SEAL HEADER PRESSURE, between 2 1/2 and 9 psig, if necessary.

[10] CHECK SPE Vacuum, as indicated on 3-PI-66-54, STEAM PACKING EXHAUST VACUUM, is between 10 and 12 in H2O vacuum.

End of Event 1. Proceed to Event 2, Raise Reactor Power Using NRC Control Rods.

Page 2 of 39 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-3 Event No.: 2 Page 1 of 4 Event

Description:

Raise Reactor Power using Control Rods Time Position Applicants Actions or Behavior If the crew does not proceed to Control Rod withdrawal, request NRC that the Driver insert Event 2.

If requested by the Chief Examiner, contact the NUSO as the Shift Driver Manager and direct the crew to continue the Reactor Startup.

During Control Rod withdrawal, Event 3, IRM Failure will NRC automatically be inserted. No action is required by the driver to insert Event 3.

(The crew may elect to conduct a reactivity re-focus brief)

Assumes the Reactivity Manager position.

Directs the Operator at the Controls (OATC) to raise Reactor Power to NUSO 6-7% in preparation for placing the Reactor MODE SWITCH in RUN, in accordance with 3-GOI-100-1A, Unit Startup and Power Operation, 3-OI-85, Control Rod Drive System, and 3-SR-3.1.3.5(A), Control Rod Coupling Integrity Check.

3-GOI-100-1A, Unit Startup Section 5.4, Withdrawal of Control Rods while in Mode 2 OATC [83] CONTINUE to withdraw Control Rods to raise Reactor Power to 6%

to 7% per 3-OI-85, Control Rod Drive System, and 3-SR-3.1.3.5(A),

Control Rod Coupling Integrity Check.

[84] ENSURE all operable APRM downscale alarms are reset and no rod blocks exist.

3-OI-85, Control Rod Drive System Section 6.6, Control Rod Withdrawal 6.6.1 Initial Conditions Prior to Withdrawing Control Rods

[1] REVIEW Precautions and Limitations in Section 3.7 and Section 3.8.

OATC

[2] CHECK the following prior to Control Rod movement:

CRD POWER, 3-HS-85-46 in ON.

Rod Worth Minimizer is operable and LATCHED into the correct ROD GROUP when Rod Worth Minimizer is enforcing (not required with no fuel in RPV)

Page 3 of 39 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-3 Event No.: 2 Page 2 of 4 Event

Description:

Raise Reactor Power using Control Rods Time Position Applicants Actions or Behavior 3-OI-85, Control Rod Drive System Section 6.6.2, Actions Required during and Following Control Rod Withdrawal

[1] IF the Control Rod fails to withdraw, THEN Refer to Section 8.15 for additional methods to reposition Control Rod.

[2] IF the Control Rod double notches, or withdraws past its correct/desired position, THEN Refer to Section 6.7 for inserting Control Rod to its correct/desired position. [NRC IR 84-02]

[3] IF at any time while driving a selected rod during the performance of this section, the Control Rod moves more than one notch from its intended position, THEN Refer to 3-AOI-85-7, Mispositioned Control Rod.

[4] OBSERVE the following during Control Rod repositioning:

Control Rod reed switch position indicators (four rod display)

OATC agree with the indication on the Full Core Display.

Nuclear Instrumentation responds as Control Rods move through the core (This ensures Control Rod is following drive during Control Rod movement.)

[5] ATTEMPT to minimize automatic RBM Rod Block as follows:

STOP Control Rod withdrawal (if possible) prior to reaching any RBM Rod Block using the RBM displays on Panel 3-9-5 and PERFORM Step 6.6.2[6].

[6] IF Control Rod movement was stopped to keep from exceeding a RBM setpoint or was caused by a RBM Rod Block, THEN PERFORM the following at the Unit SROs discretion to REINITIALIZE the RBM:

[6.1] PLACE 3-HS-85-46, CRD POWER in the OFF position to deselect the Control Rod.

[6.2] PLACE 3-HS-85-46, CRD POWER, in the ON position.

[6.3] IF desired, THEN CONTINUE to withdrawal Control Rods and PERFORM applicable section for Control Rod withdrawal.

Page 4 of 39 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-3 Event No.: 2 Page 3 of 4 Event

Description:

Raise Reactor Power using Control Rods Time Position Applicants Actions or Behavior Order of Control Rod withdrawal in accordance with 3-SR-3.1.3.5(A), Control Rod Coupling Integrity Check:

NRC Page 5 of 39 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-3 Event No.: 2 Page 4 of 4 Event

Description:

Raise Reactor Power using Control Rods Time Position Applicants Actions or Behavior 3-OI-85, Control Rod Drive System Section 6.6.3, Control Rod Notch Withdrawal

[1] SELECT the desired Control Rod by depressing the appropriate 3-XS-85-40, CRD ROD SELECT pushbutton.

[2] ENSURE CRD DRIVE WATER HEADER DP is between 250 -270 psid on 3-PDI-85-17A by throttling 3-HS-85-23A, CRD DRIVE WATER PRESS CONTROL VALVE, as necessary.

[3] N/A

[4] OBSERVE the following for the selected Control Rod:

OATC CRD ROD SELECT pushbutton is brightly ILLUMINATED White light on the Full Core Display ILLUMINATED Rod Out Permit light ILLUMINATED

[5] CHECK Rod Worth Minimizer is operable and LATCHED into the correct ROD GROUP when the Rod Worth Minimizer is enforcing.

[6] PLACE CRD CONTROL SWITCH, 3-HS-85-48, in ROD OUT NOTCH, and RELEASE.

[7] OBSERVE the Control Rod settles into the desired position and the ROD SETTLE light extinguishes.

[8] N/A

[9] N/A 3-OI-85, Control Rod Drive System 6.6.5 Return to Normal After Completion of Control Rod Withdrawal OATC [1] WHEN Control Rod movement is no longer desired AND deselecting Control Rods is desired, THEN:

[1.1] PLACE 3-HS-85-46, CRD POWER in OFF.

[1.2] PLACE 3-HS-85-46, CRD POWER in ON.

End of Event 2. Event 3, IRM Failure is automatically inserted on NRC simulator setup. No action is required by the Driver to insert Event 3.

Page 6 of 39 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-3 Event No.: 3 Page 1 of 2 Event

Description:

IRM Failure Time Position Applicants Actions or Behavior Event 3, IRM Failure, is automatically inserted on Simulator setup.

NRC No action is required by the Driver to insert Event 3.

IRM G will need to be ranged up after approximately 10 Control NRC Rods are pulled (about Control Rod 42-11 in Group 26).

During Control Rod withdrawal will be ranging IRM Switches to prevent a Control Rod Block or Reactor SCRAM Signal. When the OATC OATC attempts to range IRM G UP, the reading will not change.

Stops Control Rod withdrawal. Notifies the NUSO.

Directs the OATC to bypass IRM G in accordance with 3-OI-92A, NUSO Intermediate Range Monitors.

