Text

2015-12 Draft Operating Test
	ML16012A339
Person / Time
Site:	Fort Calhoun
Issue date:	12/14/2015
From:	Vincent Gaddy; Operations Branch IV
To:	;
References
	Download: ML16012A339 (460)
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Appendix C JPM WORKSHEET Form ES-C-1 Facility: FCS JPM # NRC SA5 Task # 1453 K/A # 2.4.41 2.9 / 4.6

Title:

Classify an Emergency Plan Event Examinee (Print):

Testing Method:

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

Alternate Path: Plant:

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

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

Initial Conditions: Given the following conditions:

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

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

Aircraft arrival time is estimated at 60 minutes.

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

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

Emergency Classification Level __________

IC/EAL Classification __________

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

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

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

Validation Time: 5 minutes Completion Time: ________ minutes Critical Time limit: 15 minutes Comments:

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

Page 1 of 4 NRC Admin JPM SA5 Rev. 4

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

PROVIDE the examinee with a copy of:

EPIP-OSC-1, Emergency Plan.

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

NOTE:

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

Page 2 of 4 NRC Admin JPM SA5 Rev. 4

Appendix C JPM STEPS Form ES-C-1

- Check Mark Denotes Critical Step CRITICAL START TIME:

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

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

Perform Step: 1 DETERMINE the Event Category.

Standard: REFERRED to FCS Emergency Action Levels:

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

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

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

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

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

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

Comment: SAT UNSAT Examiner Note: Declaration shall be made within 15 minutes of start time of JPM.

Perform Step: 4 Classify the event.

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

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

Terminating Cue: This JPM is complete.

Comment: SAT UNSAT CRITICAL STOP TIME:

Page 3 of 4 NRC Admin JPM SA5 Rev. 4

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

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

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

Aircraft arrival time is estimated at 60 minutes.

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

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

Emergency Classification Level __________

IC/EAL Classification __________

THIS IS A TIME CRITICAL JPM Page 4 of 4 NRC Admin JPM SA5 Rev. 4

Appendix D Scenario Outline Form ES-D-1 Facility: Fort Calhoun Station Scenario No.: 1 Op Test No.: Dec 2015 NRC Examiners: Operators:

Initial Conditions: 100% power MOL - RCS Boron is 482 ppm (by sample).

Turnover: Maintain steady-state power conditions. Chemistry requests two Charging Pumps be placed in service per OI-CH-1, CVCS System Normal Operation.

Critical Tasks:

Manually Actuate Pressurizer Pressure Low Signal (PPLS) when RCS Pressure 1600 psia and before 1350 psia to ensure SIAS / VIAS / CIAS Activation. (Event 8)

Stop All Reactor Coolant Pumps (RCPs) when Subcooling is less than 20°F due to Loss of Net Positive Suction Head (NPSH) per RCP NPSH Curve. (Event 6)

Commence a Cooldown and Depressurization of the Reactor Coolant System to Reestablish RCS Inventory Control while maintaining RCS Heat Removal. (Event 6)

Event No. Malf. No. Event Type* Event Description 1 N (ATCO) Raise Charging and Letdown Flow per OI-CH-1, CVCS System

+15 min Normal Operation, Attachment 3.

2 C (ATCO, CRS) Component Cooling Water (CCW) Pump Trip.

+25 min TS (CRS) Start Either Standby CCW Pump.

3 C (BOPO, CRS) Plant Air System Leak.

+35 min Start Instrument Air Compressors.

4 I (ATCO, CRS) Pressurizer Pressure Control Channel PT-103X Fails to 2150 psia

+45 min TS (CRS) on 15 Minute Ramp. Transfer Pressure Control to PT-103Y.

5 R (ATCO) Condenser Evacuation Pump Trip with Auto Start Failure.

+55 min C (BOPO, CRS) Partial Loss of Condenser Vacuum. Reduce Turbine Load.

6 M (ATCO, BOPO, Inadvertent Main Turbine Trip.

+55 min CRS) Pressurizer Safety Valve Fails 50% Open on Reactor Trip.

7 C (BOPO) Total Loss of Condenser Vacuum.

+55 min Place HCV-1040, Atmospheric Dump Valve in Service.

8 I (ATCO) Pressurizer Pressure Low Signal Actuation Failure.

+65 min Manually Initiate Safety Injection.

9 C (ATCO) Low Pressure Safety Injection (LPSI) Pumps Start Failure.

+65 min Manually Start LPSI Pumps.

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Technical Specifications Actual Target Quantitative Attributes 3 Malfunctions after EOP entry (1-2) 5 Abnormal events (2-4) 1 Major transients (1-2) 1 EOPs entered/requiring substantive actions (1-2) 0 EOP contingencies requiring substantive actions (0-2) 3 Critical tasks (2-3)

NRC Simulator Scenario 1 Outline Rev. 6

Scenario Event Description NRC Scenario 1 SCENARIO

SUMMARY

NRC 1 The crew will assume the shift at 100% power per OP-4, Load Change and Normal Power Operation.

The scheduled activity is to start a second Charging Pump per OI-CH-1, CVCS System Normal Operation, Attachment 3, Raising Charging and Letdown Flows per Chemistry request.

The next event is a Component Cooling Water Pump Trip with auto start failure of the standby pumps.

The crew enters AOP-11, Loss of Component Cooling Water, and restores flow by starting either CCW Pump AC-3A or AC-3B. The SRO will refer to Technical Specification LCO 2.4(1) - Component Cooling Water Pump.

The next event is a Plant Air System leak and entry into AOP-17, Loss of Instrument Air, is required.

Crew should recognize that the Control Room Standby Instrument Air Compressor is not loading (ammeter at 0) and start a 3rd Air Compressor. Procedure exit occurs when the Plant Air System is locally isolated from the Instrument Air System.

When plant conditions are stable, Pressurizer Pressure Control Channel, PT-103X, will fail to 2150 psia over 15 minutes. Operator actions are per ARP-CB-1/2/3/A4, Windows A-4 & B-4, PRESSURIZER PRESSURE OFF NORMAL HI-LO CHANNEL X & CHANNEL Y. The crew will transfer to the standby channel PT-103Y and restore Reactor Coolant System (RCS) pressure. The SRO will refer to Technical Specification LCO 2.10.4 - DNBR Margin during Power Operation above 15% of Rated Power.

The next event is a partial Loss of Condenser Vacuum. The crew enters AOP-26, Turbine Malfunctions,Section I, Loss of Vacuum. Actions include starting a Condenser Evacuation Pump and transitioning to AOP-05, Emergency Shutdown, to lower Turbine load and restore Condenser vacuum. When power has been reduced 3% to 5%, an inadvertent Main Turbine trip will occur.

The inadvertent Main Turbine trip results in lifting of a Pressurizer Safety Valve resulting in a Small Break Loss of Coolant Accident (Vapor Space LOCA). The crew enters EOP-00, Standard Post Trip Actions, and manually actuates Safety Injection when it is determined that a Pressurizer Pressure Low Signal Actuation failure has occurred. When Diagnostic Actions are completed at the end of EOP-00, a transition will be made to EOP-03, Loss of Coolant Accident. Two Reactor Coolant Pumps are secured while in EOP-00 when pressure drops to 1350 psia. Eventually all RCPs will be secured due to a loss of subcooling (< 20°F). Upon entry into EOP-03, Containment Cooling Fans VA-7C and VA-7D will need to be started. Containment pressure remains less than 3 psig throughout the event.

The event is complicated by total Loss of Condenser Vacuum which will require placing the Atmospheric Steam Dump Valve, HCV-1040 in service and manual starting of the Low Pressure Safety Injection Pumps due to an automatic start failure.

This scenario is terminated when a cooldown and depressurization is commenced while in EOP-03 using HR-12, Secondary Heat Removal Operation, and PC-11, Pressure Control.

Risk Significance:

Failure of risk important system prior to trip: Loss of CCW Pump

Risk significant core damage sequence: Small Break LOCA Safety Injection Actuation Failure

Risk significant operator actions: Manually Actuate Safety Injection Stop RCPs Upon Loss of Subcooling Cooldown and Depressurize RCS NRC Simulator Scenario 1 Outline Rev. 6

Scenario Event Description NRC Scenario 1 BOOTH OPERATOR INSTRUCTIONS for SIMULATOR SETUP RESET to IC- #1 (or any 100% MOL IC) and LOAD & EXECUTE NRC 1.sce for NRC Scenario 1.

Preset Item - Event 2 - Block Autostart of Non-running CCW Pumps Type Item Value Condition Expert CCAAFU_STDBY_AC_3BCC 1 Scenario Event: AC-3B (AC-3B standby fuse failure) Stbyfuse blown CCBPFU_STDBY_AC_3ACC 1 Scenario Event: AC-3A (AC-3A standby fuse failure) Stby Fuse blown Preset Item - Event 3 - Block Autostart of CA-1B Type Item Value Condition Remote REM:CA1B_3SS (CA-1B control Off (value = 3) Scenario Event: Block start selector switch) of CA-1B Preset Item - Event 5 - Block Auto Start of Condenser Evacuation Pump FW-8C Type Item Value Condition Expert CEACWL_CLTVSP Triggered Scenario Event: block start FW-8C Preset Item - Event 8 - PPLS Fail to Actuate Type Item Value Condition Malfunction ESF07 (PPLS Actuation - Train A) Block Scenario Event: PPLS auto ESF08 (PPLS Actuation - Train B) Block fail Preset Item - Event 9 - LPSI Pumps Fail to Automatically Start Type Item Value Condition Expert ESEARL62_2_1X_SI_1BTVSP Deenergized Scenario Event: LPSI fail ESEBRL62_2_2X_SI_1BTVSP Deenergized to start ESCBRL62_1_2X_SI_1ATVSP Deenergized ESCARL62_1_1X_SI_1ATVSP Deenergized Event 2 - CCW Pump AC-3C Trips Type Item Value Condition Malfunction BUS_1B3C_4C_4_BKR_TRIP trip When directed by examiner, (CCW pump AC-3C breaker fail to trigger/activate this event.

the trip position) Scenario Event: CCW Pump AC-3C Trip Event 3 - Plant Air Leak Type Item Value Condition Malfunction CAS02C (Plant Air Leak) 25 When directed by examiner, trigger/activate this event.

Scenario Event: Plant Air Leak Remote REM:CAS_CA630 0 When directed to close CA-REM:CAS_PCV1753 0 121 to isolate the instrument air leak, trigger/activate this event. Scenario Event:

When directed to close CA121 NRC Simulator Scenario 1 Outline Rev. 6

Scenario Event Description NRC Scenario 1 Event 4 - Pressurizer Pressure Transmitter PT-103X Fails High Type Item Value Condition Transmitter RCS_PT103X 2150 When directed by examiner, Ramp: 900 seconds trigger/activate this event.

Scenario Event: PT-103x fail high Event 5 - Running Condenser Evacuation Pump Trips, Degrading Condenser Vacuum Type Item Value Condition Malfunction CES06 (Condenser Evacuation FW- Trip When directed by examiner, 8B Pump trips) trigger/activate this event.

CND01 (Loss of Main Condenser 3%, ramp = 60 sec Scenario Event: Cond Vacuum) Evac trip Event 6 - Inadvertent Trip, Pressurizer Safety Valve Opens Type Item Value Condition Remote REM:86-1/G1-TRP (relay 86-1/G1 Trip When directed by examiner, fail to trip position) trigger/activate this event.

REM: 86-2/G1-TRP (relay 86-2/G1 Trip Scenario Event: Trip, fail to trip position) safety valve open Malfunction RCS_RC141 (safety valve RC-141) After reactor trip, value = 50, ramp =

15 seconds, delay =

5 seconds Event 7 - Total Loss of Condenser Vacuum Type Item Value Condition Malfunction CND01 (Loss of Main Condenser 100%, 300 second 60 seconds after reactor trip, Vacuum) ramp automatically trigger/activate event:

Complete Loss of Cond Vacuum NRC Simulator Scenario 1 Outline Rev. 6

Scenario Event Description NRC Scenario 1 Booth Operator: INITIALIZE to IC-1 and LOAD NRC 1.sce.

ENSURE all Simulator Annunciator Alarms are ACTIVE.

ENSURE all Control Board Tags are removed.

ENSURE CH-1B, Charging Pump is running.

ENSURE AC-3C, Component Cooling Water Pump running.

ENSURE Channel X Pressurizer Pressure and Level selected.

ENSURE FW-8B, Condenser Evacuation Pump running.

ENSURE Reactivity Briefing Sheet printout provided with Turnover.

ENSURE Middle-of-Life Thumb Rule Sheet provided with Turnover.

ENSURE Containment Pressure Relief (CPR) is secured.

ENSURE procedures in progress provided to crew in Briefing Room:

- COPY of OI-CH-1, Chemical and Volume Control System Normal Operation, Attachment 3, Raising Charging and Letdown Flows, INITIALED through Step 2.i.

Control Room Annunciators in Alarm:

NONE 0B Procedure List Event 1: OP-4, Load Change and Normal Power Operation Event 1: OI-CH-1, CVCS System Normal Operation, Attachment 3, Raising Charging and Letdown Flows Event 2: AOP-11, Loss of Component Cooling Water Event 3: AOP-17, Loss of Instrument Air Event 4: ARP-CB-1/2/3/A4, Windows A-4 & B-4, PRESSURIZER PRESSURE OFF NORMAL HI-LO CHANNEL X & CHANNEL Y Event 5: AOP-26, Turbine Malfunctions,Section I, Loss of Vacuum Event 6: EOP-00, Standard Post Trip Actions Event 6: EOP-03, Loss of Coolant Accident Event 6: HR-12, Secondary Heat Removal Operation Event 6: PC-11, Pressure Control NRC Simulator Scenario 1 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 1 Page 6 of 29 Event

Description:

Raise Charging and Letdown Flow Time Position Applicants Actions or Behavior Booth Operator: When directed, RESPOND to requests from Control Room. Report back that plant conditions requested are normal unless otherwise scripted.

Indications Available:

NONE Examiner Note: The following steps are from OI-CH-1, Chemical and Volume Control System Normal Operation, Attachment 3, Raising Charging and Letdown Flows.

+1 min ATCO START the selected Charging Pump CH-1B. [Step 3]

PLACE CH-1B switch to START.

NOTES

1. PIC-210 Letdown Press Cntrlr should be continuously monitored while adjusting letdown flow.

2. Steps 4 and 5 may be performed concurrently without the procedure in hand. Sign-offs may be completed after these steps are performed.

RAISE bias on HIC-101-1/101-2, Letdown Throttle Valves Controller, and ATCO OBSERVE an increase in Letdown flow. [Step 4]

ROTATE HIC-101-1/101-2 in COUNTERCLOCKWISE direction to increase Letdown flow.

Examiner Note: It is acceptable to place letdown pressure control and flow control in manual or automatic control during rotation of charging pumps.