3-OI-92A, Intermediate Range Monitors Section 6.1, Bypassing an IRM Channel NOTES

1) It is not necessary for a bypassed IRM channel to have its detector inserted into the Core.

2) Only one IRM in each trip system can be bypassed at a time.

3) All operations are performed on Panel 3-9-5 unless specifically stated otherwise.

OATC CAUTION NPG-SPP 10.4 requires approval of the Plant Manager or his designee prior to any planned operation with IRMs bypassed unless bypassing is specifically allowed within approved procedures.

[1] REVIEW all precautions and limitations in Section 3.0.

[2] PLACE the appropriate IRM Bypass selector switch to the BYPASS position:

3-HS-92-7A/S4A, IRM BYPASS 3-HS-92-7A/S4B, IRM BYPASS

[3] CHECK that the Bypassed light is illuminated.

Page 7 of 39 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-3 Event No.: 3 Page 2 of 2 Event

Description:

IRM Failure Time Position Applicants Actions or Behavior If contacted by the crew as the Shift Manager, acknowledge any report given.

Driver If contacted as the Plant Manager / Shift Manager for permission to continue Control Rod withdrawal with IRM G bypassed, acknowledge the request and ask for their recommendation.

Concur with any recommendation given.

OATC Informs the NUSO that IRM G is bypassed.

References Technical Specification 3.3.1.1, RPS Instrumentation.

Declares an Information Only LCO based on only three IRM channels being required per trip system in accordance with Table 3.3.1.1-1 (page 1 of 3).

NUSO End of Event 3. Request that the driver insert Event 4, Control Rod NRC Accumulator Inoperable.

Page 8 of 39 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-3 Event No.: 4 Page 1 of 4 Event

Description:

Control Rod Accumulator Inoperable Time Position Applicants Actions or Behavior When requested by the Chief Examiner, insert Event 5, Control Rod Driver Accumulator Inoperable.

NRC The alarm will occur on Control Rod 22-27 Accumulator.

Acknowledges and reports the following alarm to the NUSO:

OATC CRD ACCUMULATOR PRESSURE LOW / LEVEL HIGH, 3-9-5A, Window 29 Directs the Balance of Plant Operator (BOP) to respond in accordance NUSO with the appropriate Alarm Response Procedure.

Alarm Response Procedure, 3-ARP-9-5A CRD ACCUMULATOR PRESSURE LOW / LEVEL HIGH, 3-9-5A, Window 29 Operator Action:

A. CHECK alarm by amber background light illuminated on Full Core Display.

B. LOG in the narrative log the Control Rod number and time alarm was received.

NOTE OATC If any of the following fuses is/are cleared, the local indications at Panel 25-4 and 25-22 will NOT illuminate.

C. IF multiple accumulator lights are lit on Panel 3-9-5, THEN CHECK for cleared fuses 3-FU1-085-25-004G, -004H in Panel 25-4 and 3-FU1-085-25-022G, -022H in Panel 25-22.

D. DISPATCH personnel to Panel 25-4 (east side), Panel 25-22 (west side) El 565', to determine if level high or pressure low.

E. DEPRESS push-button for associated HCU to determine if alarm is caused by level high or pressure low as follows:

1. If alarm is due to high level, the red light will extinguish.

2. If light stays illuminated, alarm is due to low N2 pressure.

Page 9 of 39 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-3 Event No.: 4 Page 2 of 4 Event

Description:

Control Rod Accumulator Inoperable Time Position Applicants Actions or Behavior If contacted as the Reactor Building AUO to respond to the CRD Accumulator alarm, acknowledge the direction. Wait two minutes Driver and report to the crew that CRD 22-27 Accumulator Pressure is 900 psig and lowering. You are unable to raise Accumulator Nitrogen Pressure by recharging.

F. IF alarm is valid, THEN REFER TO 3-OI-85, Control Rod Drive OATC System and 3-AOI-85-3, CRD System Failure.

NOTE If accumulator pressure is greater than 940 psig the accumulator is not required to be declared Inoperable when the star valve is CLOSED, unless accumulator is unattended.

G. IF Accumulator Pressure is less than or equal to 940 psig, THEN DECLARE Control Rod HCU INOPERABLE.

H. IF the associated HCUs nitrogen pressure is found less than 940 psi, THEN INITIATE a Condition Report (CR) to calibrate the pressure switch. The HCU will NOT be declared operable until the switch has OATC been calibrated.

I. IF alarm is due to low pressure with pressure greater than 940 psig and accumulator CANNOT be recharged within one hour, THEN EVALUATE per Tech Spec 3.1.5.

1. IF the Control Rod is declared SLOW REFER TO TECH SPEC 3.1.4. Currently no more than 13 OPERABLE Control Rods shall be slow and no more than 2 OPERABLE Control Rods that are slow shall occupy adjacent locations.

2. IF the Control Rod is declared INOPERABLE, THEN REFER TO TECH SPEC 3.1.3.

J. RECORD this evaluation in narrative log.

K. N/A Page 10 of 39 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-3 Event No.: 4 Page 3 of 4 Event

Description:

Control Rod Accumulator Inoperable Time Position Applicants Actions or Behavior Technical Specification 3.1.5, Control Rod SCRAM Accumulators LCO 3.1.5: Each Control Rod SCRAM Accumulator shall be OPERABLE APPLICABILITY: Modes 1 and 2 NUSO NOTE: Separate Condition entry is allowed for each Control Rod SCRAM Accumulator.

CONDITION:

A.1 One Control Rod SCRAM Accumulator inoperable with Reactor Steam Dome Pressure 900 psig.

REQUIRED ACTION: COMPLETION TIME:

A.1 A.1 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months NOTE: Only applicable if the associated Control Rod SCRAM time was within the limits of Table 3.1.4-1 during the last SCRAM time Surveillance.

NUSO Declare the associated Control Rod SCRAM Time slow.

OR A.2- Declare the associated Control A.2 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Rod INOPERABLE.

If Tech Spec 3.1.5 A.2 is Declared, THEN Tech Spec 3.1.3 is NRC applicable.

Page 11 of 39 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-3 Event No.: 4 Page 4 of 4 Event

Description:

Control Rod Accumulator Inoperable Time Position Applicants Actions or Behavior If the Control Rod is declared INOPERABLE, references Tech Spec 3.1.3, Control Rod OPERABILITY.

LCO 3.1.3 Each Control Rod shall be OPERABLE APPLICABILITY: MODES 1 and 2 NUSO NOTE: Separate Condition entry is allowed for each Control Rod.

CONDITION:

C. One or more Control Rods INOPERABLE for reasons other than Condition A or B.

REQUIRED ACTION: COMPLETION TIME:

NOTE: RWM may be bypassed as allowed by LCO 3.3.2.1, if required, to allow insertion of INOPERABLE Control Rod and NUSO continued operation.

C.1 - Fully insert INOPERABLE C.1 - 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months Control Rod.

AND C.2 - Disarm the associated CRD. C.2 - 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months If the Control Rod is declared SLOW, references Tech Spec 3.1.4, Control Rod SCRAM Times.

NUSO Currently, there are less than 13 OPERABLE Control Rods that are slow, and there are no OPERABLE Control Rods that are slow that occupy adjacent locations. Therefore, there are no actions required for Tech Spec 3.1.4, Control Rod SCRAM Times.

End of Event 4. Request that the driver insert Event 5, Failure of 3A NRC Stack Dilution Fan, Standby Fan Fails to Automatically Start.

Page 12 of 39 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-3 Event No.: 5 Page 1 of 1 Event

Description:

Failure of 3A Stack Dilution Fan, Standby Fan Fails to Automatically Start Time Position Applicants Actions or Behavior When requested by the Chief Examiner, insert Event 4, Failure of Driver 3A Stack Dilution Fan, Standby Fan Fails to Automatically Start.

Acknowledges and reports the following alarm to the NUSO:

BOP STACK GAS DILUTION AIR FLOW LOW, 3-9-7A, Window 3 Directs the BOP to respond in accordance with the appropriate Alarm NUSO Response Procedure.

It is acceptable if BOP elects to start 3B Stack Dilution Fan in NRC accordance with 3-OI-66, Section 8.6 instead of using the ARP below.

Alarm Response Procedure, 3-ARP-9-7A STACK GAS DILUTION AIR FLOW LOW, Window 3 BOP Operator Action:

A. CHECK alternate fan ON and damper open, (red light illuminated) on Panel 3-9-7.

Determines that the standby Stack Dilution Fan did not automatically BOP start and manually starts 3B Stack Dilution Fan.

B. DISPATCH personnel to stack to check and report status of the following for both fans:

1. Fan motor.

BOP

2. Fan belts.

3. Damper stuck closed.

C. CHECK Breaker 5E on 480V Diesel Aux Bd A and B.

Driver If contacted as an AUO, acknowledge any direction given.

End of Event 5. Request that the Driver insert Event 6, 3A NRC Emergency Diesel Generator (EDG) Logic Breaker Tripped.

Page 13 of 39 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-3 Event No.: 6 Page 1 of 3 Event

Description:

3EA Emergency Diesel Generator (EDG) Logic Breaker Tripped Time Position Applicants Actions or Behavior When requested by the Chief Examiner, insert Event 6, 3EA Driver Emergency Diesel Generator (EDG) Logic Breaker Tripped.

Acknowledges and reports the following alarm to the NUSO:

BOP DIESEL GENERATOR 3A CONTROL POWER OFF, 3-9-23A, Window 14 Directs the BOP to respond in accordance with the appropriate Alarm NUSO Response Procedure.

Alarm Response Procedure, 3-ARP-9-23A DIESEL GENERATOR 3A CONTROL POWER OFF, Window 14 Operator Action:

A. OBSERVE any other alarms on panels 9-8 or 9-23 which may BOP indicate problem area.

B. CHECK panels, breakers and batteries. IF necessary, THEN CHECK fuses and relays.

C. IF loss of normal power has occurred, THEN TRANSFER to alternate power source. REFER TO 0-OI-57D DC Electrical System.

D. REFER TO TS 3.8.1, 3.8.2, 3.8.4, and 3.8.5.

If contacted as the Outside NUSO, Work Control, AUO, or Electrical Maintenance, acknowledge any direction given.

Driver Wait 45 seconds and report that the Logic Breaker for EDG 3EA is tripped. If directed to attempt to close the Logic Breaker, report that the breaker will not close or RESET.

There are no required actions for Tech Spec 3.3.8.1, LOP NRC Instrumentation or Tech Spec 3.8.7, Distribution Systems -

Operating.

Page 14 of 39 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-3 Event No.: 6 Page 2 of 3 Event

Description:

3EA Emergency Diesel Generator (EDG) Logic Breaker Tripped Time Position Applicants Actions or Behavior Declares 3EA EDG INOPERABLE References Technical Specification 3.8.1, AC Sources - Operating.

LCO 3.8.1 The following AC electrical power sources shall be OPERABLE:

a. Two qualified circuits between the offsite transmission network and the onsite Class 1E AC Electrical Power Distribution System;

b. Unit 3 diesel generators (DGs) with two divisions of 480 V NUSO load shed logic and common accident signal logic OPERABLE; and

c. Unit 1 and 2 DG(s) capable of supplying the Unit 1 and 2 4.16 kV shutdown board(s) required by LCO 3.8.7, "Distribution Systems - Operating."

APPLICABILITY: MODES 1, 2, and 3.

CONDITION: One required Unit 3 DG INOPERABLE REQUIRED ACTION: COMPLETION TIME:

B.1 - Verify power availability from B.1 - 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months the offsite transmission network. AND Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter AND B.2 - Evaluate availability of both B.2 - 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months NUSO temporary diesel generators (TDGs). AND Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter AND B.3. - Declare required feature(s), B.3 - 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months from discovery of supported by the inoperable Unit 3 Condition B concurrent with DG, inoperable when the redundant inoperability of redundant required feature(s) are inoperable. required feature(s)

Page 15 of 39 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-3 Event No.: 6 Page 3 of 3 Event

Description:

3EA Emergency Diesel Generator (EDG) Logic Breaker Tripped Time Position Applicants Actions or Behavior REQUIRED ACTION: (continued) COMPLETION TIME:

AND B.4.1 Determine OPERABLE Unit 3 B.4.1 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months DG(s) are not inoperable due to common cause failure.

OR B.4.2 - Perform SR 3.8.1.1 for B.4.2 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months OPERABLE Unit 3 DG(s).

AND B.5 - Restore Unit 1 and 2 DG to B.5 - 7 days from discovery of OPERABLE status unavailability of TDG(s)

NUSO AND 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from discovery of Condition B entry 6 days concurrent with unavailability of TDG(s)

AND 14 days AND 21 days from discovery of failure to meet LCO In the interest of time management, if the crew begins to perform 0-SR-3.8.1.A.1, Verification of Offsite Power Availability to 4.16 kV NRC Shutdown Boards, inform the crew that another operator will complete the surveillance.

End of Event 6. Request that the driver insert Event 7, High NRC Suppression Pool Water Level / Emergency Depressurization.

Page 16 of 39 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-3 Event No.: 7 Page 1 of 10 Event

Description:

High Suppression Pool Water Level / Emergency Depressurization Time Position Applicants Actions or Behavior When requested by the Chief Examiner, insert Event 7, High Driver Suppression Pool Water Level / Emergency Depressurization.

Event 8, Two Control Rods Fail to Insert and Event 9, 480V Shutdown Board Trip, will occur during Event 7 and are NRC automatically entered by Simulator Setup. No action is required by the driver to insert Event 8 or Event 9.

Acknowledges and reports the following alarms as they are received:

DRYWELL TO SUPPRESSION CHAMBER DIFFERENTIAL BOP PRESSURE ABNORMAL, 3-9-3B, Window 26 SUPPRESSION CHAMBER WATER LEVEL ABNORMAL, 3-9-3B, Window 15 Directs the BOP to respond in accordance with the appropriate Alarm NUSO Response Procedures.