ADJUST PIC-210, Letdown Press Controller as necessary to maintain ATCO Letdown pressure approximately 300 psig. [Step 5]

Continue to ADJUST bias on HIC-101-1/101-2 until Pressurizer level is ATCO STABILIZED at the programmed setpoint. [Step 6]

When Letdown flow is stable, PROCEED to Event 2.

NRC Simulator Scenario 1 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 2 Page 7 of 29 Event

Description:

Component Cooling Water Pump Trip Time Position Applicants Actions or Behavior Booth Operator: When directed, EXECUTE Event 2.

When contacted to report pump conditions, Auxiliary Building Operator reports normal conditions. Water Plant Operator reports breaker tripped on overcurrent Indications Available:

CB-1/2/3/A2 - CCW PUMPS TRIP CB-1/2/3/A2 - CC WATER FROM DISCH HEADER FLOW LO CB-1/2/3/A2 - CCW PUMPS AC-3A/B/C STANDBY START CB-1/2/3/A2 - AUXILIARY COOLANT FROM CRDM FLOW LO CCW Pump AC-3C white TRIP and green STOP lights lit Multiple loss of CCW flow alarms

+30 sec ATCO RESPOND to Annunciator Response Procedures.

ATCO INFORM CRS of CCW Pump AC-3C trip with NO auto start of standby pump.

Examiner Note: ATCO may Operate to Mitigate per OPD 3-01 and START a CCW Pump.

CRS REFER to AOP-11, Loss of Component Cooling Water.

Examiner Note: The following steps are from AOP-11, Loss of Component Cooling Water.

ATCO VERIFY normal CCW/RW System operation: [Step 4.1]

START CCW Pump AC-3A or AC-3B. [Step 4.1.a]

VERIFY CCW System pressure 60 psig. [Step 4.1.b]

DETERMINE AC-1B, Raw Water CCW Heat Exchanger in service.

[Step 4.1.c]

DETERMINE RCP Coolers CCW Valves, HCV-438A/B/C/D all OPEN.

[Step 4.1.d]

ATCO VERIFY Raw Water Pump operating. [Step 4.2]

ATCO If CCW Surge Tank level < 42 inches, FILL the CCW Surge Tank: [Step 4.3]

OPEN LCV-2801, CCW Surge Tank Makeup Valve, to refill CCW Surge Tank. [Step 4.3.a]

PLACE LCV-2801 in CLOSE or AUTO. [Step 4.3.b]

NRC Simulator Scenario 1 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 2 Page 8 of 29 Event

Description:

Component Cooling Water Pump Trip Time Position Applicants Actions or Behavior CRS EVALUATE Technical Specification LCO 2.4, Containment Cooling

LCO 2.4.(1).a - Component Cooling Water Pump AC-3C

CONDITION 2.4.(1).a - Component Cooling Water Pump AC-3C inoperable.

ACTION 2.4.(1).b - RESTORE Component Cooling Water Pump AC-3C within 7 days OR PLACE Reactor in HOT SHUTDOWN condition within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

When Technical Specifications are addressed, PROCEED to Event 3.

NRC Simulator Scenario 1 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 3 Page 9 of 29 Event

Description:

Plant Air System Leak Time Position Applicants Actions or Behavior Booth Operator: When directed, EXECUTE Event 3.

- Plant Air System leak @ 25%.

Indications Available:

CB-10,11/A21 - PLANT AIR PRESS LO PI-1700, Plant Air Press lowering on CB-10,11

+30 sec BOPO RESPOND to Annunciator Response Procedures.

BOPO INFORM CRS of Plant Air System pressure less than 96 psig and lowering.

Examiner Note: BOPO may Operate to Mitigate per OPD 3-01 and START an Air Compressor.

CRS REFER to AOP-17, Loss of Instrument Air.

Examiner Note: The following steps are from AOP-17, Loss of Instrument Air.

BOPO ENSURE all available Air Compressors start. [Step 4.1]

START Air Compressor CA-1A.

START Air Compressor CA-1B.

Booth Operator: If contacted, REPORT Compressors, Dryers, and Filters appear to be operating normally.

Booth Operator: If contacted, PLACE standby Air Compressor CA-1B in service.

CONTACT Equipment Operator to ensure proper operation of Instrument Air BOPO Compressors, Dryers, and Filters. [Step 4.2]

ANNOUNCE and REPEAT message using Plant Communication System:

CREW

[Step 4.3]

"Attention all personnel, attention all personnel; there is a plant air leak in progress. Report any large air usage to the Control Room."

CRS DIRECT available operators to search for source of air leakage. [Step 4.4]

NRC Simulator Scenario 1 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 3 Page 10 of 29 Event

Description:

Plant Air System Leak Time Position Applicants Actions or Behavior Booth Operator: When contacted, REPORT leak is downstream of PCV-1753. When directed, execute simulator operation to isolate leak and report CA-121, Service Air Supply System Manual Isolation Valve is CLOSED.

DETERMINE Instrument Air pressure is < 80 psig, and CONTACT BOPO Equipment Operator to VERIFY PCV-1753, Service Air System Automatic Isolation Valve CLOSED. [Step 4.5]

DETERMINE Instrument Air pressure slowly returning to normal after service CRS air was isolated. [Step 4.6]

VERIFY CA-121, Service Air Supply System Manual Isolation Valve is closed. [Step 4.6.a]

GO TO Section 5.0, Exit Conditions. [Step 4.6.b]

Examiner Note: Plant Air System remains isolated for the duration of the Scenario.

When Instrument Air pressure returns to normal, PROCEED to Event 4.

NRC Simulator Scenario 1 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 4 Page 11 of 29 Event

Description:

Pressurizer Pressure Control Channel Failure Time Position Applicants Actions or Behavior Booth Operator: When directed, EXECUTE Event 4.

- Pressurizer Pressure Control Channel PT-103X fails to 2150 psia on 15 minute ramp.

Indications Available:

CB-1/2/3/A4 - PRESSURIZER PRESSURE OFF NORMAL HI-LO CHANNEL Y (1st alarm)

CB-1/2/3/A4 - PRESSURIZER PRESSURE OFF NORMAL HI-LO CHANNEL X (2nd alarm ~ 2 min later)

Examiner Note: Due to the nature of this failure, Channel Y alarm comes in 1st as it senses PZR pressure < 2080 psia (alarm setpoint) even though Channel X is the Controlling Channel. As the Channel X setpoint failure ramps in and reaches

> 2145 psia (alarm setpoint), Channel X annunciator will alarm.

+30 sec ATCO RESPOND to Annunciator Response Procedures.

Examiner Note: ATCO may Operate to Mitigate per OPD 4-09 and TRANSFER to Channel Y.

Examiner Note: The following steps are from ARP-CB-1/2/3/A4, Window A-4 for Channel X.

ATCO VERIFY RCS pressure using all available indications. [Step 1]

MONITOR Pressurizer Pressure and operation of PC-103X. [Step 1.1]

DETERMINE PC-103X is not controlling pressure and PLACE HC-103, Pressurizer Pressure Channel Selector Switch to CHAN Y position. [Step 1.1.1]

MAY REFER to OI-RC-7, Reactor Coolant System Pressure Control CRS Normal Operation. [Step 1.1.2]

ATCO PERFORM the following for the low pressure condition: [Step 2]

REFER to Technical Specification LCO 2.10.4.(5) if pressure 2075 CRS psia. [Step 2.1]

DETERMINE Pressurizer Spray Valves PCV-103-1 and PCV-103-2 are ATCO CLOSED. [Step 2.2]

ENSURE all Pressurizer Heater Control Switches in AUTO or ON.

ATCO

[Step 2.3]

ATCO

ENERGIZE additional Pressurizer Heaters as required. [Step 2.4]

DETERMINE Pressurizer level NOT lowering on LR-101X/LR-101Y.

ATCO

[Step 2.5]

NRC Simulator Scenario 1 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 4 Page 12 of 29 Event

Description:

Pressurizer Pressure Control Channel Failure Time Position Applicants Actions or Behavior ATCO

VERIFY VCT level trend on LI-219. [Step 2.6]

CRS EVALUATE Technical Specification LCO 2.10.4, Power Distribution Limits

LCO 2.10.4.(5) - DNBR Margin during Power Operation above 15% of Rated Power

CONDITION 2.10.4.(5).(a).(ii) - Pressurizer Pressure < 2075 psia.

ACTION 2.10.4.(5).(b) - RESTORE Pressurizer Pressure within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or REDUCE power to less than 15% of rated power within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

When Technical Specifications have been addressed, PROCEED to Event 5.

NRC Simulator Scenario 1 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 5 Page 13 of 29 Event

Description:

Partial Loss of Condenser Vacuum / Condenser Evacuation Pump Trip With Auto Start Failure Time Position Applicants Actions or Behavior Booth Operator: When directed, EXECUTE Event 5

- Partial Loss of Condenser vacuum @ 5% on 3 minute ramp.

- Condenser Evacuation Pump FW-8B trip.

- Condenser Evacuation Pump FW-8C Auto Start failure.

Examiner Note: rate of lowering condenser vacuum may be modified at your discretion to advance or retard the pace of this and the next event.

Indications Available:

CB-10,11/A9 - VACUUM PUMP B STOPPED OR SEAL WATER TEMP HI Emergency Response Facility Computer System (ERFCS) Alarm on Low Condenser Vacuum Condenser Evacuation Pump FW-8B green STOP light lit Lowering Condenser Vacuum on PI-925A/B or P0976A/B

+30 sec BOPO RESPOND to Annunciator Response Procedures.

INFORM CRS of lowering Condenser vacuum and Condenser Evacuation BOPO Pump FW-8B trip.

Examiner Note: BOPO may Operate to Mitigate per OPD 4-09 and START FW-8C.

CRS REFER to AOP-26, Turbine Malfunctions,Section I, Loss of Vacuum.

Examiner Note: The following steps are from AOP-26, Turbine Malfunctions,Section I, Loss of Vacuum.

MONITOR Condenser vacuum on ERF Computer System/PI-925A/

BOPO PI-925B/P0976A/P0976B. [Step 4.1]

BOPO ENSURE all Condenser Evacuation Pumps are running. [Step 4.2]

START FW-8C, Condenser Evacuation Pump.

CAUTION The Turbine should not be operated with a Generator load of less than 150 MW when vacuum is less than or equal to 23.85" Hg (ERF, P0976A/B) or 6.07" Hg absolute (PI-925A/B) due to possible overheating of final stage blades.

If Condenser vacuum is < 25" Hg or 4.92" Hg Absolute, COMMENCE a plant CRS shutdown to restore vacuum per AOP-05 Emergency Shutdown. [Step 4.3]

NRC Simulator Scenario 1 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 5 Page 14 of 29 Event

Description:

Partial Loss of Condenser Vacuum / Condenser Evacuation Pump Trip With Auto Start Failure Time Position Applicants Actions or Behavior Examiner Note: The following steps are from AOP-05, Emergency Shutdown.

NOTE TDB-III-23a and the Power Ascension/Power Reduction Strategy (PAPRs) provide guidance for the shutdown.

CRS CONTACT Reactor Engineer if additional guidance is required. [Step 4.1]

NOTE Operation of more than one Charging Pump will raise the rate of the power reduction.

Examiner Note: Unless directed, boration will occur from the Safety Injection Refueling Water Tank (SIRWT) when in AOP-05 to avoid time constraints.

If borating from SIRWT, COMMENCE boration by performing the following:

CRS

[Step 4.2]

DETERMINE Charging Pump CH-1A is RUNNING.

ATCO

[Step 4.2.a]

OPEN LCV-218-3, Charging Pump Suction SIRWT Isolation Valve.

ATCO

[Step 4.2.b]

ATCO

CLOSE LCV-218-2, VCT Outlet Valve. [Step 4.2.c]

CRS DETERMINE Boration alignment from CVCS NOT required. [Step 4.3]

CRS NOTIFY Energy Marketing of power reduction. [Step 4.4]

NOTE During the power reduction, maintain TC PER TDB Figure III.1, Tave Program.

MAINTAIN RCS Temperature Control via Turbine Load per HR-12, BOPO Secondary Heat Removal Operation: [Step 4.5]

MAINTAIN TCOLD 527°F to 547°F AND

MAINTAIN TCOLD +0°F to -1°F of program.

NRC Simulator Scenario 1 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 5 Page 15 of 29 Event

Description:

Partial Loss of Condenser Vacuum / Condenser Evacuation Pump Trip With Auto Start Failure Time Position Applicants Actions or Behavior MAINTAIN Pressurizer Level via Charging and Letdown per IC-11, Inventory ATCO Control: [Step 4.6]

MAINTAIN Pressurizer Level 45% to 60% AND

MAINTAIN Pressurizer Level within 4% of program.

When power has been lowered 3% to 5%, PROCEED to Events 6, 7, 8, and 9.

NRC Simulator Scenario 1 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 6, 7, 8, & 9 Page 16 of 29 Event

Description:

Inadvertent Main Turbine Trip / Pressurizer Safety Valve Failure / Loss of Condenser Vacuum /

Pressurizer Pressure Low Signal Actuation Failure / Low Pressure Safety Injection Pumps Start Failure Time Position Applicants Actions or Behavior Booth Operator: When directed, EXECUTE Events 6, 7, 8, and 9.

- Inadvertent Turbine Trip.

- Pressurizer Safety Valve fails 50% open on Reactor Trip.

- Loss of Condenser Vacuum @ 100%.

- Pressurizer Pressure Low Signal (PPLS) Actuation failure.

- Low Pressure Safety Injection Pumps start failure.

Indications Available:

Numerous Reactor Trip and Turbine Trip Alarms.

+10 sec ATCO RECOGNIZE Reactor Trip due to Turbine Trip.

CRS DIRECT performance of EOP-00, Standard Post Trip Actions.

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

ATCO VERIFY Reactivity Control: [Step 1]

VERIFY ALL of the following: [Step 1.a]

VERIFY no more than one Regulating or Shutdown CEA NOT inserted.

VERIFY Reactor Power is LOWERING.

VERIFY Startup Rate is NEGATIVE.

MONITOR plant for an uncontrolled RCS Cooldown. [Step 1.b]

CRS DETERMINE Reactivity Control criteria SATISFIED.

BOPO VERIFY Turbine Trip: [Step 2]

VERIFY HP & LP Stop and Intercept Valves CLOSED.

BOPO ENSURE all Generator Breakers are tripped: [Step 3]

DETERMINE Generator Output Breaker 3451-4 tripped.

DETERMINE Generator Output Breaker 3451-5 tripped.

DETERMINE Generator Field Breaker 41E/G1F tripped.