Alarm Response Procedure, 3-ARP-9-3B DRYWELL TO SUPPRESSION CHAMBER DIFFERENTIAL PRESSURE ABNORMAL, Window 26 BOP Operator Action:

A. CHECK alarm by checking Drywell to Suppression Chamber DP.

B. REFER TO 3-OI-64, Primary Containment System.

C. REFER TO Tech Spec Section 3.6.2.6, Drywell-to-Suppression Chamber Differential Pressure.

Due to time constraints, Tech Spec evaluation for this event is not NRC required and should not be used to evaluate the applicants Tech Spec competency.

Page 17 of 39 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-3 Event No.: 7 Page 2 of 10 Event

Description:

High Suppression Pool Water Level / Emergency Depressurization Time Position Applicants Actions or Behavior Alarm Response Procedure, 3-ARP-9-3B SUPPRESSION CHAMBER WATER LEVEL ABNORMAL, Window 15 Operator Action:

A. CHECK level using multiple indications.

B. IF level is low, THEN DISPATCH personnel to check for leaks.

C. IF level is high, THEN CHECK for RCIC, HPCI, Core Spray, or RHR draining to Suppression Pool, and CHECK 3-TR-64-161, SUPPRESSION POOL WATER TEMPERATURE and 3-TR-64-162, SUPPRESSION POOL WATER TEMPERATURE.

BOP D. REFER TO 3-OI-74, Residual Heat Removal System, Sections 8.2, 8.3, and 8.4.

E. REFER TO Tech Spec Section 3.6.2.2, Suppression Pool Water Level.

F. IF level is above -1" or below -6.25" AND NOT in Mode 4 or Mode 5 THEN (otherwise N/A) ENTER 3-EOI-2, Primary Containment Control.

G. IF level is above -1" or below -6.25" AND in Mode 4 or Mode 5 THEN (otherwise N/A)

1. EVALUATE plant conditions to DETERMINE if 3-EOI-2 entry is appropriate.

2. RECORD actions in NOMS log.

Determines that the cause for the Drywell to Suppression Chamber DP BOP alarm is rising Suppression Pool Water Level, and informs the NUSO.

Directs the BOP to monitor Suppression Pool Water Level and to NUSO provide an update when level reaches (-) 1 inch.

Acknowledges and reports the following alarm to the NUSO when BOP received:

SUPPRESSION POOL LEVEL HIGH, 3-9-3F, Window 12 Directs the BOP to respond in accordance with the appropriate Alarm NUSO Response Procedure.

Page 18 of 39 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-3 Event No.: 7 Page 3 of 10 Event

Description:

High Suppression Pool Water Level / Emergency Depressurization Time Position Applicants Actions or Behavior Alarm Response Procedure, 3-ARP-9-3F SUPPRESSION POOL LEVEL HIGH, Window 12 A. CHECK CST 3 and Suppression Pool level using multiple indications.

B. ENSURE HPCI Suction automatically transfers to the Suppression BOP Pool.

C. IF automatic transfer fails, THEN REFER TO 3-OI-73, High Pressure Coolant Injection System.

D. REFER TO Tech Spec 3.5.1, ECCS - Operating, 3.5.2 and 3.6.2.2, Suppression Pool Water Level.

Due to time constraints, Tech Spec evaluation for this event is not NRC required and should not be used to evaluate the applicants Tech Spec competency.

When appropriate, enters 3-EOI-2, Primary Containment Control on high NUSO Suppression Pool Water Level (level above (-) 1 inch).

3-EOI-2, Primary Containment Control NUSO SP/L-1 MONITOR and CONTROL Suppression Pool Water Level (-) 6 inches to (-) 1 inch.

IF THEN NUSO Suppression Pool Water Level CANNOT be maintained below

(-) 1 inch.

Suppression Pool Water Level CANNOT be maintained above NO ACTION REQUIRED

(-) 6 inches.

Page 19 of 39 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-3 Event No.: 7 Page 4 of 10 Event

Description:

High Suppression Pool Water Level / Emergency Depressurization Time Position Applicants Actions or Behavior SP/L-3 MAINTAIN Suppression Pool Water Level below 19 ft.

NUSO (APPX 18, 20K)

SP/L-4 WHEN Suppression Pool Level CANNOT be maintained below (APPX 9) 19 feet STOP DW Sprays Directs the BOP to control Suppression Pool Water Level in accordance NUSO with 3-EOI-Appendix-18, Suppression Pool Water Inventory Removal and Makeup.

If contacted as an AUO to perform any steps locally per 3-EOI-Appendix-18, Suppression Pool Water Inventory Removal Driver and Makeup, acknowledge any direction given.

If contacted as an AUO to verify any valves closed, acknowledge the direction.

3-EOI-Appendix-18, Suppression Pool Water Inventory Removal and Makeup

[1] N/A

[2] N/A BOP [3] IF Directed by the NUSO, THEN REMOVE water from Suppression Pool as follows:

[3.1] DISPATCH personnel to perform the following (Unit 3 RB, Elevation 519 ft, Torus Area):

[3.1.1] VERIFY OPEN 3-SHV-074-0786A (B), RHR DRAIN PUMP A (B) DISCHARGE SHUTOFF VALVE.

Page 20 of 39 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-3 Event No.: 7 Page 5 of 10 Event

Description:

High Suppression Pool Water Level / Emergency Depressurization Time Position Applicants Actions or Behavior

[3.1.2] OPEN the following valves:

3-SHV-074-0564A(B), RHR DRAIN PUMP A(B) SEAL WATER SUPPLY 3-SHV-074-0529A (B), RHR DRAIN PUMP A (B)

SHUTOFF VALVE

[3.1.3] UNLOCK and OPEN 3-SHV-074-0765A (B), RHR DRAIN PUMP A(B) DISCHARGE.

[3.1.4] NOTIFY Unit Operator that RHR Drain Pump 3A (3B) is lined up to remove water from Suppression Pool.

[3.1.5] REMAIN at torus area UNTIL Unit 3 Operator directs starting of RHR Drain Pump 3A (3B).

BOP

[3.2] IF Main Condenser is desired drain path, THEN OPEN 3-FCV-74-62, RHR MAIN CONDENSER FLUSH VALVE.

[3.3] IF Radwaste is desired drain path, THEN PERFORM the following:

[3.3.1] ESTABLISH communications with Radwaste

[3.3.2] OPEN 3-FCV-74-63, RHR RADWASTE SYSTEM FLUSH VALVE.

[3.4] NOTIFY personnel in Unit 3 RB, El 519 ft, Torus Area to start RHR Drain Pump 3A(3B).

[3.5] THROTTLE 3-FCV-74-108, RHR DRAIN PUMP 3A/B DISCHARGE HEADER VALVE, as necessary.

After 2 minutes, report that the outside portions of 3-EOI-Appendix-18, Suppression Pool Water Inventory Removal and Makeup are complete. If directed to start the RHR Drain Pump, Driver report that the RHR Drain Pump has been started.