BOPO VERIFY 4160 V Safeguards Buses 1A3 and 1A4 are ENERGIZED. [Step 4]

NRC Simulator Scenario 1 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 6, 7, 8, & 9 Page 17 of 29 Event

Description:

Inadvertent Main Turbine Trip / Pressurizer Safety Valve Failure / Loss of Condenser Vacuum /

Pressurizer Pressure Low Signal Actuation Failure / Low Pressure Safety Injection Pumps Start Failure Time Position Applicants Actions or Behavior CRS DETERMINE Maintenance of Vital Auxiliaries criteria SATISFIED.

Examiner Note: Diesel Generators only start after safeguards (PPLS) actuation.

VERIFY both Diesel Generators RUNNING on Safety Injection Actuation BOPO Signal. [Step 5]

VERIFY 4160 V Non-Safeguards Buses 1A1 and 1A2 are ENERGIZED.

BOPO

[Step 6]

BOPO VERIFY 125 VDC Buses 1 and 2 are ENERGIZED. [Step 7]

BOPO VERIFY Instrument Air is AVAILABLE: [Step 8]

DETERMINE Instrument Air pressure 90 psig.

DETERMINE Instrument Air Compressor CA-1A RUNNING.

ATCO VERIFY Component Cooling Water System operation NORMAL: [Step 9]

DETERMINE at least one CCW pump RUNNING. [Step 9.a]

DETERMINE CCW Pump discharge pressure 60 psig. [Step 9.b]

DETERMINE HCV-438A/B/C/D, CCW to RCP Coolers OPEN. [Step 9.c]

DETERMINE at least one Raw Water Pump RUNNING. [Step 9.d]

CRS VERIFY RCS Inventory Control criteria satisfied: [Step 10]

DETERMINE PZR level NOT between 30% and 70% and NOT ATCO TRENDING to between 45% and 60%.

DETERMINE RCS subcooling 20°F:

[CA] RESTORE Inventory Control by manually controlling Charging and Letdown. [Step 10.1.a]

CRS DETERMINE RCS Inventory Control criteria NOT SATISFIED.

CRS VERIFY RCS Pressure Control criteria satisfied: [Step 11]

NRC Simulator Scenario 1 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 6, 7, 8, & 9 Page 18 of 29 Event

Description:

Inadvertent Main Turbine Trip / Pressurizer Safety Valve Failure / Loss of Condenser Vacuum /

Pressurizer Pressure Low Signal Actuation Failure / Low Pressure Safety Injection Pumps Start Failure Time Position Applicants Actions or Behavior

DETERMINE RCS pressure NOT between 1800 psia and 2300 psia and ATCO NOT trending to between 2050 psia and 2150 psia.

[CA] DETERMINE RCS pressure < 2300 psia and PORV NOT open.

[Step 11.1]

[CA] DETERMINE RCS pressure 1350 psia and TRIP one RCP in each loop. [Step 11.2]

[CA] DETERMINE RCS pressure 1600 psia and ENSURE PPLS actuated. [Step 11.3]

Manually Actuate Pressurizer Pressure Low Signal (PPLS) when RCS Pressure CRITICAL TASK 1600 psia and before 1350 psia to ensure SIAS / VIAS / CIAS Activation.

STATEMENT Pressure at Time of PPLS Trip ______ psia.

CRITICAL TASK ATCO DETERMINE PPLS relays NOT tripped and manually ACTUATE PPLS.

[CA] INSERT and TURN keys at 86A/PPLS Test Switch & 86B/PPLS ATCO Test Switch on AI-30A & AI-30B. [Step 11.3.a]

[CA] DETERMINE PPLS relays 86A/PPLS / 86B/PPLS / 86A1/PPLS /

86B1/PPLS have TRIPPED. [Step 11.3.b]

[CA] DETERMINE VIAS relays 86A/VIAS / 86B1/VIAS / 86B/VIAS /

86A1/VIAS have TRIPPED. [Step 11.3.c]

[CA] DETERMINE SIAS relays 86A/SIAS / 86AX/SIAS / 86B1/SIAS /

86B1X/SIAS / 86B/SIAS / 86BX/SIAS / 86A1/SIAS / 86A1X/SIAS have TRIPPED. [Step 11.3.d]

[CA] DETERMINE CIAS relays 86A/CIAS / 86B1/CIAS / 86B/CIAS /

86A1/CIAS have TRIPPED. [Step 11.3.e]

[CA] ENSURE HPSI / LPSI / Charging Pumps RUNNING. [Step 11.3.f]

DETERMINE HPSI Pumps SI-2A & SI-2B or SI-2B & SI-2C RUNNING.

DETERMINE LPSI Pumps NOT RUNNING and manually ATCO START SI-1A and SI-1B.

DETERMINE Charging Pumps CH-1A, CH-1B, & CH-1C RUNNING.

[CA] ENSURE adequate SI flow per IC-13, Safety Injection Flow vs.

Pressurizer Pressure. [Step 11.3.g]

[CA] DETERMINE Emergency Boration in progress. [Step 11.3.h]

NRC Simulator Scenario 1 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 6, 7, 8, & 9 Page 19 of 29 Event

Description:

Inadvertent Main Turbine Trip / Pressurizer Safety Valve Failure / Loss of Condenser Vacuum /

Pressurizer Pressure Low Signal Actuation Failure / Low Pressure Safety Injection Pumps Start Failure Time Position Applicants Actions or Behavior CRS DETERMINE RCS Pressure Control criteria NOT SATISFIED.

Examiner Note: The following steps are from RC-11, Emergency Boration Verification.

ATCO ENSURE the following valves are CLOSED: [Step 1]

FCV-269X, Demin Water Makeup Valve

FCV-269Y, Boric Acid Makeup Valve

HCV-264, CH-4A Recirc Valve

HCV-257, CH-4B Recirc Valve ATCO VERIFY all the following valves OPEN: [Step 2]

HCV-268, Boric Acid Pump Header to Charging Pumps Isolation Valve

HCV-265, CH-11A Gravity Feed Valve

HCV-258, CH-11B Gravity Feed Valve ATCO ENSURE all available Boric Acid Pumps RUNNING: [Step 3]

CH-4A, Boric Acid Pump

CH-4B, Boric Acid Pump ATCO ENSURE all available Charging Pumps RUNNING: [Step 4]

CH-1A, Charging Pump

CH-1B, Charging Pump

CH-1C, Charging Pump ATCO ENSURE the following valves are CLOSED: [Step 5]

LCV-218-2, VCT Outlet Valve

LCV-218-3, Charging Pump Suction SIRWT Isolation Valve Examiner Note: The following steps continue from EOP-00, Standard Post Trip Actions.

CRS VERIFY Core Heat Removal criteria satisfied: [Step 12]

NRC Simulator Scenario 1 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 6, 7, 8, & 9 Page 20 of 29 Event

Description:

Inadvertent Main Turbine Trip / Pressurizer Safety Valve Failure / Loss of Condenser Vacuum /

Pressurizer Pressure Low Signal Actuation Failure / Low Pressure Safety Injection Pumps Start Failure Time Position Applicants Actions or Behavior ATCO

DETERMINE RCP NPSH requirements met.

DETERMINE at least one RCP operating.

DETERMINE Core T 10°F.

Examiner Note: Depending on Crew actions, RCS subcooling will be lost in either EOP-00, SPTAs or EOP-03, LOCA.

Stop All Reactor Coolant Pumps (RCPs) when Subcooling is approaching or is CRITICAL TASK less than 20°F, before 0°F due to Loss of Net Positive Suction Head (NPSH) per STATEMENT RCP NPSH Curve.

CRITICAL TASK ATCO DETERMINE RCP subcooling < 20°F and PERFORM the following:

ATCO

STOP all RCPs.

[CA] PLACE TCV-909, Temperature Controller in MANUAL on DCS.

BOPO

[Step 12.2.a]

[CA] ENSURE TCV-909, Temperature Controller OUTPUT is zero BOPO (0). [Step 12.2.b]

CRS * [CA] VERIFY Natural Circulation in at least one Loop. [Step 12.2.c]

[CA] DETERMINE Core T 50°F.

[CA] DETERMINE difference between CETs and RCS THOT is 10°F on ERF "CHR" display.

[CA] DETERMINE RCS subcooling is 20°F.

[CA] DETERMINE THOT and TCOLD are stable or lowering.

CRS DETERMINE Core Heat Removal criteria NOT SATISFIED.

NOTE If Instrument Air to valves HCV-1105, HCV-1106, HCV-1107A/B and HCV-1108A/B is not available, throttling of these valves is not possible.

Open or close operation of these valves is possible for a minimum of three cycles.

CRS VERIFY RCS Heat Removal criteria satisfied:

NRC Simulator Scenario 1 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 6, 7, 8, & 9 Page 21 of 29 Event

Description:

Inadvertent Main Turbine Trip / Pressurizer Safety Valve Failure / Loss of Condenser Vacuum /

Pressurizer Pressure Low Signal Actuation Failure / Low Pressure Safety Injection Pumps Start Failure Time Position Applicants Actions or Behavior VERIFY Main Feedwater is restoring SG levels to 35% to 80% NR and 73%

BOPO to 94% WR. [Step 13]

DETERMINE FCV-1101 & FCV-1102 Feed Regulating Valves CLOSED.

[Step 13.a]

DETERMINE HCV-1105 & HCV-1106 Feed Regulating Bypass Valves ramped to between 40% & 45% OPEN. [Step 13.b]

BOPO

PLACE 43/FW switch in OFF. [Step 13.c]

ENSURE no more than one Main Feedwater Pump RUNNING.

[Step 13.d]

ENSURE no more than one Condensate Pump RUNNING. [Step 13.e]

STOP running Heater Drain Pumps FW-5A, FW-5B, and/or FW-5C.

BOPO

[Step 13.f]

ENSURE both sets of SG Blowdown Isolation Valves CLOSED.

[Step 13.g]

VERIFY Steam Dump and Bypass Valves controlling both of the following:

BOPO

[Step 14]

VERIFY RCS TCOLD between 525°F and 535°F.

VERIFY Steam Generator pressure between 850 psia & 925 psia.

[CA] DETERMINE loss of Condenser vacuum and PLACE HCV-1040, BOPO Atmosphere Dump Valve in service.

SELECT HCV-1040 on DCS Secondary Screen.

CRS DETERMINE RCS Heat Removal criteria SATISFIED.

CRS VERIFY Containment Isolation criteria satisfied:

ATCO VERIFY Normal Containment conditions exist: [Step 15]

DETERMINE unexpected rise in Containment Sump level. [Step 15.a]

DETERMINE Containment Area Radiation Monitors in ALARM.

ATCO

[Step 15.b]

DETERMINE Containment Ventilation Radiation Monitors RM-050 and ATCO RM-051 in ALARM. [Step 15.c]

[CA] ENSURE VIAS has ACTUATED and 86A/VIAS, 86A1/VIAS, 86B/VIAS, & 86B1/VIAS relays TRIPPED.

NRC Simulator Scenario 1 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 6, 7, 8, & 9 Page 22 of 29 Event

Description:

Inadvertent Main Turbine Trip / Pressurizer Safety Valve Failure / Loss of Condenser Vacuum /

Pressurizer Pressure Low Signal Actuation Failure / Low Pressure Safety Injection Pumps Start Failure Time Position Applicants Actions or Behavior

[CA] ENSURE RM-050 & RM-051 Containment Radiation Monitor Sample Pump STOPPED.

[CA] ENSURE RM-065, Post Accident Control Room Iodine Monitor RUNNING.

DETERMINE SG Blowdown & Condenser Off Gas Radiation Monitors (RM-054A / RM-054B / RM-057) NOT in alarm. [Step 15.d]

DETERMINE SG Blowdown & Condenser Off Gas Radiation Monitors (RM-054A / RM-054B / RM-057) NOT trending to alarm. [Step 15.e]

ATCO

VERIFY Containment conditions: [Step 15.f]

DETERMINE Containment pressure < 3 psig.

DETERMINE Containment temperature > 120°F.

CRS DETERMINE Containment Integrity criteria NOT SATISFIED.

CRS DIAGNOSE event in progress: [Step 16]

DETERMINE Reactivity Control requirements met.

DETERMINE both DC buses energized.

DETERMINE at least one Vital 4160 V Bus energized.

DETERMINE at least one Non-Vital 4160 V Bus energized.

VERIFY at least one RCP running.

If not, CONSIDER EOP-02, Loss of Offsite Power/Forced Circulation.

DETERMINE adequate Feedwater flow to at least one SG.

VERIFY Pressurizer pressure > 1800 psia with high subcooled margin, normal SG pressure, and no indications of primary to secondary leakage.

If not, CONSIDER EOP-03, Loss of Coolant Accident.

NOTE Certain events (i.e., LOCA, SGTR, UHE and Loss of All Feedwater) do not require offsite power in order to adequately, mitigate the effects of the accident. For this reason, the LOCA, SGTR, UHE or Loss of All Feedwater procedure may be implemented even if a Loss of Offsite Power has also occurred.

DETERMINE all Safety Function Acceptance Criteria NOT SATISFIED.

NRC Simulator Scenario 1 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 6, 7, 8, & 9 Page 23 of 29 Event

Description:

Inadvertent Main Turbine Trip / Pressurizer Safety Valve Failure / Loss of Condenser Vacuum /

Pressurizer Pressure Low Signal Actuation Failure / Low Pressure Safety Injection Pumps Start Failure Time Position Applicants Actions or Behavior

DETERMINE single event in progress and transition to EOP-03, Loss of Coolant in Accident.

Examiner Note: The following steps are from EOP-03, Loss of Coolant Accident.

CRS CONFIRM Standard Post Trip Actions have been performed. [Step 1]

CRS CONFIRM Loss of Coolant Accident Diagnosis: [Step 2]

VERIFY Safety Function Status Check Acceptance Criteria being satisfied. [Step 2.a]

VERIFY CIAS is NOT present and SAMPLE both SGs. [Step 2.b]

CRS IMPLEMENT the Emergency Plan. [Step 3]

Time: __________

NOTE Floating Step BB, Minimizing DC Loads, requires operator action within 15 minutes of loss of either battery charger.

CREW MONITOR the Floating Steps. [Step 4]

DETERMINE RCS pressure 1600 psia and Containment pressure 5 psig CRS and CSAS NOT present. [Step 5]

DETERMINE RCS pressure 1600 psia and VERIFY Engineered ATCO Safeguards Actuation: [Step 6]

DETERMINE PPLS relays 86A/PPLS / 86B/PPLS / 86A1/PPLS /

86B1/PPLS have TRIPPED. [Step 6.a]

DETERMINE VIAS relays 86A/VIAS / 86B1/VIAS / 86B/VIAS /

86A1/VIAS relays TRIPPED. [Step 6.b]

DETERMINE SIAS relays 86A/SIAS / 86AX/SIAS / 86B1/SIAS /

86B1X/SIAS / 86B/SIAS / 86BX/SIAS / 86A1/SIAS / 86A1X/SIAS have TRIPPED. [Step 6.c]

DETERMINE CIAS relays 86A/CIAS / 86B1/CIAS / 86B/CIAS /

86A1/CIAS relays TRIPPED. [Step 6.d]

NRC Simulator Scenario 1 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 6, 7, 8, & 9 Page 24 of 29 Event

Description:

Inadvertent Main Turbine Trip / Pressurizer Safety Valve Failure / Loss of Condenser Vacuum /

Pressurizer Pressure Low Signal Actuation Failure / Low Pressure Safety Injection Pumps Start Failure Time Position Applicants Actions or Behavior CRS DETERMINE Containment pressure 5 psig. [Step 7]

ATCO DETERMINE SIAS actuated and OPTIMIZE SI flow. [Step 8]

ENSURE HPSI / LPSI / Charging Pumps RUNNING. [Step 8.a]

DETERMINE HPSI Pumps SI-2A & SI-2B RUNNING.

DETERMINE LPSI Pumps SI-1A and SI-1B RUNNING.

DETERMINE Charging Pumps CH-1A, CH-1B, & CH-1C RUNNING.

DETERMINE Emergency Boration in progress per RC-11, Emergency Boration Verification. [Step 8.b]

DETERMINE adequate SI flow per IC-13, Safety Injection Flow vs.

Pressurizer Pressure. [Step 8.c]

CRS VERIFY RCP operating parameters: [Step 9]

ATCO

ENSURE at least one RCP stopped if TCOLD < 500°F. [Step 9.a]

ENSURE one RCP stopped in each loop if RCS pressure 1350 psia.

ATCO

[Step 9.b]

ENSURE all RCPs STOPPED if RCS pressure < NPSH requirements ATCO per PC-12, RCS Pressure-Temperature Limits. [Step 9.c]

CRS RECORD time of SIAS initiation. [Step 10]

Time: __________

VERIFY normal Component Cooling Water (CCW) and Raw Water (RW)

ATCO System operation: [Step 11]

ENSURE at least 2 CCW Pumps RUNNING. [Step 11.a]

VERIFY CCW Pump discharge pressure 60 psig. [Step 11.b]

ENSURE at least 2 RW Pumps RUNNING. [Step 11.c]

ENSURE at least 3 CCW Heat Exchangers in service. [Step 11.d]

ENSURE all RCP Coolers CCW Valves OPEN. [Step 11.e]

NRC Simulator Scenario 1 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 6, 7, 8, & 9 Page 25 of 29 Event

Description:

Inadvertent Main Turbine Trip / Pressurizer Safety Valve Failure / Loss of Condenser Vacuum /

Pressurizer Pressure Low Signal Actuation Failure / Low Pressure Safety Injection Pumps Start Failure Time Position Applicants Actions or Behavior NOTE Do NOT isolate a PORV if the pressurizer is water solid.

ATCO VERIFY PORVs and PZR Code Safety Valves are CLOSED. [Step 12]

DETERMINE Quench Tank temperature, pressure, and level in ALARM.

[Step 12.a]

DETERMINE PZR Safety Valve discharge temperature high in ALARM.

[Step 12.b]

ATCO

NOTIFY CRS that a PZR Safety Valve is OPEN.

DETERMINE PORV Acoustic Flow Alarms are CLEAR. [Step 12.c]

NOTE Rising Radiation Monitor RM-053 count rate, rising CCW surge tank level or rising CCW surge tank pressure may be indications of a RCS-to-CCW leak.

ATCO DETERMINE RCS to CCW leak is NOT in progress. [Step 13]

CRS DETERMINE LOCA is inside Containment. [Step 14]

ATCO PERFORM the following for a LOCA inside Containment: [Step 15]

PLACE HC-504A, CNTMT SUMP PUMP WD-3A CONTROL SWITCH, in PULL-TO-LOCK. [Step 15.a]

PLACE HC-504B, CNTMT SUMP PUMP WD-3B CONTROL SWITCH, in PULL-TO-LOCK. [Step 15.a]

CLOSE HCV-506A, Containment Sump Isolation Valve. [Step 15.b]

CLOSE HCV-506B, Containment Sump Isolation Valve. [Step 15.b]

ATCO VERIFY all the following conditions exist: [Step 16]

DETERMINE all HPSI Pumps are operating.

DETERMINE SI flowrate is acceptable per IC-13 SI Flow vs. PZR Pressure.

DETERMINE Representative CET temperature less than superheat.

DETERMINE Reactor Vessel Level Monitoring System > 43% and NOT lowering.

NRC Simulator Scenario 1 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 6, 7, 8, & 9 Page 26 of 29 Event

Description:

Inadvertent Main Turbine Trip / Pressurizer Safety Valve Failure / Loss of Condenser Vacuum /

Pressurizer Pressure Low Signal Actuation Failure / Low Pressure Safety Injection Pumps Start Failure Time Position Applicants Actions or Behavior ATCO ENSURE SI-2C, HPSI Pump Control Switch in PULL-TO-LOCK.

ATCO DETERMINE NONE of the following conditions exist: [Step 17]

SI flowrate is less than IC-13 SI Flow vs. PZR Pressure.

Representative CET temperature greater than superheat.

Reactor Vessel Level Monitoring System < 43% and lowering.

CRS DETERMINE RCS leak is NOT isolated. [Step 18]

DETERMINE Steam Generator Isolation Signal (SGIS) NOT actuated.

BOPO

[Step 19]

DETERMINE SG levels between 35% and 85% NR using Main Feedwater.

BOPO

[Step 20]

MAINTAIN Feedwater flow per HR-15, Main Feed Pump Operation.

[Step 20.a]

CONTROL Feedwater flow per HR-11, Manual Feet Control (DCS).

[Step 20.b]

CAUTION Failure to place the Containment Spray Pumps to Pull to Lock may allow actuation of Spray into Containment. This can lead to Containment Sump Blockage.

ATCO SECURE all Containment Spray flow: [Step 21]

PLACE SI-3A, Containment Spray Pump in PULL-TO-LOCK.

PLACE SI-3B, Containment Spray Pump in PULL-TO-LOCK.

PLACE SI-3C, Containment Spray Pump in PULL-TO-LOCK.

CAUTION

1) When TC is 178°F or greater, the maximum RCS cooldown rate is 100°F/hr. When TC is less than 178°F, the maximum RCS cooldown rate is 50°F/hr.

2) No more than three RCPs shall be in operation when RCS temperature is less than 500°F.

COMMENCE a Steam Generator cooldown per HR-12, Secondary Heat CRS Removal Operation. [Step 22]

NRC Simulator Scenario 1 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 6, 7, 8, & 9 Page 27 of 29 Event

Description:

Inadvertent Main Turbine Trip / Pressurizer Safety Valve Failure / Loss of Condenser Vacuum /

Pressurizer Pressure Low Signal Actuation Failure / Low Pressure Safety Injection Pumps Start Failure Time Position Applicants Actions or Behavior

Time: __________

MAINTAIN RCS pressure per PC-12, Pressure-Temperature Limits.

CRS

[Step 23]

CONTROL RCS heat removal per HR-12, Secondary Heat Removal BOPO Operation. [Step 23.a]

ATCO

CONTROL RCS pressure per PC-11, Pressure Control. [Step 23.b]

If HPSI Stop and Throttle criteria are met, CONTROL Charging, ATCO Letdown, and HPSI flow per IC-11, Inventory Control. [Step 23.c]

NOTE Voiding of the RCS is indicated by the inability to depressurize to SDC entry pressure.

Attachment IC-14, RCS Void Elimination, provides guidance to correct this condition.

COMMENCE depressurizing RCS to 300 psia using any of the following CRS per PC-11, Pressure Control: [Step 24]

CONTROL Pressurizer Spray flow.

CONTROL Charging and Letdown flow.

THROTTLE HPSI Pumps.

Time: __________

Commence a Cooldown and Depressurization of the Reactor Coolant System CRITICAL TASK before Reactor Vessel Level Monitoring System (RFLMS) is less than 100%,

STATEMENT indicating a bubble has formed in the head, to Reestablish RCS Inventory Control while maintaining RCS Heat Removal.

CRITICAL IMPLEMENT HR-12, Secondary Heat Removal Operation, to lower RCS TASK BOPO temperature.

Examiner Note: The following steps are from HR-12, Secondary Heat Removal Operation.

BOPO ENSURE Turbine Control is in MANUAL. [Step 1]

BOPO * [CA] DETERMINE Turbine NOT online and GO TO Step 4.

NRC Simulator Scenario 1 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 6, 7, 8, & 9 Page 28 of 29 Event

Description:

Inadvertent Main Turbine Trip / Pressurizer Safety Valve Failure / Loss of Condenser Vacuum /

Pressurizer Pressure Low Signal Actuation Failure / Low Pressure Safety Injection Pumps Start Failure Time Position Applicants Actions or Behavior NOTES

1. In MANUAL, single arrows are 1% / double arrows are 5% change in OUTPUT value.

2. While Steam Dump and Bypass Control is in Temperature/Pressure Mode, the controllers PC0910 and TC0909_PI will alternate between controls, depending on the higher output signal. A red square outlining the controlling function signify which parameter is in control.

3. Steps 4 through 11 may be performed as needed, and in any order.

CAUTIONS When TC is 178°F or greater, the maximum RCS cooldown rate is 100°F/hr.

When TC is less than 178°F, the maximum RCS cooldown rate is 50°F/hr.

If Steam Dump and Bypass (SD&B) is available, CONTROL RCS BOPO temperature with a single SD&B Valve. [Step 4]

[CA] DETERMINE Steam Dump and Bypass is NOT available and GO BOPO TO Step 9.

Examiner Note: HCV-1040, Atmospheric Dump Valve, may already be in service following the Loss of Condenser Vacuum that occurred on Reactor Trip.

BOPO If HCV-1040, is available, CONTROL RCS temperature as follows: [Step 9]

DEPRESS the valve toggle to SELECT HCV-1040. [Step 9.a]

PUSH UP and DOWN arrows as required to ADJUST HCV-1040 output as needed. [Step 9.b]

CRITICAL TASK ATCO IMPLEMENT PC-11, Pressure Control, to lower RCS pressure.

Examiner Note: The following steps are from PC-11, Pressure Control. PC-12, RCS Pressure-Temperature Limits (graph), is maintained on a Control Room hardcopy.

CAUTION A charging header flow path must be maintained at all times.

MAINTAIN RCS pressure per the PC-12, RCS Pressure-Temperature Limits ATCO graph: [Step 1]

DETERMINE Steps 1.a through 1.d N/A. [Step 1.e]

NRC Simulator Scenario 1 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 6, 7, 8, & 9 Page 29 of 29 Event

Description:

Inadvertent Main Turbine Trip / Pressurizer Safety Valve Failure / Loss of Condenser Vacuum /

Pressurizer Pressure Low Signal Actuation Failure / Low Pressure Safety Injection Pumps Start Failure Time Position Applicants Actions or Behavior

CONTROL Auxiliary Spray flow as necessary by operating the following:

[Step 1.e]

HCV-240, PZR Auxiliary Spray Isolation Valve

HCV-249, PZR Auxiliary Spray Isolation Valve

HCV-238, Loop 1 Charging Isolation Valve

HCV-239, Loop 2 Charging Isolation Valve

If HPSI Stop and Throttle criteria is met, CONTROL Pressurizer level using Charging, Letdown, and/or HPSI flow per IC-11, Inventory Control.

[Step 1.f]

CONTROL RCS heat removal per HR-12, Secondary Heat Removal Operation. [Step 1.g]

When RCS Cooldown and Depressurization is in progress, TERMINATE the scenario.

NRC Simulator Scenario 1 Outline Rev. 6

Appendix D Scenario Outline Form ES-D-1 Facility: Fort Calhoun Station Scenario No.: 2 Op Test No.: Dec 2015 NRC Examiners: Operators:

Initial Conditions: 100% power MOL - RCS Boron is 482 ppm (by sample).

Turnover: Maintain steady-state power conditions. Perform Containment Spray Pump SI-3A Operability Test per OI-CS-1, Containment Spray Pump Normal Operation, Attachment 1A.

Critical Tasks:

Commence an Emergency Boration of the RCS Due to 2 or more Stuck CEAs when Diesel Generator DG-1 Breaker is Closed and Bus 1A3 is Reenergized to Restore Reactivity Control. (Event 7)

Restore Feedwater Flow to At Least One Steam Generator to Reestablish any SG as a Heat Sink. (Event 6)

Restore Power to any 4160 V Safeguards Bus using a Diesel Generator to Reestablish Maintenance of Vital Auxiliaries and Allow Branching to Meet other Safety Functions During a Station Blackout. (Event 8)

Event No. Malf. No. Event Type* Event Description 1 N (ATCO) Perform OI-CS-1, Containment Spray Normal Operation,

+15 min Attachment 1A, SI-3A Containment Spray Pump Operability Test.

2 Severe Thunderstorm Watch from the National Weather Service.

+20 min AOP-01, Acts of Nature,Section II, Severe Weather Entry Required.

3 I (BOPO, CRS) Steam Generator RC-2A Level Channel LT-903Y Fails High.

+30 min TS (CRS) Feedwater Control System Automatically Shifts to Manual.

4 I (ATCO, CRS) VCT Level Transmitter LT-219 Fails Low due to CVCS leak.

+45 min TS (CRS) 5 C (BOPO, CRS) Loss of 161 KV Line.

+55 min Place Condensate Pump FW-2A in service.

6 M (ATCO, BOPO, Loss of Offsite Power.

+55 min CRS) Reactor Trip.

7 C (ATCO) Four (4) Stuck CEAs on Reactor Trip.

+55 min Emergency Boration Required Upon Power Restoration.

8 M (ATCO, BOPO, Diesel Generator DG-01 Breaker Failure with Diesel Generator

+60 min CRS) DG-02 Overspeed Trip. Station Blackout.

9 C (BOPO) Diesel Driven Auxiliary Feedwater Pump FW-54 Start Failure.

+70 min EOP-20, Functional Recovery Entry Required.

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Technical Specifications Actual Target Quantitative Attributes 2 Malfunctions after EOP entry (1-2) 4 Abnormal events (2-4) 2 Major transients (1-2) 1 EOPs entered/requiring substantive actions (1-2) 1 EOP contingencies requiring substantive actions (0-2) 3 Critical tasks (2-3)

NRC Simulator Scenario 2 Outline Rev. 6

Scenario Event Description NRC Scenario 2 SCENARIO

SUMMARY

NRC 2 The crew will assume the shift at 100% power per OP-4, Load Change and Normal Power Operation.

The scheduled activity is to perform OI-CS-1, Containment Spray Normal Operation, Attachment 1A, SI-3A Containment Spray Pump Operability Test.