If contacted as the Rad Waste Operator, acknowledge any reports or direction given.

NUSO SP/L-6 MAINTAIN Suppression Pool Level within the safe area of Curve 4 (APPX 18, 20K)

Page 21 of 39 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-3 Event No.: 7 Page 6 of 10 Event

Description:

High Suppression Pool Water Level / Emergency Depressurization Time Position Applicants Actions or Behavior NUSO SP/L-7 WHEN Suppression Pool Level CANNOT be maintained within the safe area of Curve 4 (APPX 9)

NUSO Critical Task:

When Suppression Pool Water Level cannot be maintained within the safe area of Curve 4, SRV Tail Pipe Level Limit, the crew will insert a manual Reactor SCRAM as directed by the Nuclear Unit Senior Operator (NUSO).

CREW Critical Task Failure Criteria:

The operating crew fails to proceed without delay and in a controlled manner to insert a manual SCRAM before Suppression Pool Water Level enters the Action Required area of Curve 4, SRV Tail Pipe Level Limit.

Page 22 of 39 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-3 Event No.: 7 Page 7 of 10 Event

Description:

High Suppression Pool Water Level / Emergency Depressurization Time Position Applicants Actions or Behavior Directs the OATC to insert a manual Reactor SCRAM and directs the NUSO crew to enter 3-AOI-100-1, Reactor SCRAM.

Event 8, 2 Control Rods Fail to Insert, is automatically entered on NRC Simulator Setup. No action is required by the driver to insert Event 8. See page 27 of 39 for Event 8 actions.

Inserts a manual Reactor SCRAM 3-AOI-100-1, Reactor SCRAM Immediate Actions

[1] DEPRESS 3-HS-99-5A/S3A, REACTOR SCRAM A and OATC 3-HS-99-5A/S3B, REACTOR SCRAM B, on Panel 3-9-5.

[2] PLACE REACTOR MODE SWITCH, 3-HS-99-5A/S1, in SHUTDOWN.

[3] IF all Control Rods can NOT be verified fully inserted, THEN INITIATE ARI. (Otherwise MARK N/A).

OATC Determines that there are two (2) rods out.

When the Reactor MODE SWITCH is placed in SHUTDOWN, the Feedwater Heater Outlet Isolation Valves will close.

NRC See page 30 of 39 for actions for Event 9, 480V Shutdown Board Trip.

[4] IF Reactor Power is 5% or BELOW, THEN (Otherwise MARK N/A)

REPORT the following to the UNIT SRO:

Reactor SCRAM Mode Switch is in Shutdown OATC All rods in or rods out Reactor Water Level and trend (recovering or lowering)

Reactor Pressure and trend MSIV position (Open or Closed)

Power level Page 23 of 39 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-3 Event No.: 7 Page 8 of 10 Event

Description:

High Suppression Pool Water Level / Emergency Depressurization Time Position Applicants Actions or Behavior Following the Reactor SCRAM, enters 3-EOI-1A, ATWS RPV Control and directs the crew to perform the following:

Maintain Reactor Pressure to ensure that Suppression Pool Level is maintained within the safe area of Curve 4 in accordance with NUSO 3-EOI-Appendix-8B, Reopening MSIVs/Bypass Valve Operation Maintain Reactor Water Level using in accordance with 3-EOI-Appendix-5D, Injection System Lineup HPCI or 3-EOI-Appendix-5C, Injection System Lineup RCIC Insert Control Rods (Continuing actions of 3-EOI-2, Primary Containment Control)

SP/L-8 WHEN Suppression Pool Level and RPV Pressure CANNOT be maintained within the safe area of Curve 4 (APPX 9).

SP/L-9 STOP injection into RPV from sources external to Primary Containment EXCEPT from systems required to assure Adequate NUSO Core Cooling or shut down the Reactor SP/L-10 WHEN Suppression Pool Level and RPV Pressure CANNOT be restored and maintained within the safe area of Curve 4 Page 24 of 39 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-3 Event No.: 7 Page 9 of 10 Event

Description:

High Suppression Pool Water Level / Emergency Depressurization Time Position Applicants Actions or Behavior Critical Task:

When Suppression Pool Water Level and RPV Pressure cannot be restored and maintained within the safe area of Curve 4, the NUSO determines that Emergency Depressurization is required. The Operator initiates Emergency Depressurization as directed by the NUSO.

CREW Critical Task Failure Criteria:

The operating crew fails to proceed without delay and in a controlled manner to initiate Emergency Depressurization from the time that Suppression Pool Water Level and RPV Pressure cannot be restored and maintained within the safe area of Curve 4.

Enters 2-C-2, Emergency RPV Depressurization C2-1 IF THEN RPV Water Level CANNOT be NO ACTION REQUIRED determined NUSO It is anticipated that RPV depressurization will result in NO ACTION REQUIRED loss of injection required for Adequate Core Cooling Containment Water Level CANNOT be maintained below NO ACTION REQUIRED 44 feet C2-2 IF Drywell Pressure is above 2.45 psig NUSO PREVENT injection from ONLY those Core THEN Spray and LPCI pumps NOT required to assure Adequate Core Cooling (APPX 4)

Page 25 of 39 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-3 Event No.: 7 Page 10 of 10 Event

Description:

High Suppression Pool Water Level / Emergency Depressurization Time Position Applicants Actions or Behavior C2-3 EMERGENCY DEPRESSURIZE the RPV IF Suppression Pool Water Level is above 5.5 feet THEN OPEN 6 MSRVs (ADS Valves preferred)

OK to exceed 100 F/hr Cooldown Rate IF THEN Drywell Control Air becomes NUSO NO ACTION REQUIRED unavailable Less than 4 MSRVs can be opened AND NO ACTION REQUIRED RPV Pressure is 80 psi or more above Suppression Chamber Pressure The first loop of Low Pressure Injection (Core Spray/RHR) that the NRC crew attempts to use will result in a loss of the 480V Shutdown Board for that loop. See Event 9 on 31.

End of Event 7. When the crew has inserted all Control Rods, Emergency Depressurized the Reactor, and has control of Reactor NRC Water level above the Top of Active Fuel ((-) 162 inches) using low pressure systems, end of Scenario.

Page 26 of 39 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-3 Event No.: 8 Page 1 of 3 Event

Description:

Two Control Rods Fail to Insert Time Position Applicants Actions or Behavior Event 8 is automatically entered by simulator setup. No action is NRC required by the Driver to insert Event 8.

Following the Reactor SCRAM and after initiating Alternate Rod OATC Insertion (ARI), determines that all Control Rods are not in. Informs the NUSO that two rods are out, and that Reactor Power is less than 5%.

Directs the OATC to insert Control Rods in accordance with NUSO 3-AOI-100-1, Reactor SCRAM.

Not all Subsequent Actions of 3-AOI-100-1, Reactor SCRAM, are NRC listed below.