The next event is a Severe Thunderstorm Watch from the National Weather Service requiring entry into AOP-01, Acts of Nature,Section II, Severe Weather. Once plant announcements have been made, a high failure of Steam Generator RC-2A Level Channel LT-903Y will occur. Initial operator actions are per ARP-DCS-FW, Feedwater DCS Annunciator Response Procedure, and include verifying Feedwater Control is in MANUAL and bypassing the failed input. Once completed, Feedwater Control is restored to AUTO per OI-FW-3, Steam Generator Level Control, Attachment 4, Level Controller Operation. The SRO will refer to Technical Specification LCO 2.15.3 - Steam Generator Narrow Range Level Instrument at AI-179.

The next event is a sensing line leak resulting in a low failure of Volume Control Tank (VCT) Level Transmitter LT-219. Actions are per ARP-CB-1/2/3/A2, VOLUME CONTROL TANK LEVEL HI-LO, until it is determined that a leak exists. Once identified, AOP-33, CVCS Leak, is entered. The SRO will refer to Technical Specification LCO 2.15.3 - Volume Control Tank Level Instrument at AI-185.

A lightning strike in the Fort Calhoun Switchyard will open 161 KV Breakers 110 and 111 and result in a loss of the 161 KV lines. A successful Fast Bus Transfer initially maintains power to all 4160 V Buses.

The crew enters AOP-31, 161 KV Grid Malfunctions,Section II, All 4160 V Buses Fed from 22 KV.

AOP-31 requires placing FW-2A, Condensate Pump in service to balance electric system loads per OI-FW-1, Condensate System Normal Operation. When AOP-31, Step 6, Matching Breaker Flags is performed, a Plant Trip will occur.

A Loss of Offsite Power occurs on the Plant Trip and initiates a failure of both Emergency Diesel Generators. When the Reactor Trip is verified, four (4) CEAs will be identified as stuck and an Emergency Boration is required but cannot be initiated due to loss of power. EOP-00, Standard Post Trip Actions, is entered and feedwater flow must be aligned to the Steam Generators using AFW-10, Steam Driven Auxiliary Feedwater (AFW) Pump via the AFW Nozzles or the Feed Ring. When Diagnostic Actions are performed in EOP-00 the crew should recognize a loss of both Reactivity Control (4 Stuck CEAs with no Emergency Boration flow) and Maintenance of Vital Auxiliaries (no energized 4160 V Safeguards Bus) and enter EOP-20, Functional Recovery. EOP-20, Resource Assessment Trees RC-1, CEA Insertion and MVA-AC, Restoration of AC are the significant Safety Functions to be addressed.

The event is complicated by a start failure of FW-54, Diesel Driven AFW Pump (normal post-trip AFW source), and requires starting and aligning of FW-10, Turbine Driven AFW Pump. The scenario is terminated in EOP-20 when power is restored to Safeguards Bus 1A3 via a replaced DG-01 Output Breaker and Emergency Boration flow is initiated to the Reactor Coolant System.

Risk Significance:

Failure of risk important system prior to trip: Loss of 161 KV Line

Risk significant core damage sequence: Loss of Reactivity Control Station Blackout/Loss of Feedwater

Risk significant operator actions: Establish Feedwater Flow Emergency Borate for 4 Stuck CEAs Restore Power to Safeguards Bus NRC Simulator Scenario 2 Outline Rev. 6

Scenario Event Description NRC Scenario 2 BOOTH OPERATOR INSTRUCTIONS for SIMULATOR SETUP RESET to IC-#2 (or any 100% MOL IC) and LOAD & EXECUTE NRC 2.sce for NRC Scenario 2.

Provide Lead Examiner key for HCV-2958 for performance of Normal evolution, Event 1.

Preset item - Event 7 - 4 Rods Stuck out on Reactor Trip Type Item Value Condition Malfunction ROD_PWR_229_2 Stuck Scenario Event: 4 stuck ROD_PWR_B15_2 Stuck rods ROD_PWR_B16_2 Stuck ROD_PWR_228_2 Stuck Preset Item - Event 8 - Diesel Generator #1 Breaker Failure Type Item Value Condition Malfunction BUS_1A3_20_BKR_Trip (1AD-1 True Scenario Event: DG1 Breaker failure to Trip position) Breaker Failure Preset Item - Event 9 - FW-54 Fails to Start Remotely Type Item Value Condition Remote REM:AFW_FWC04 Local Scenario Event: FW-54 REM:AFW_FWC02 Stop Start Failure Event 2 - Notification of Severe Thunderstorm Watch from National Weather Service Type Direction Booth Call on the NAWAS phone by dialing 98*, wait 5-10 seconds and REPORT:

Operator This is the National Advance Warning Alert System with an update. The National weather service in Valley, Nebraska has issued a Severe Thunderstorm Watch for Washington county Nebraska until (60 minute from current time). Current radar indicates conditions are met to produce severe thunderstorms with potentially heavy rain, high winds, and damaging hail. Individuals in the path of the storm are recommended to be attentive to weather conditions and consider moving to shelter in a sturdy structure.

Event 3 - Steam Generator Level Transmitter LT-903X Fails High Type Item Value Condition Transmitter LT-903Y 100, ramp = 30 sec When directed by examiner, LT-903Y-1 100, ramp = 30 sec trigger/activate this event.

Scenario Event: LT903X fail high Event 4 - VCT Level Transmitter Fails Low, VCT Leak Type Item Value Condition Transmitter CVC_LT219 0, ramp = 30 sec When directed by examiner, Malfunction CVX07B (VCT Leak) 2% trigger/activate this event.

Scenario Event: VCT LT-219 Fail Low When directed as Aux Building operator to isolate leak, delete malfunction CVX07B.

NRC Simulator Scenario 2 Outline Rev. 6

Scenario Event Description NRC Scenario 2 Event 5 - Loss of 161 KV Line Type Item Value Condition Malfunction 87L/161 (Relay 87L/161 trip) True When directed by examiner, trigger/activate this event.

Scenario Event: Loss of 161KV line Event 6 - Loss of Offsite Power Type Item Value Condition Malfunction SWD01 True When directed by examiner, trigger/activate this event.

Scenario Event: Loss of Offsite Power Event 8 - Diesel Generator #2 Overspeed Trip Type Item Value Condition Expert H_PD2_301TL_1 1 Event is triggered REM:FDP_RCW1_1 2 automatically 10 minutes REM:FDP_RCW1_2 2 after reactor trip.

REM:FDP_RCW1_5 2 Scenario Event: DG2 H_PD2_311_1 1200, ramp = 10 sec overspeed trip H_PD2_311_1 0, Delay=11, ramp = 3 DGAQRL112x2TVSP 0 DGAQRL112X1TVSP 0 NRC Simulator Scenario 2 Outline Rev. 6

Scenario Event Description NRC Scenario 2 Booth Operator: INITIALIZE to IC-1 and LOAD NRC 2.sce.

ENSURE all Simulator Annunciator Alarms are ACTIVE.

ENSURE all Control Board Tags are removed.

ENSURE Condensate Pumps FW-2B and FW-2C in service.

ENSURE Synchroscope Switch in a location other than DG-01 Breaker 1AD1.

ENSURE Reactivity Briefing Sheet printout provided with Turnover.

ENSURE Middle-of-Life Thumb Rule Sheet provided with Turnover.

ENSURE procedures in progress provided to crew in Briefing Room:

- COPY of OI-CS-1, Containment Spray Normal Operation, Attachment 1A, SI-3A Containment Spray Pump Operability Test, INITIALED through Prerequisites. Provide Key for HCV-2958.

Control Room Annunciators in Alarm:

NONE 0B Procedure List Event 1: OI-CS-1, Containment Spray Normal Operation, Attachment 1A, SI-3A Containment Spray Pump Operability Test Event 2: AOP-01, Acts of Nature,Section II, Severe Weather Event 3: ARP-DCS-FW, Feedwater DCS Annunciator Response Procedure Event 3: OI-FW-3, Steam Generator Level Control, Attachment 4, Level Controller Operation Event 4: ARP-CB-1/2/3/A2, Window B-2U, VOLUME CONTROL TANK LEVEL HI-LO Event 4: AOP-33, CVCS Leak Event 5: AOP-31, 161 KV Grid Malfunctions,Section II, All 4160 V Buses Fed from 22 KV Event 5: OI-FW-1, Condensate System Normal Operation, Attachment 4, Rotating Condensate Pumps Event 6: EOP-00, Standard Post Trip Actions Event 9: EOP-20, Functional Recovery Event 9: EOP-20, Resource Assessment Trees RC-1, CEA Insertion Event 9: EOP-20, Resource Assessment Trees MVA-AC, Restoration of AC NRC Simulator Scenario 2 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 2 Event # 1 Page 6 of 33 Event

Description:

Containment Spray Pump Operability Test Time Position Applicants Actions or Behavior Booth Operator: When directed, RESPOND to requests from Control Room.

Indications Available:

NONE Examiner Note: The following steps are from OI-CS-1, Containment Spray Normal Operation, Attachment 1A, SI-3A Containment Spray Pump Operability Test.

Booth Operator: If requested, report Containment Spray Pump SI-3A is ready to start and all conditions locally are normal.

ATCO PLACE the following switches to TEST. [Step 1]

CNTMT Spray Valve HCV-344 Test Switch HC-344/Test.

CNTMT Spray Valve HCV-345 Test Switch HC-345/Test.

ATCO VERIFY the following annunciators are in ALARM: [Step 2]

HCV-344/345 SET SPRAY PUMPS TEST PERMIT at AI-30A, A33-1, Window H-5.

HCV-344/345 SET SPRAY PUMPS TEST PERMIT at AI-30B, A34-1, Window H-3.

ATCO RUN SI-3A by completing the following: [Step 3]

REVIEW Technical Specification LCO 2.4 requirements and LOG into CRS the appropriate T.S. LCOs 2.4(2)c for HCV-345). [Step 3.a]

ATCO

OVERRIDE and CLOSE HCV-345: [Step 3.b]

PLACE HC-345, Containment Spray Valve Control Switch, to OVERRIDE

VERIFY annunciator SPRAY VALVE HCV-345 HEADER ISOLATED at AI-30B, A34-1, Window H-2 is in ALARM. [Step 3.c]

Examiner Note: Provide Key for HCV-2958 controls to candidate.

CLOSE HCV-2958, Containment Spray Pump SI-3A Discharge at AI123A. [Step 3.d]

VERIFY annunciator SI PUMPS VALVES OFF NORMAL at AI-30A, A33-1, ATCO Window J-1 is in ALARM. [Step 4]

NRC Simulator Scenario 2 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 2 Event # 1 Page 7 of 33 Event

Description:

Containment Spray Pump Operability Test Time Position Applicants Actions or Behavior Booth Operator: When contacted, REPORT SI-138 is OPEN and BACKSEATED.

Direct operator to ENSURE SI-138, Containment Spray Pump SI-3A ATCO Minimum Recirc Isolation Valve is OPEN and BACKSEATED. [Step 5]

ATCO DETERMINE the following valves are OPEN: [Step 6]

HCV-385, SIRWT Tank Recirculation Valve

HCV-386, SIRWT Tank Recirculation Valve ATCO START SI-3A, CNTMT Spray Pump. [Step 7]

Examiner Input: When timing is started, REPORT time compress and that 5 minutes has elapsed.

ATCO When five minutes has elapsed, STOP SI-3A, CNTMT Spray Pump. [Step 8]

OPEN HCV-2958, Containment Spray Pump SI-3A Discharge ay AI-128A.

ATCO

[Step 9]

VERIFY annunciator SI-3A, SI PUMPS VALVES OFF NORMAL at AI-30A, ATCO A33-1, Window J-1) is CLEAR. [Step 10]

PLACE HC-345, Containment Spray Valve HCV-345 Control Switch, to ATCO AUTO. [Step 11]

VERIFY annunciator SPRAY VALVE HCV-345 HEADER ISOLATED at ATCO AI-30B, A34-1, Window H-2 is CLEAR. [Step 12]

ATCO PLACE the following switches to OFF: [Step 13]

CNTMT Spray Vlv HCV-344 Test Switch HC-344/Test

CNTMT Spray Vlv HCV-345 Test Switch HC-345/Test NRC Simulator Scenario 2 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 2 Event # 1 Page 8 of 33 Event

Description:

Containment Spray Pump Operability Test Time Position Applicants Actions or Behavior ATCO VERIFY the following annunciators are CLEAR: [Step 14]

HCV-344/345 SET SPRAY PUMPS TEST PERMIT at AI-30A, A33-1, Window H-5

HCV-344/345 SET SPRAY PUMPS TEST PERMIT at AI-30B, A34-1, Window H-3 CRS LOG out of Technical Specification 2.4 LCOs. [Step 15]

Booth Operator: When contacted, REPORT signs are removed.

CONTACT Auxiliary Operator to remove Protective Equipment Signs.

ATCO

[Step 16]

When CRS has logged out of Technical Specifications, PROCEED to Event 2.

NRC Simulator Scenario 2 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 2 Event # 2 Page 9 of 33 Event

Description:

Severe Thunderstorm Watch from the National Weather Service Time Position Applicants Actions or Behavior Booth Operator: When directed, REPORT Event 2.

- Severe Thunderstorm Watch from the National Weather Service Indications Available:

NONE Booth Operator: CONTACT Control Room on NAWAS phone READ prepared message.

Call on the NAWAS phone by dialing 98*, wait 5-10 seconds and REPORT:

This is the National Advance Warning Alert System with an update. The National weather service in Valley, Nebraska has issued a Severe Thunderstorm Watch for Washington county Nebraska until (60 minute from current time). Current radar indicates conditions are met to produce severe thunderstorms with potentially heavy rain, high winds, and damaging hail. Individuals in the path of the storm are recommended to be attentive to weather conditions and consider moving to shelter in a sturdy structure.

CRS REFER to AOP-01, Acts of Nature,Section II, Severe Weather.

Examiner Note: The following steps are from AOP-01, Acts of Nature,Section II, Severe Weather.

NOTE The Shift Manager and Station Duty Manager should discuss the potential for wind-generated missiles and the necessity to restore any Engineered Safeguards Equipment that may be out of service.

NOTIFY Manager-Shift Operations and Station Duty Manager of weather CRS conditions. [Step 4.1]

If weather conditions allow, PERFORM a visual inspection of the Protected CRS Area and Switchyard per SO-G-119, Site Wind Generated Missile Protection Standards. [Step 4.2]

NOTES

1. Guidance for announcements for the Administration Building and Training Center are located in EPIP-OSC-15, Communicator Actions.

2. Steps 3 through 6 can be performed in order and as needed as required by weather conditions.

CRS If a severe thunderstorm watch exists, PERFORM the following: [Step 4.3]

NRC Simulator Scenario 2 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 2 Event # 2 Page 10 of 33 Event

Description:

Severe Thunderstorm Watch from the National Weather Service Time Position Applicants Actions or Behavior

MONITOR NAWAS to determine changes in weather conditions.