3-AOI-100-1, Reactor SCRAM OATC [16] IF all rods are NOT inserted to Position 02 or beyond, THEN DIRECT Reactor Engineer to commence determination that the Reactor will remain subcritical under all conditions without boron.

If contacted as the Reactor Engineer, acknowledge any direction or report given.

As necessary, insert the following Events when / if the EOI Appendices are requested:

Event 23 - CLOSE 3-85-586, CRD CHARGING WATER Driver VALVE Event 24 EOI-Appendix-1F, Manual SCRAM Event 25 EOI-Appendix-2, Defeating ARI Logic Trips Event 26 EOI-Appendix-8E, Bypassing Group 6 Low RPV Level and High Drywell Pressure Isolation Interlocks Event 27 - OPEN 3-85-586, CRD CHARGING WATER VALVE

[17] IF any Control Rod fails to fully insert and it is required to Re-SCRAM, THEN PERFORM the following, as required. (Otherwise OATC N/A)

[17.1] RESET the SCRAM per Steps 4.2[24] thru 4.2[24.12].

Page 27 of 39 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-3 Event No.: 8 Page 2 of 3 Event

Description:

Two Control Rods Fail to Insert Time Position Applicants Actions or Behavior

[17.2] CHECK WEST and EAST CRD DISCH VOL WTR LVL HIGH HALF SCRAM annunciators (3-XA-55-4A, Window 1 and 3-XA 4A, Window 29) are reset.

[17.3] INITIATE a manual SCRAM.

OATC

[17.4] REPEAT Step 4.2[17], as necessary, as long as rod motion is observed.

[18] IF any Control Rod fails to fully insert and it is required to Drive Control Rods, THEN REFER TO 3-OI-85, Control Rod Drive System.

3-OI-85, Control Rod Drive System Section 6.7, Control Rod Insertion

[1] REVIEW Precautions and Limitations in Sections 3.7 and 3.8.

[2] ENSURE the following prior to Control Rod movement:

3-HS-85-46, CRD POWER in ON ROD WORTH MINIMIZER is operable and LATCHED in to the correct ROD GROUP, when Rod Worth Minimizer is enforcing OATC [3] OBSERVE the following during Control Rod repositioning:

Control Rod reed switch position indicators (four rod display) agree with the indication on the Full Core Display.

Nuclear Instrumentation responds as Control Rods move through the core (This ensures Control Rod is following drive during Control Rod movement.)

[4] PERFORM the following to insert the Control Rod as appropriate.

Control Rod Notch Insertion per Section 6.7.2 Control Rod Continuous Insertion per Section 6.7.3 3-OI-85, Control Rod Drive System Section 6.7.3, Continuous Insertion of Control Rod OATC [1] CHECK Section 6.7.1 has been performed.

[2] SELECT the desired Control Rod by depressing the appropriate 3-XS-85-40, CRD ROD SELECT pushbutton.

Page 28 of 39 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-3 Event No.: 8 Page 3 of 3 Event

Description:

Two Control Rods Fail to Insert Time Position Applicants Actions or Behavior

[3] OBSERVE the following for the selected Control Rod:

CRD ROD SELECT pushbutton is brightly ILLUMINATED White light on the Full Core Display ILLUMINATED OATC [4] PLACE and HOLD CRD CONTROL SWITCH, 3-HS-85-48, in ROD IN.

[5] WHEN Control Rod notch reaches the even rod notch position prior to the desired final Control Rod notch position, THEN RELEASE 3-HS-85-48, CRD CONTROL SWITCH.

When the OATC has selected each stuck Control Rod and begins to drive the rod in, the malfunction will clear to allow rod insertion.

NRC As a result, the Control Rod will drive in at a much faster rate than normal.

[6] OBSERVE the Control Rod settles into desired position and the ROD SETTLE light extinguishes.

[7] IF the Control Rod double notches or inserts past its correct/desired position, THEN with Unit SROs permission return the Control Rod to the intended position per Section 6.6. (Otherwise N/A)

[8] IF the Control Rod moves more than one notch from its intended OATC position, THEN PERFORM 3-AOI-85-7, Mispositioned Control Rod.

(Otherwise N/A)

[9] WHEN Control Rod movement is no longer desired AND deselecting Control Rods is desired, THEN:

[9.1] PLACE 3-HS-85-46, CRD POWER in OFF.

[9.2] PLACE 3-HS-85-46, CRD POWER in ON.

End of Event 8. When the crew has inserted all Control Rods, Emergency Depressurized the Reactor, and has control of Reactor NRC Water Level above the Top of Active Fuel (TAF, (-) 162 inches) using low pressure systems, end of Scenario.

Page 29 of 39 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-3 Event No.: 9 Page 1 of 4 Event

Description:

480V Shutdown Board Trip Time Position Applicants Actions or Behavior Event 9, 480V Shutdown Board Trip, is automatically entered by Simulator Setup. No action is required by the driver to insert Event 9.

NRC The first loop of Low Pressure Injection (Core Spray/RHR) that the crew attempts to use will result in a loss of the 480V Shutdown Board for that loop.

During Emergency Depressurization, directs the BOP to maintain Reactor Water Level using Core Spray or RHR in accordance with any of the following EOI Appendices:

3-EOI-Appendix-6B, Injection Subsystems Lineup - RHR System I LPCI Mode (see below)

NUSO 3-EOI-Appendix-6C, Injection Subsystems Lineup - RHR System II LPCI Mode (see page 32 of 39) 3-EOI-Appendix-6D, Injection Subsystems Lineup -Core Spray System (see page 31 of 39) 3-EOI-Appendix-6E, Injection Subsystems Lineup - Core Spray System II (see page 33 of 39)

If the crew selects Loop II of RHR or Core Spray to maintain Reactor Water Level, proceed to page 32 of 39 for the procedure(s)

NRC for injection. If the crew selects Loop I of RHR see below for the procedure(s) for injection.

IF USING LOOP I OF RHR FOR INJECTION:

3-EOI-Appendix-6B, Injection Subsystems Lineup RHR System I LPCI Mode

[1] IF Adequate core cooling is assured AND it becomes necessary to BOP bypass LPCI Injection Valve auto open signal to control injection, THEN PLACE 3-HS-74-155A, LPCI SYSTEM I OUTBD INJECTION VALVE BYPASS SELECT, in BYPASS.

[2] ENSURE OPEN 3-FCV-74-7, RHR SYSTEM I MINIMUM FLOW VALVE.