[Step 4.3.a]

ANNOUNCE and REPEAT the following over the plant communications system: [Step 4.3.b]

"Attention all personnel. Attention all personnel. A severe thunderstorm watch has been issued for area surrounding the plant until 10 PM tonight."

When Plant announcement has been made, PROCEED to Event 3.

NRC Simulator Scenario 2 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 2 Event # 3 Page 11 of 33 Event

Description:

Steam Generator Level Channel Failure Time Position Applicants Actions or Behavior Booth Operator: When directed, EXECUTE Event 3.

- Steam Generator RC-2A Level Channel LT-903Y fails high.

Indications Available:

Feedwater Digital Control System Alarm

+30 sec BOPO RESPOND to Annunciator Response Procedures.

INFORM CRS Steam Generator RC-2A Level Transmitter LT-903Y failed BOPO high.

CRS DIRECT actions of ARP-DCS-FW, LT-903Y.

Examiner Note: The following steps are from ARP-DCS-FW, Feedwater Digital Control System.

DETERMINE failure is NOT from a Feedwater Flow, Steam Flow, or Steam BOPO Pressure Instrument. [Step 1]

BOPO PERFORM the following for Level Instrument LT-903Y failure: [Step 2]

VERIFY FWCS IN MANUAL is displayed on SECONDARY Feedwater Regulating System display. [Step 2.1]

TOUCH display with the BAD process. A 'B' will be displayed beside the level indication. [Step 2.2]

DETERMINE BAD input NOT automatically bypassed. [Step 2.3]

TOUCH Bypass on verification faceplate to BYPASS BAD input.

[Step 2.3.1]

VERIFY point displays GOOD status and 'B' is no longer displayed.

[Step 2.3.2]

When Steam Generator level is in band, RETURN Level Controller, LC0903_1E back to AUTO per OI-FW-3. [Step 2.4]

Examiner Note: The following steps are from OI-FW-3, Steam Generator Level Control, Attachment 4, Level Controller Operation, Step 5.

PERFORM the following to return Level Controller LC0903_E1 (DCS) to BOPO AUTOMATIC control: [Step 5]

NRC Simulator Scenario 2 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 2 Event # 3 Page 12 of 33 Event

Description:

Steam Generator Level Channel Failure Time Position Applicants Actions or Behavior

SELECT Level Controller LC0903_1E (DCS). [Step 5.a]

PERFORM one of the following to DISPLAY controller: [Step 5.a]

TOUCH FWTR CTRL-2A pushbutton from Feedwater Regulating System display, OR

TOUCH Feedwater Level Control Button from the LVLS display.

PERFORM the following for Steam Generator RC-2A: [Step 5.b]

TOUCH AUTO on LC0903_1E (DCS) Level Controller and VERIFY the 'T' is displayed. [Step 5.b.1)]

DETERMINE FC1101, S/G RC-2A FW REG VLV (DCS) is in AUTO.

[Step 5.b.2)]

DETERMINE HC1105 is in AUTO. [Step 5.b.3)]

VERIFY Feed Regulating System return to 3 ELEMENT AUTO.

[Step 5.b.4)]

Examiner Note: The following steps continue from ARP-DCS-FW.

CRS DETERMINE other Steam Generator level instruments NOT affected. [Step 2]

CRS DETERMINE BAD input bypassed MANUALLY. [Step 3]

BOPO MONITOR Steam Generator levels. [Step 4]

VERIFY XC-105, Secondary Calorimetric, is valid. [Step 5]

CRS Examiner Note: This step is normally performed by the STA - The CRS may not address XC-105 at this time.

Booth Operator: When contacted, REPORT Level Transmitter LT-903Y-1 is failed high at AI-179.

DETERMINE LT-903 is cause of alarm and CONTACT Auxiliary Operator to CRS VERIFY level indication at AI-179. [Step 6]

CRS NOTIFY Work Week Manager of LT-903Y malfunction. [Step 7]

NRC Simulator Scenario 2 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 2 Event # 3 Page 13 of 33 Event

Description:

Steam Generator Level Channel Failure Time Position Applicants Actions or Behavior EVALUATE Technical Specification LCO 2.15.3, Alternate Shutdown and CRS Auxiliary Feedwater Panel

LCO 2.15.3.(1) - Steam Generator Narrow Range Level Instrument at AI-179 (Table 2.6 / Item #3.d)

CONDITION 2.15.3.(1) - One Steam Generator Narrow Range Level Transmitter inoperable

ACTION 2.15.3.(1) - RESTORE the required channel to OPERABLE status within seven (7) days.

When Technical Specifications have been addressed, PROCEED to Event 4.

NRC Simulator Scenario 2 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 2 Event # 4 Page 14 of 33 Event

Description:

VCT Level Transmitter Leak Time Position Applicants Actions or Behavior Booth Operator: When directed, EXECUTE Event 4.

- VCT Level Transmitter leak on LT-219 line.

Indications Available:

CB-1,2,3/A2 - VOLUME CONTROL TANK LEVEL HI-LO VCT level indication LIC-219 slowly lowering

+30 sec ATCO RESPOND to Annunciator Response Procedures.

ATCO INFORM CRS of VCT Level Transmitter LT-219 slowly lowering.

REFER to ARP-CB-1,2,3/A2, Window B-2U - VOLUME CONTROL TANK CRS LEVEL HI-LO.

Examiner Note: The following steps are from ARP-CB-1.2.3/A2, Window B-2U - VOLUME CONTROL TANK LEVEL HI-LO.

DETERMINE VCT Level Indication on LIC-219 NOT between 51.7% and ATCO 91.2%. [Step 1]

Booth Operator: When contacted, WAIT 2 minutes then REPORT indications of leakage from the VCT Level Transmitter.

ATCO If level is low, PERFORM the following: [Step 2]

ALIGN LCV-218-1, VCT Inlet Valve is aligned to VCT. [Step 2.1]

DETERMINE Pressurizer level is at program. [Step 2.2]

DETERMINE makeup to VCT NOT required. [Step 2.3]

CONTACT Auxiliary Operator to check CVCS System for leakage.

[Step 2.4]

ATCO DETERMINE VCT level is NOT high. [Step 3]

Booth Operator: When contacted, WAIT one minute and REPORT VCT level indication at AI-185 indicates 0%.

NRC Simulator Scenario 2 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 2 Event # 4 Page 15 of 33 Event

Description:

VCT Level Transmitter Leak Time Position Applicants Actions or Behavior ATCO DETERMINE VCT level indication is due to a leak. [Step 4]

CONTACT Auxiliary Operator to verify level indication at AI-185, Alternate Shutdown Panel. [Step 4.1]

If low level is due to a system leak, IMPLEMENT AOP-33, CVCS Leak. [Step 4.1.1]

EVALUATE Technical Specification LCO 2.15.3, Alternate Shutdown and CRS Auxiliary Feedwater Panel

LCO 2.15.3.(1) - Volume Control Tank Level Instrument at AI-185 (Table 2.6 / Item #4.b)

CONDITION 2.15.3.(1) - One Volume Control Tank Level Instrument inoperable

ACTION 2.15.3.(1) - RESTORE the required channel to OPERABLE status within seven (7) days.

Examiner Note: The following steps are from AOP-33, CVCS Leak.

CRS PERFORM the following to isolate CVCS: [Step 4.1]

ATCO

CLOSE both Letdown Isolation Valves. [Step 4.1.a]

CLOSE TCV-202.

CLOSE HCV 204.

Examiner Note: With all 3 Charging Pumps in PULL-TO-LOCK, Technical Specification LCO 2.2.4 would be temporarily entered until a Charging Pump is restarted later in the AOP. This is an identified Procedural Enhancement Opportunity.

ATCO

PLACE Charging Pump Control Switches in PULL-TO-LOCK. [Step 4.1.b]

PLACE CH-1A in PULL-TO-LOCK

PLACE CH-1B in PULL-TO-LOCK

PLACE CH-1C in PULL-TO-LOCK ATCO

ENSURE the following valves are CLOSED: [Step 4.1.c]

CLOSE HCV-238, Loop 1 Charging Isolation.

CLOSE HCV-239, Loop 2 Charging Isolation.

VERIFY HCV-240, PZR Auxiliary Spray Isolation Valve CLOSED.

NRC Simulator Scenario 2 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 2 Event # 4 Page 16 of 33 Event

Description:

VCT Level Transmitter Leak Time Position Applicants Actions or Behavior

VERIFY HCV-249, PZR Auxiliary Spray Isolation Valve CLOSED.

CRS IMPLEMENT the Emergency Plan. [Step 4.2]

Booth Operator: REPORT as Auxiliary Operator that indications of leakage are from the VCT level transmitter line. REPORT as Chemistry that no sampling is in progress.

Booth Operator: If directed to isolate the level transmitter, EXECUTE remote function to LOCALLY CLOSE CH-227 and CH-206 which isolates LT-218 and LT-219.

CRS PERFORM the following to locate the leak: [Step 4.3]

AO

VISUALLY inspect CVCS system piping. [Step 4.3.a]

CHECK all the following levels: [Step 4.3.b]

DETERMINE Spent Regen Tank level normal.

DETERMINE Aux Building Sump Tank RISING.

DETERMINE Containment Sump level normal.

CRS

DIRECT Chemistry to isolate all CVCS sample lineups. [Step 4.3.c]

NOTE If leak is contained by the actions in Step 1, PZR level will lower at a rate of approximately 1% every 12 minutes due to Reactor Coolant Pump Bleedoff flow.

CRS DETERMINE Pressurizer level NOT lowering abnormally. [Step 4.4]

NOTE VCT level will tend to rise approximately 1% every 6 minutes due to Reactor Coolant Pump Bleedoff flow.

DETERMINE VCT level is lowering and CLOSE LCV-218-2, VCT Outlet ATCO Valve. [Step 4.5]

ATCO DETERMINE VCT level continues to lower and ISOLATE the VCT: [Step 4.6]

PLACE LCV-218-1, VCT Inlet Valve, in RWTS. [Step 4.6.a]

NRC Simulator Scenario 2 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 2 Event # 4 Page 17 of 33 Event

Description:

VCT Level Transmitter Leak Time Position Applicants Actions or Behavior

ENSURE HCV-208, RCP Bleedoff to RCDT Isolation Valve, is open.

[Step 4.6.b]

CLOSE all of the following valves: [Step 4.6.c]

HCV-241, RCP Bleedoff to VCT Isolation Valve

HCV-206, RCP Bleedoff to VCT Isolation Valve Booth Operator: REPORT as Auxiliary Operator that SL-130 and SL-135 are CLOSED.

SL-130, SAMPLE RETURN TO VOLUME CONTROL TANK CH-14 ISOLATION VALVE in Room 60.

SL-135, VCT CH-14 RCS SAMPLE RETURN ISOLATION VALVE in Room 60.

Booth Operator: CONTACT as Shift Manager and PERFORM RCS Makeup from the SIRWT.

CRS PERFORM Step a or b to MAINTAIN PZR level 45-60%: [Step 4.7]

COMMENCE RCS makeup at existing boron concentration per Attachment A, Blended Makeup to the Charging Pump Suction Header.

[Step 4.7.a]

PERFORM the following and COMMENCE RCS makeup from SIRWT:

[Step 4.7.b]

OPEN LCV-218-3, Charging Pump Suction SIRWT Isolation Valve.

[Step 4.7.b.1)]

OPEN both Charging Isolation Valves: [Step 4.7.b.2)]

OPEN HCV-238.

OPEN HCV-239.

START at least one Charging Pump. [Step 4.7.b.3)]

CRS EVALUATE need to implement AOP-09, High Radioactivity. [Step 4.8]

When Boron addition via SIRWT is commenced, PROCEED to Event 5.

NRC Simulator Scenario 2 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 2 Event # 5 Page 18 of 33 Event

Description:

Loss of 161 KV Line Time Position Applicants Actions or Behavior Booth Operator: When directed, EXECUTE Event 5.

- Loss of 161 KV Line.

Indications Available:

CB-20/A15 - BREAKER 111 TRIPPED CB-20/A15 - 161 KV SUPPLY BKR LOCKOUT RELAY OPERATED CB-20/A15 - BREAKER 110 TRIPPED CB-20/A15 - PLANT 161 KV LINE LOW FREQUENCY CB-20/A17 - TRANS T1A-3 SECONDARY LOW VOLTAGE CB-20/A17 - BKR 1A33 AUTO TRIP CB-20/A17 - TRANS TIA-3 LOCKOUT RELAY OPERATED 86/TIA-3 CB-20/A18 - TRANS T1A-4 SECONDARY LOW VOLTAGE CB-20/A18 - BKR 1A44 AUTO TRIP CB-20/A18 - TRANS TIA-4 LOCKOUT RELAY OPERATED 86/TIA-4 Supply Breakers 110 AND 111 white trip lights lit

+30 sec BOPO RESPOND to Annunciator Response Procedures.

BOPO INFORM CRS of loss of 161 KV line.

REFER to AOP-31, 161 KV Grid Malfunctions,Section II, All 4160 V Buses CRS Fed from 22 KV.

Booth Operator: If contacted, REPORT as T&D System Operation that cause of 161 KV line loss appears to be lightning strike in Fort Calhoun Station Switchyard. If requested, report repair teams are being dispatched.

Examiner Note: The following steps are from AOP-31, 161 KV Grid Malfunctions,Section II, All 4160 V Buses Fed from 22 KV.

CAUTION To protect Bus 1A1 in the event of a fault, FW-2A and FW-4A should not both be left running when the Feedwater System is realigned.

DETERMINE Reactor power is 50% and ENSURE all the following CRS conditions are satisfied: [Step 4.1]

BOPO

DETERMINE two Condensate Pumps are RUNNING. [Step 4.1.a]

DETERMINE two Feedwater Pumps are RUNNING. [Step 4.1.b]

DETERMINE two Heater Drain Pumps are RUNNING. [Step 4.1.c]

NRC Simulator Scenario 2 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 2 Event # 5 Page 19 of 33 Event

Description:

Loss of 161 KV Line Time Position Applicants Actions or Behavior BOPO ADJUST Main Generator terminal voltage to less than 22,000 Volts. [Step 4.2]

NOTIFY Energy Marketing of the need to adjust voltage. [Step 4.2.a]

ADJUST Voltage Regulator per OI-ST-1, Turbine Generator Normal Operation. [Step 4.2.b]

VERIFY terminal voltage is less than 22,000 volts. [Step 4.2.c]

ESTABLISH balanced 4160 V Bus loading on T1A1 and T1A2 by ensuring CRS ALL of the following pumps on Bus 1A1 are operating: [Step 4.3]

DETERMINE FW-2A, CONDENSATE Pump is NOT running and BOPO REFER to OI-FW-1 Condensate System Normal Operation. [Step 4.3.a]

DETERMINE FW-4A, Main Feedwater Pump is RUNNING. [Step 4.3.b]

DETERMINE FW-5A, Heater Drain Pump is RUNNING. [Step 4.3.c]

Examiner Note: The following steps are from OI-ST-1, Turbine Generator Normal Operation, Attachment 6, Generator VAR Adjustments (Automatic Mode).

NOTE Lowering voltage of the Main Generator while synchronized to the grid may cause low voltage indications on T1A1 and T1A2.

BOPO PERFORM the following to lower Generator Reactive Load (VARS): [Step 1]

ROTATE Generator G1 AC Regulator Voltage Adjuster (90P) in the COUNTER-CLOCKWISE direction until desired load is attained, as indicated on VAR/G1. [Step 1.a]

PLACE CS-70E/G1F, Generator G1 DC Regulator Voltage Adjuster (70P), in the LOWER position until the V/G1R, Generator ST-2 Voltage Regulator Transfer Voltage, reads 0 Volts DC. [Step 1.b]

Examiner Note: The following steps are from OI-FW-1, Condensate System Normal Operation, Attachment 4, Rotating Condensate Pumps.

NOTE FW-2B or FW-2C are the preferred pumps during one pump operation because of their ability to continue to run after a SIAS or CSAS actuation.

NRC Simulator Scenario 2 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 2 Event # 5 Page 20 of 33 Event

Description:

Loss of 161 KV Line Time Position Applicants Actions or Behavior BOPO SELECT Condensate Pump to be started: [Step 1]

FW-2A STA SUSPEND GARDEL data feed per OI-ERFCS-2, Attachment 6. [Step 2]

CAUTIONS

1. The 43/FW Switch affects the Auto-Start operation of the Main Condensate Pumps, Main Feedwater Pumps and Heater Drain Pumps.

2. The Standby (Auto-Start) feature for these pumps will be inhibited when the 43/FW Switch is placed in off.

3. The 43/FW Switch must be in off to start a Condensate Pump.

BOPO PERFORM the following at CB-10/11: [Step 3]

PLACE Cond. & FW Pumps Transfer Switch 43/FW to OFF. [Step 3.a]

VERIFY annunciator CB-10/11/A10, Window B-6L - 43/FW TRANSFER SWITCH OFF-AUTO in alarm. [Step 3.b]

NOTE During rotation of the Condensate Pumps, at the time the designated standby pump is started, declare XC105 INVALID and log in the Control Room Log. Once pump rotation is complete, and the 12-minute validity period has passed (Ref. OI-ERFCS-3), the STA should review all XC105 input parameters and determine they are at steady-state, then XC105 can be declared valid and available for monitoring reactor core output.

BOPO START Condensate Pump FW-2A. [Step 4]

VERIFY FW-2A ammeter returns to < 250 amps in < 15 seconds and BOPO STABILIZES on CB-10/11. [Step 5]

Booth Operator: When contacted, REPORT FW-2A discharge pressure of ~520 psig.

BOPO ENSURE Condensate Pump minimum flow is being maintained: [Step 6]

VERIFY discharge pressure of 490-600 psig at FW-2A. [Step 6.a]

Booth Operator: When contacted, REPORT all FW-2A parameters are normal.

NRC Simulator Scenario 2 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 2 Event # 5 Page 21 of 33 Event

Description:

Loss of 161 KV Line Time Position Applicants Actions or Behavior BOPO MONITOR Condensate Pump FW-2A parameters: [Step 7]

CHECK for unusual noise or vibration. [Step 7.a]

VERIFY lube oil levels in middle of sightglass. [Step 7.b]

CHECK PI-1214, Seal Water inlet pressure between 70 and 90 psig.

[Step 7.c]

CHECK PI-1232A/B/C, FW-2A/B/C Discharge Pressure at 490-600 psig at pump. [Step 7.d]

CHECK PI-1181A/B/C, FW-2A/B/C Discharge Pressure at 490-600 psig BOPO on CB-10/11. [Step 7.e]

BOPO

CHECK flow and temperatures on ERF Computer: [Step 7.f]

F1172, PRINT XC092 FOR COND. FLOW

T1179A, COND PMP A DISCH HDR TEMP

T1179B, COND PMP B DISCH HDR TEMP

T1184A/B/C, COND PMP A/B/C MTR OUT BRG TEMP

T1185A/B/C, COND PMP A/B/C MTR IN BRG TEMP NOTE Condensate Pump Control Switch should be positioned in AFTER-STOP or PULL-STOP per the Shift Manager or CRS, to prevent possible water hammer at power levels below 50%.

BOPO STOP Condensate Pump FW-2C. [Step 8]

BOPO PLACE FW-2C Condensate Pump Control Switch in AFTER-STOP. [Step 9]

BOPO VERIFY FW-2C Condensate Pump ammeter drops to 0. [Step 10]

Booth Operator: When contacted, REPORT no reverse rotation on FW-2C.

BOPO VERIFY FW-2C NOT rotating in reverse direction. [Step 11]

DETERMINE 43-SIAS/FW2, Post-SIAS/CSAS Running Condensate Pump BOPO Switch in FW-2B position. [Step 12]

BOPO PERFORM the following at CB-10/11: [Step 13]

PLACE 43/FW Switch in AUTO. [Step 13.a]

NRC Simulator Scenario 2 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 2 Event # 5 Page 22 of 33 Event

Description:

Loss of 161 KV Line Time Position Applicants Actions or Behavior

VERIFY annunciator CB-10/11/A10, Window B-6L - 43/FW TRANSFER SWITCH OFF-AUTO is clear. [Step 13.b]

STA RESTORE GARDEL data feed per OI-ERFCS-2, Attachment 7. [Step 14]

Examiner Note: The following steps continue from AOP-31,Section II.

BOPO DETERMINE all 480 V Buses greater than 430 volts: [Step 4.4]

OBSERVE Bus 1B3A voltage at ~470 V.

OBSERVE Bus 1B3B voltage at ~470 V.

OBSERVE Bus 1B3C voltage at ~470 V.

OBSERVE Bus 1B4A voltage at ~470 V.

OBSERVE Bus 1B4B voltage at ~460 V.

OBSERVE Bus 1B4C voltage at ~470 V.

NOTIFY NRC Operation Center within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months of loss of 161 KV Line.

CRS

[Step 4.5]

Examiner Note: The Loss of Offsite Power event is triggered 30 seconds after the flag is matched on Breaker 1A44.

BOPO MATCH flags on all the following breakers: [Step 4.6]

Breaker 110 flag MATCHED.

Breaker 111 flag MATCHED.

Breaker 1A31 flag already matched.

Breaker 1A33 flag MATCHED.

Breaker 1A42 flag already matched.

Breaker 1A44 flag MATCHED.

When Breaker 1A44 flag is MATCHED, PROCEED to Events 6, 7, 8, and 9.

NRC Simulator Scenario 2 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 2 Event # 6, 7, 8, & 9 Page 23 of 33 Event

Description:

Loss of Offsite Power / Four Stuck CEAs / Train A Diesel Generator Breaker Failure / Train B Diesel Generator Overspeed Trip / Diesel Driven Auxiliary Feedwater Pump Start Failure Time Position Applicants Actions or Behavior Booth Operator: When the control switch is matched for Breaker 1A44, Events 6, 7, 8, and 9 will automatically execute.

- Loss of Offsite Power.

- Four Stuck CEAs on Reactor Trip.

- Diesel Generator DG-01 Output Breaker failure.

- Diesel Generator DG-02 overspeed trip.

- Diesel Driven Auxiliary Feedwater Pump FW-54 start failure.

Indications Available:

Numerous Reactor Trip and Loss of Offsite Power Alarms.

CREW RECOGNIZE Reactor Trip due to Loss of Offsite Power.

CRS DIRECT performance of EOP-00, Standard Post Trip Actions.

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

ATCO VERIFY Reactivity Control: [Step 1]

VERIFY ALL of the following: [Step 1.a]

DETERMINE more than one Regulating or Shutdown CEA NOT inserted.

VERIFY Reactor Power is LOWERING.

VERIFY Startup Rate is NEGATIVE.

DETERMINE uncontrolled RCS Cooldown NOT in progress. [Step 1.b]

Examiner Note: Applicant will be unable to Emergency Borate until power is restored in EOP-20, Functional Recovery, due to the Station Blackout (SBO).

[CA] If more than one CEA is NOT fully inserted, PERFORM the following to initiate Emergency Boration: [Step 1.2]

[CA] ENSURE both following valves CLOSED: [Step 1.2.a]

[CA] FCV-269X, Demin Water Makeup Valve.

[CA] FCV-269Y, Boric Acid Makeup Valve.

[CA] OPEN all the following valves: [Step 1.2.b]

NRC Simulator Scenario 2 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 2 Event # 6, 7, 8, & 9 Page 24 of 33 Event

Description:

Loss of Offsite Power / Four Stuck CEAs / Train A Diesel Generator Breaker Failure / Train B Diesel Generator Overspeed Trip / Diesel Driven Auxiliary Feedwater Pump Start Failure Time Position Applicants Actions or Behavior

[CA] HCV-268, Boric Acid Pump Header to Charging Pumps Isolation Valve.

[CA] HCV-265, CH-11A Gravity Feed Valve.

[CA] HCV-258, CH-11B Gravity Feed Valve.

[CA] START all the following pumps: [Step 1.2.c]

[CA] Boric Acid Pump CH-4A.

[CA] Boric Acid Pump CH-4B.

[CA] Charging Pump CH-1A.

[CA] Charging Pump CH-1B.

[CA] Charging Pump CH-1C.

[CA] CLOSE LCV-218-2, VCT Outlet Valve. [Step 1.2.d]

[CA] ENSURE the following valves CLOSED: [Step 1.2.e]

[CA] LCV-218-3, Charging Pump Suction SIRWT Isolation Valve

[CA] HCV-257, CH-4B Recirc Valve.

[CA] HCV-264, CH-4A Recirc Valve.

[CA] BORATE until adequate Shutdown Margin is established.

[Step 1.2.f]

CRS DETERMINE Reactivity Control criteria NOT SATISFIED.

BOPO VERIFY Turbine Trip: [Step 2]

VERIFY HP & LP Stop and Intercept Valves CLOSED.

BOPO ENSURE all Generator Breakers are tripped: [Step 3]

DETERMINE Generator Output Breaker 3451-4 tripped.

DETERMINE Generator Output Breaker 3451-5 tripped.

DETERMINE Generator Field Breaker 41E/G1F tripped.

DETERMINE both 4160 V Safeguards Buses 1A3 & 1A4 are DEENERGIZED.

BOPO

[Step 4]

NRC Simulator Scenario 2 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 2 Event # 6, 7, 8, & 9 Page 25 of 33 Event

Description:

Loss of Offsite Power / Four Stuck CEAs / Train A Diesel Generator Breaker Failure / Train B Diesel Generator Overspeed Trip / Diesel Driven Auxiliary Feedwater Pump Start Failure Time Position Applicants Actions or Behavior

[CA] PERFORM the following with either Bus 1A3 or Bus 1A4 DEENERGIZED. [Step 4.1]

Booth Operator: When contacted, Wait 5 minutes, then REPORT minimizing DC loads in progress.

Booth Operator: When contacted about condition of Diesel Generators, WAIT 1 minute and REPORT DG-01 Output Breaker overcurrent relays are TRIPPED. Electrical Maintenance on station investigating breaker replacement.

Booth Operator: When contacted about condition of Diesel Generators, WAIT 1 minute and REPORT DG-02 tripped with oil vapor in the room.

Booth Operator: When contacted, EXECUTE local actions to align emergency boration by manually opening HCV-268 and manually closing LCV-218-3, and local alignment of potable water cooling to an air compressor.

[CA] Minimize DC Loads within 15 minutes of loss of bus per MVA-24, Minimizing DC Loads. [Step 4.1.a]

[CA] DEPRESS Diesel Generator EMERGENCY START pushbuttons. [Step 4.1.b]

CRS DETERMINE Maintenance of Vital Auxiliaries criteria NOT SATISFIED.

DETERMINE Safety Injection Actuation Signal has NOT occurred and DG-01 BOPO is RUNNING and DG-02 is STOPPED. [Step 5]

DETERMINE 4160 V Non-Safeguards Buses 1A1 and 1A2 are BOPO DEENERGIZED. [Step 6]

BOPO DETERMINE 125 VDC Buses 1 and 2 are ENERGIZED. [Step 7]

BOPO VERIFY Instrument Air is AVAILABLE: [Step 8]

DETERMINE Instrument Air pressure < 90 psig.

DETERMINE Instrument Air Compressors NOT RUNNING due to loss of power.

[CA] If Instrument Air pressure is < 90 psig, PERFORM the following to restore Instrument Air: [Step 8.1]

NRC Simulator Scenario 2 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 2 Event # 6, 7, 8, & 9 Page 26 of 33 Event

Description:

Loss of Offsite Power / Four Stuck CEAs / Train A Diesel Generator Breaker Failure / Train B Diesel Generator Overspeed Trip / Diesel Driven Auxiliary Feedwater Pump Start Failure Time Position Applicants Actions or Behavior BOPO * [CA] START a Bearing Water Pump.

BOPO * [CA] START an Air Compressor.

ATCO VERIFY Component Cooling Water System operation NORMAL: [Step 9]

DETERMINE NO CCW Pumps RUNNING. [Step 9.a]

DETERMINE CCW Pump discharge pressure < 60 psig. [Step 9.b]

[CA] DETERMINE all Reactor Coolant Pumps STOPPED.

[Step 9.1]

DETERMINE HCV-438A/B/C/D, CCW to RCP Coolers OPEN. [Step 9.c]

DETERMINE NO Raw Water Pumps RUNNING. [Step 9.d]

CRS VERIFY RCS Inventory Control criteria satisfied: [Step 10]

DETERMINE PZR level between 30% and 70% and TRENDING to ATCO between 45% and 60%.

DETERMINE RCS subcooling > 20°F.

CRS DETERMINE RCS Inventory Control criteria SATISFIED.

CRS VERIFY RCS Pressure Control criteria satisfied: [Step 11]

ATCO

DETERMINE RCS pressure between 1800 psia and 2300 psia.

DETERMINE RCS pressure TRENDING between 2050 psia and 2150 psia.

DETERMINE PORVs are CLOSED.

CRS DETERMINE RCS Pressure Control criteria SATISFIED.

CRS VERIFY Core Heat Removal criteria satisfied: [Step 12]

ATCO

DETERMINE RCP NPSH requirements met.

DETERMINE all RCPs STOPPED.

[CA] PLACE TCV-909, Temperature Controller in MANUAL on DCS.