Page 30 of 39 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-3 Event No.: 9 Page 2 of 4 Event

Description:

480V Shutdown Board Trip Time Position Applicants Actions or Behavior

[3] ENSURE OPEN the following valves:

3-FCV-74-1, RHR PUMP 3A SUPPRESSION POOL SUCTION VALVE 3-FCV-74-12, RHR PUMP 3C SUPPRESSION POOL SUCTION VALVE

[4] ENSURE CLOSED the following valves:

3-FCV-74-61, RHR SYSTEM I DRYWELL SPRAY INBOARD VALVE BOP 3-FCV-74-60, RHR SYSTEM I DRYWELL SPRAY OUTBOARD VALVE 3-FCV-74-57, RHR SYSTEM I SUPPRESSION CHAMBER/POOL ISOLATION VALVE 3-FCV-74-58, RHR SYSTEM I SUPPRESSION CHAMBER SPRAY VALVE 3-FCV-74-59, RHR SYSTEM I SUPPRESSION POOL COOLING/TEST VALVE

[5] ENSURE RHR Pump 3A and / or 3C running Determines that 3A 480V Shutdown Board has tripped. Proceeds to CREW use the appropriate EOI Appendix for injection with low pressure systems for the opposite loop.

IF USING LOOP I OF CORE SPRAY FOR INJECTION:

3-EOI-Appendix-6D, Injection Subsystems Lineup Core Spray System II

[1] VERIFY OPEN the following valves:

3-FCV-75-2, CORE SPRAY PUMP 3A SUPPRESSION POOL BOP SUCTION VALVE 3-FCV-75-11, CORE SPRAY PUMP 3C SUPPRESSION POOL SUCTION VALVE 3-FCV-75-23, CORE SPRAY SYS I OUTBD INJECTION VALVE

[2] VERIFY CLOSED 3-FCV-75-22, CORE SPRAY SYS I TEST VALVE.

[3] VERIFY Core Spray Pump 3A and/or 3C running.

Page 31 of 39 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-3 Event No.: 9 Page 3 of 4 Event

Description:

480V Shutdown Board Trip Time Position Applicants Actions or Behavior Determines that 3A 480V Shutdown Board has tripped. Proceeds to CREW use the appropriate EOI Appendix for injection with low pressure systems for the opposite loop.

IF USING LOOP II OF RHR FOR INJECTION:

3-EOI-Appendix-6C, Injection Subsystems Lineup RHR System II LPCI Mode

[1] IF Adequate Core Cooling is assured AND, it becomes necessary to bypass LPCI Injection Valve auto open signal to control injection, THEN PLACE 3-HS-74-155B, LPCI SYS II OUTBOARD INJECTION VALVE BYPASS SELECT IN BYPASS.

[2] ENSURE OPEN 3-FCV-74-30, RHR SYSTEM II MINIMUM FLOW VALVE.

[3] ENSURE OPEN the following valves:

3-FCV-74-24, RHR PUMP 3B SUPPRESSION POOL SUCTION VALVE.

BOP 3-FCV-74-35, RHR PUMP 3D SUPPRESSION POOL SUCTION VALVE.

[4] ENSURE CLOSED the following valves:

3-FCV-74-75, RHR SYSTEM II DRYWELL SPRAY INBOARD VALVE 3-FCV-74-74, RHR SYSTEM II DRYWELL SPRAY OUTBOARD VALVE 3-FCV-74-71, RHR SYSTEM II SUPPR CHAMBER/POOL ISOLATION VALVE 3-FCV-74-72, RHR SYSTEM II SUPPR CHAMBER SPRAY VALVE 3-FCV-74-73, RHR SYSTEM II SUPPR POOL COOLING/TEST VALVE

[5] ENSURE RHR Pump 3B and/or 3D running.

Page 32 of 39 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 21-04 Scenario No. NRC-3 Event No.: 9 Page 4 of 4 Event

Description:

480V Shutdown Board Trip Time Position Applicants Actions or Behavior Determines that 3B 480V Shutdown Board has tripped. Proceeds to CREW use the appropriate EOI Appendix for injection with low pressure systems for the opposite loop.

IF USING LOOP II OF CORE SPRAY FOR INJECTION:

3-EOI-Appendix-6E, Injection Subsystems Lineup Core Spray System II

[1] VERIFY OPEN the following valves:

3-FCV-75-30, CORE SPRAY PUMP 3B SUPPRESSION POOL SUCTION VALVE BOP 3-FCV-75-39, CORE SPRAY PUMP 3D SUPPRESSION POOL SUCTION VALVE 3-FCV-75-51, CORE SPRAY SYS II OUTBD INJECTION VALVE

[2] VERIFY CLOSED 3-FCV-75-50, CORE SPRAY SYSTEM II TEST VALVE.

[3] VERIFY CS Pump 3B and/or 3D running.

Determines that 3B 480V Shutdown Board has tripped. Proceeds to CREW use the appropriate EOI Appendix for injection with low pressure systems for the opposite loop.

End of Event 9. When the crew has Emergency Depressurized the Reactor and has control of Reactor Water Level above the Top of NRC Active Fuel (TAF, (-) 162 inches) using low pressure systems, end of Scenario.

Page 33 of 39 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Scenario Setup UNIT 3 IC 38 Exam IC 263 Procedure Revision Procedure Revision Procedure Revision OI-47C 23 ARP-7A 30 APPX 18 3 OI-85 92 ARP-23B 29 TS 3.1.3 212 OI-92A 18 EOI-1 13 TS 3.1.5 212 AOI-85-3 13 EOI-2 13 TS 3.1.4 212 AOI-100-1 74 APPX-6B 7 SR 3.1.3.5(A) 27 ARP-3B 23 APPX-6C 8 GOI-100-1A 133 ARP-3F 37 APPX-6D 4 ARP-5A 54 APPX-6E 4 Verify RWM is properly latched.

Simulator Setup Verify camera system is powered down (admin password = abcd1234)

Start CPERF PRIOR to placing the Simulator in RUN Schedule Files(s): 2104 NRC Scenario 3 UNIT 3.sch Event Files(s): 2104 NRC Scenario 3 UNIT 3.evt Schedule File - 2104 NRC Scenario 3 ES-D-2 UNIT 3.sch Event Action Description 2104 NRC Scenario 3 UNIT 3.evt Insert override XS-92-7/42G to 8 CHANNEL G IRM RANGE SWITCH Insert malfunction RD08R2227 to 4 CRD ACCUMULATOR LOW PRESSURE 22-27 95.00000 Insert malfunction PMP-66-31A to 42_CONTACTOR STACK DILUTION AIR FAN B FAIL_CONTROL_POWER 5 Insert override HS-66-29A to STOP STACK DILUTION FAN 3A 15 Delete malfunction PMP-66-31A 42_CONTACTOR STACK DILUTION AIR FAN B Insert override ZLOHS6631A_1 to On HS-66-31A-Green* STACK DILUTION FAN 3B Page 34 of 39 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Schedule File - 2104 NRC Scenario 3 ES-D-2 UNIT 2.sch Event Action Description Insert override ZLOHS6631A_2 to Off HS-66-31A-RED STACK DILUTION FAN 3B 15 Delete override ZLOHS6631A_1 HS-66-31A-Green* STACK DILUTION FAN 3B 15 Delete override ZLOHS6631A_2 HS-66-31A-RED STACK DILUTION FAN 3B 5 Insert override ZLOHS6629A_1 to Off HS-66-29A-GREEN STACK DILUTION FAN 3A 6 Insert remote DG01A to OPEN UNIT 3 DIESEL GENERATOR 3A LOGIC BREAKER ZI-74-34-Green* RHR PUMP 3B CST SUCTION Insert override ZLOZI7434_1 to On VALVE ZI-74-45-Green* RHR PUMP 3D CST SUCTION Insert override ZLOZI7445_1 to On VALVE Insert override ZLOZI7411_1 to On ZI-74-11-Green* RHR PUMP 3A CST SUCTION VLV ZI-75-31-Green* CORE SPRAY PUMP 3B CST Insert override ZLOZI7531_1 to On SUCTION VLV ZI-75-40-Green* CORE SPRAY PUMP 3D CST Insert override ZLOZI7540_1 to On SUCTION VLV ZI-75-12-Green* CORE SPRAY PUMP 3C CST Insert override ZLOZI7512_1 to On SUCTION VALVE ZI-75-3-Green* CORE SPRAY PUMP 3A CST Insert override ZLOZI753_1 to On SUCTION VALVE RHR PUMP B CONDENSATE SUCTION VALVE 7 Insert remote RH07 to OPEN HCV-74-34 RHR PUMP D CONDENSATE SUCTION VALVE 7 Insert remote RH08 to OPEN HCV-74-45 Insert remote RH05 after 180 to RHR PUMP A CONDENSATE SUCTION VALVE 7

OPEN HCV-74-11 Insert remote RH06 after 180 to RHR PUMP C CONDENSATE SUCTION VALVE 7

OPEN HCV-74-23 ALIGN CONDENSATE STORAGE TANK TO CORE 7 Insert remote CS06A to ALIGN SPRAY LOOP 1 ALIGN CONDENSATE STORAGE TANK TO CORE 7 Insert remote CS06B to ALIGN SPRAY LOOP 2 7 Insert remote FW11 to XCON CROSS CONNECT CSTS Page 35 of 39 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Schedule File - 2104 NRC Scenario 3 ES-D-2 UNIT 2.sch Event Action Description ALIGN CONDENSATE STORAGE TANK TO CORE 17 Insert remote CS06A to NORM SPRAY LOOP 1 ALIGN CONDENSATE STORAGE TANK TO CORE 17 Insert remote CS06B to NORM SPRAY LOOP 2 Insert remote CS06A after 120 to ALIGN CONDENSATE STORAGE TANK TO CORE 17 ALIGN SPRAY LOOP 1 Insert remote CS06B after 120 to ALIGN CONDENSATE STORAGE TANK TO CORE 17 ALIGN SPRAY LOOP 2 Insert malfunction RD06R1023 STICK ANY CONTROL ROD 10-23 Insert malfunction RD06R1443 STICK ANY CONTROL ROD 14-43 18 Delete malfunction RD06R1023 STICK ANY CONTROL ROD 10-23 28 Delete malfunction RD06R1443 STICK ANY CONTROL ROD 14-43 23 Insert remote RD06 to CLOSE after 60 CRD CHARGING WATER VALVE FCV-3-85-586 Schedule F:\Scenarios\Scenario 3\App.

24 1F.sch Schedule F:\Scenarios\Scenario 3\App.

25 2.sch Schedule F:\Scenarios\Scenario 3\App.

26 8E.sch 27 Insert remote RD06 to OPEN after 60 CRD CHARGING WATER VALVE FCV-3-85-586 19 Insert malfunction ED10A after 3 480V SHUTDOWN BOARD 3A FAILURE 20 Insert malfunction ED10A after 3 480V SHUTDOWN BOARD 3A FAILURE 21 Insert malfunction ED10B after 3 480V SHUTDOWN BOARD 3B FAILURE 22 Insert malfunction ED10B after 3 480V SHUTDOWN BOARD 3B FAILURE Insert override HS-3-75A after 30 to 17 HP HTR 3A1 FW OUTLET ISOL VLV CLOSE Insert override HS-3-76A after 30 to 17 HP HTR 3B1 FW OUTLET ISOL VLV CLOSE Insert override HS-3-77A after 30 to 17 HP HTR 3C1 FW OUTLET ISOL VLV CLOSE Page 36 of 39 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Schedule File: APP. 1F.sch Event Action Description Insert remote RP13A to BYP after 120 DEFEAT CHANNEL A1 AUTO SCRAM Insert remote RP13B to BYP after 120 DEFEAT CHANNEL B1 AUTO SCRAM Insert remote RP13C to BYP after 120 DEFEAT CHANNEL A2 AUTO SCRAM Insert remote RP13D to BYP after 120 DEFEAT CHANNEL B2 AUTO SCRAM Schedule File: APP. 2.sch Event Action Description Insert remote RP12A to TEST after OPERATE LOCAL ATWS MODE SWITCH 180 3-HS-68-118A Insert remote RP12B to TEST after OPERATE LOCAL ATWS MODE SWITCH 180 3-HS-68-118B Schedule File: APP. 8E.sch Event Action Description BYPASS GROUP 6 INBOARD LOW LVL/HI DW Insert remote RP14A to BYP after 300 PRESS INTLKS BYPASS GROUP 6 OUTBOARD LOW LVL/HI Insert remote RP14B to BYP after 300 DW PRESS INTLKS Page 37 of 39 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Event File List Details Page 38 of 39 Unit 3

UNIT 3 SHIFT TURNOVER MEETING Today DAYS ON LINE Protected Equipment Drywell Leakage (GPM)

MODE 2 0 None PRA (EOOS) -Green 1.89 Rx Power 500Kv GRID - Qualified Floor Drain (GPM) 2.2% 161Kv Grid -Qualified 0.31 MWe Last breaker closure Equipment Drain (GPM) 0 N/A 1.58 Review logs Qualifications Review RCP/Rx Brief Review LCO/OWA Actions Walkdown Panels/Verify EOOS CR Reviews Complete Leadership and Team Effectiveness CHANGES IN LCOs IRM H bypassed due to noise. Tech Spec 3.3.1.1 (Information only)

LCOs OF 72 HOURS OR LESS SIGNIFICANT ITEMS DURING PREVIOUS SHIFT/RADIOLOGICAL CHANGES Reactor Startup MAJOR EQUIPMENT CHANGES PLANNED FOR THIS SHIFT After assuming the shift, Transfer Seal Steam from Auxiliary Steam to Main Steam in accordance with 3-OI-47C, Seal Steam System, Section 6.1. RPHP is in effect.

Continue the Reactor Startup in accordance with 3-GOI-100-1A, Section 5.4[83]. Begin with Group 25 Rods, starting with Control Rod 26-35.

Contact the Ops Superintendent prior to placing the MODE SWITCH in RUN.

Thunderstorm watch was just issued for counties in North Alabama.

OPERATOR WORK AROUNDS OWAs - 0 Burdens - 0 Challenges - 0 ODMIs/ACMPs ONEAs FIRE RISK SIGNIFICANT ITEMS OOS/FPLCO Actions Due SCHEDULED ITEMS NOT COMPLETED Page 39 of 39 Unit 3