BOPO

[Step 12.2.a]

NRC Simulator Scenario 2 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 2 Event # 6, 7, 8, & 9 Page 27 of 33 Event

Description:

Loss of Offsite Power / Four Stuck CEAs / Train A Diesel Generator Breaker Failure / Train B Diesel Generator Overspeed Trip / Diesel Driven Auxiliary Feedwater Pump Start Failure Time Position Applicants Actions or Behavior

[CA] ENSURE TCV-909, Temperature Controller OUTPUT is zero BOPO (0). [Step 12.2.b]

CRS * [CA] VERIFY Natural Circulation in at least one Loop. [Step 12.2.c]

[CA] DETERMINE Core T 50°F.

[CA] DETERMINE difference between CETs and RCS THOT is 10°F on ERF "CHR" display.

[CA] DETERMINE RCS subcooling is 20°F.

[CA] DETERMINE THOT and TCOLD are stable or lowering.

CRS DETERMINE Core Heat Removal criteria NOT SATISFIED.

NOTE If Instrument Air to valves HCV-1105, HCV-1106, HCV-1107A/B and HCV-1108A/B is not available, throttling of these valves is not possible.

Open or close operation of these valves is possible for a minimum of three cycles.

CRS VERIFY RCS Heat Removal criteria satisfied:

Examiner Note: Preferred method to feed SGs during an SBO is via FW-54 (aligned through the SG Feed Ring). Applicant must recognize that FW-54 fails to start and place FW-10, Steam Driven AFW Pump, in service. There are no automatic initiated actions to restore Feedwater flow in this Scenario.

Booth Operator: If contacted about FW-54, WAIT 5 minutes and REPORT pump appears damaged.

BOPO DETERMINE Main Feedwater is NOT restoring SG levels. [Step 13]

[CA] If Main Feedwater is NOT restoring S/G level and SGLS has NOT actuated, ESTABLISH Feedwater by performing step a, b, c, d, or e: [Step 13.1]

[CA] DETERMINE Main Feedwater NOT available. [Step 13.1.a]

[CA] DETERMINE AFW Pump FW-54 did NOT start. [Step 13.1.b]

[CA] DETERMINE AFW Pump FW-06 NOT available. [Step 13.1.c]

NRC Simulator Scenario 2 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 2 Event # 6, 7, 8, & 9 Page 28 of 33 Event

Description:

Loss of Offsite Power / Four Stuck CEAs / Train A Diesel Generator Breaker Failure / Train B Diesel Generator Overspeed Trip / Diesel Driven Auxiliary Feedwater Pump Start Failure Time Position Applicants Actions or Behavior CRITICAL TASK Restore Feedwater Flow to At Least One Steam Generator to Reestablish any STATEMENT SG as a Heat Sink prior to AFAS actuating.

CRITICAL * [CA] INITIATE AFW using FW-10, AFW Pumps to AFW Nozzles:

TASK BOPO

[Step 13.1.c]

[CA] START AFW Pump FW-10 at AI-66. [Step 13.1.c.1)]

[CA] RESTORE level in at least one SG to 35% to 85% NR or 73% to 94% WR via AFW Nozzles. [Step 13.1.c.2)]

OPEN HCV-1107A at AFW Panel AI-66.

OPEN HCV-1108A at AFW Panel AI-66.

OPEN HCV-1107B at AFW Panel AI-66.

OPEN HCV-1108B at AFW Panel AI-66.

DETERMINE FCV-1101 & FCV-1102 Feed Regulating Valves CLOSED.

[Step 13.a]

DETERMINE HCV-1105 & HCV-1106 Feed Regulating Bypass Valves ramped to between 40% & 45% OPEN. [Step 13.b]

BOPO

PLACE 43/FW switch in OFF. [Step 13.c]

DETERMINE NO Main Feedwater Pumps RUNNING. [Step 13.d]

DETERMINE NO Condensate Pumps RUNNING. [Step 13.e]

DETERMINE NO Heater Drain Pumps RUNNING. [Step 13.f]

BOPO

ENSURE SG Blowdown Isolation Valves CLOSED. [Step 13.g]

HCV-1387A & HCV-1387B

HCV-1388A & HCV-1388B VERIFY Steam Dump and Bypass Valves controlling both of the following:

BOPO

[Step 14]

DETERMINE RCS TCOLD between 525°F and 535°F.

DETERMINE Steam Generator pressures ~1000 psia.

[CA] If TCOLD greater than 525°F, PERFORM the following:

[Step 14.1]

[CA] DETERMINE Steam Dump and Bypass Valves NOT available. [Step 14.1.a]

NRC Simulator Scenario 2 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 2 Event # 6, 7, 8, & 9 Page 29 of 33 Event

Description:

Loss of Offsite Power / Four Stuck CEAs / Train A Diesel Generator Breaker Failure / Train B Diesel Generator Overspeed Trip / Diesel Driven Auxiliary Feedwater Pump Start Failure Time Position Applicants Actions or Behavior

[CA] CONTROL HCV-1040, Atmospheric Dump Valve as BOPO required. [Step 14.1.b]

[CA] IF REQUIRED, OPERATE Air Assisted Main Steam Safety Valves MS-291 and MS-292. [Step 14.1.c]

CRS DETERMINE RCS Heat Removal criteria SATISFIED.

CRS VERIFY Containment Isolation criteria satisfied:

ATCO VERIFY Normal Containment conditions exist: [Step 15]

DETERMINE no unexpected rise in Containment Sump level. [Step 15.a]

DETERMINE Containment Area Radiation Monitors NOT in alarm.

[Step 15.b]

DETERMINE Containment Ventilation Radiation Monitors NOT in alarm.

[Step 15.c]

DETERMINE SG Blowdown and Condenser Off Gas Radiation Monitors NOT alarming. [Step 15.d]

DETERMINE SG Blowdown and Condenser Off Gas Radiation Monitors ATCO NOT TRENDING to alarm. [Step 15.e]

CRS DETERMINE Containment Integrity criteria SATISFIED.

CRS DIAGNOSE event in progress: [Step 16]

DETERMINE Reactivity Control requirements NOT met.

If not, GO TO EOP-20, Functional Recovery.

Booth Operator: When EOP-20 is entered, If previously contacted for repairs, REPORT as Electrical Maintenance that the DG-01 Output Breaker has been replaced and the area cordoned off for closure. Request Breaker controller to be placed in Pull-to-Lock so that the breaker can be racked up. After 2 minutes, remove malfunction and report that the breaker is ready for closure.

Examiner Note: The following steps are from EOP-20, Functional Recovery.

NRC Simulator Scenario 2 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 2 Event # 6, 7, 8, & 9 Page 30 of 33 Event

Description:

Loss of Offsite Power / Four Stuck CEAs / Train A Diesel Generator Breaker Failure / Train B Diesel Generator Overspeed Trip / Diesel Driven Auxiliary Feedwater Pump Start Failure Time Position Applicants Actions or Behavior CRS CONFIRM Standard Post Trip Actions have been performed. [Step 1]

CRS IMPLEMENT the Emergency Plan. [Step 2]

Time: __________

CREW MONITOR the Floating Steps. [Step 3]

CRS DETERMINE Feedwater flow has NOT been lost. [Step 4]

ATCO DETERMINE all RCPs are STOPPED. [Step 5]

DETERMINE that CIAS has NOT occurred and DIRECT Shift Chemist to CRS sample both Steam Generators. [Step 6]

IDENTIFY EOP-20 Success Path to satisfy each Safety Function using CRS Safety Function Status Checks or Resource Assessment Trees. [Step 7]

VERIFY Reactivity Control NOT SATISFIED and CONSIDER Reactivity Control - Resource Tree A, RC-2: Boration using CVCS, Condition 2.

VERIFY Maintenance of Vital Auxiliaries NOT SATISFIED and CONSIDER Maintenance of Vital Auxiliaries - Resource Tree B, MVA-AC: Restoration of AC.

DETERMINE RCS Inventory Control SATISFIED.

DETERMINE RCS Pressure Control SATISFIED.

DETERMINE RCS and Core Heat Removal SATISFIED.

DETERMINE Containment Integrity SATISFIED.

Examiner Note: The following steps are from EOP-20, Functional Recovery, Section 10, Maintenance of Vital Auxiliaries - AC.

DETERMINE NO 4160 V Safeguards Bus is energized and Reactivity CRS Control Safety Function is in jeopardy. [Step 10.1]

CRS DETERMINE both 4160 V Safeguards Buses are DEENERGIZED. [Step 10.2]

NRC Simulator Scenario 2 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 2 Event # 6, 7, 8, & 9 Page 31 of 33 Event

Description:

Loss of Offsite Power / Four Stuck CEAs / Train A Diesel Generator Breaker Failure / Train B Diesel Generator Overspeed Trip / Diesel Driven Auxiliary Feedwater Pump Start Failure Time Position Applicants Actions or Behavior

[CA] PERFORM step A or B to RESTORE deenergized bus.

[Step 10.2.1]

[CA] If 1A3 is deenergized, GO TO Step 3. [Step 10.2.1.a]

BOPO VERIFY NONE of the following Lockout Relays are tripped: [Step 10.3]

86/1A13

86/1A33

86/1A3-TFB If Bus 1A3 is deenergized and DG-1 is running, PERFORM the following to BOPO ENERGIZE Bus 1A3: [Step 10.4]

OPEN all the following breakers: [Step 10.4.a]

1A33

1A13

FW-6, Electric AFW Pump

RC-3C, RCP

AC-10A, RW Pump

AC-10C, RW Pump

SI-1A, LPSI Pump Restore Power to any 4160 V Safeguards Bus using a Diesel Generator to CRITICAL TASK Reestablish Maintenance of Vital Auxiliaries and Allow Branching to Meet STATEMENT other Safety Functions During a Station Blackout.

CRITICAL

If DG-1 frequency is > 60 Hz and voltage is > 4160 V, CLOSE breaker TASK BOPO 1AD1. [Step 10.4.b]

PLACE Breaker 1AD1 in CLOSE.

Examiner Note: Breaker will close when taken out of Pull-to-Lock if repairs have been made.

Time: _______

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

NRC Simulator Scenario 2 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 2 Event # 6, 7, 8, & 9 Page 32 of 33 Event

Description:

Loss of Offsite Power / Four Stuck CEAs / Train A Diesel Generator Breaker Failure / Train B Diesel Generator Overspeed Trip / Diesel Driven Auxiliary Feedwater Pump Start Failure Time Position Applicants Actions or Behavior Commence an Emergency Boration of the RCS Due to 2 or more Stuck CEAs CRITICAL TASK when Diesel Generator DG-1 Breaker is Closed and Bus 1A3 is Reenergized to STATEMENT Restore Reactivity Control.

CRITICAL * [CA] If more than one CEA is NOT fully inserted, PERFORM the TASK ATCO following to initiate Emergency Boration: [Step 1.2]

[CA] ENSURE both following valves CLOSED: [Step 1.2.a]

[CA] FCV-269X, Demin Water Makeup Valve.

[CA] FCV-269Y, Boric Acid Makeup Valve.

[CA] ENSURE all the following valves OPEN: [Step 1.2.b]

[CA] HCV-268, Boric Acid Pump Header to Charging Pumps Isolation Valve

[CA] HCV-265, CH-11A Gravity Feed Valve.

[CA] HCV-258, CH-11B Gravity Feed Valve.

[CA] START all the following pumps: [Step 1.2.c]

[CA] Boric Acid Pump CH-4A.

[CA] Charging Pump CH-1A.

[CA] CLOSE LCV-218-2, VCT Outlet Valve. [Step 1.2.d]

[CA] ENSURE the following valves CLOSED: [Step 1.2.e]

Booth Operator: When contacted, EXECUTE remote function to LOCALLY CLOSE LCV-218-3, Charging Pump Suction SIRWT Isolation Valve.

[CA] LOCALLY OPEN LCV-218-3, Charging Pump Suction SIRWT Isolation Valve.

[CA] HCV-257, CH-4B Recirc Valve.

[CA] HCV-264, CH-4A Recirc Valve.

[CA] BORATE until adequate Shutdown Margin is established.

[Step 1.2.f]

NRC Simulator Scenario 2 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 2 Event # 6, 7, 8, & 9 Page 33 of 33 Event

Description:

Loss of Offsite Power / Four Stuck CEAs / Train A Diesel Generator Breaker Failure / Train B Diesel Generator Overspeed Trip / Diesel Driven Auxiliary Feedwater Pump Start Failure Time Position Applicants Actions or Behavior Examiner Note: The following steps continue from EOP-20, Functional Recovery.

CRS VERIFY Safety Functions are being satisfied at 15 minute intervals. [Step 8]

If Safety Function Status Check Acceptance Criteria are satisfied, CRS PERFORM instructions for all Success Paths in use. [Step 9]

IMPLEMENT Section 18, Long Term Actions, when both of the following are CRS SATISFIED: [Step 10]

INSTRUCTIONS for all Success Paths have been performed.

Safety Function Status Check Acceptance Criteria for Success Paths in use are being SATISFIED.

When Emergency Boration is initiated, TERMINATE the scenario.

NRC Simulator Scenario 2 Outline Rev. 6

Appendix D Scenario Outline Form ES-D-1 Facility: Fort Calhoun Station Scenario No.: 3 Op Test No.: Dec 2015 NRC Examiners: Operators:

Initial Conditions: 100% power MOL - RCS Boron is 482 ppm (by sample).

Turnover: Maintain steady-state power conditions. Rotate Heater Drain Pumps FW-5B and FW-5C per OI-VD-1, Feedwater Heater Vents and Drains Normal Operation. Charging Pump CH-1C out of service for packing repair.

Critical Tasks:

Stop One Reactor Coolant Pump in Each Loop when Reactor Coolant System Pressure is

< 1350 psia, Prior to losing Reactor Coolant Pump Net Positive Suction Head. (Event 7)

Isolate the Affected Steam Generator with a Tube Rupture to Minimize Spread of Contamination. (Event 7)

Reduce and Maintain RCS THOT 510°F to Maintain SG Pressure Below Safety Valve Setpoint of 1000 psia. (Event 7)

Event No. Malf. No. Event Type* Event Description 1 N (BOPO) Rotate Heater Drain Pumps per OI-VD-1, Feedwater Heater Vents

+10 min and Drains Normal Operation, Attachment 2.

2 I (ATCO, CRS) Pressurizer Level Channel Transmitter LT-101X Fails Low.

+20 min Transfer Pressurizer Level Control to LT-101Y.

3 I (BOPO, CRS) Steam Generator RC-2A Steam Flow Transmitter FT-907 Fails

+30 min High. Bypass Affected Transmitter.

4 C (ATCO, CRS) Charging Pump CH-1A Trip.

+40 min TS (CRS) Restore Letdown and Charging Flow.

5 C (ATCO,BOPO, Steam Generator RC-2B Tube Leak Greater Than 150 GPD.

+50 min CRS) TS (CRS) Isolate Blowdown Flow.

6 R (ATCO) Commence Plant Shutdown per AOP-05, Emergency Shutdown.

+60 min N (BOPO, CRS) 7 M (ATCO, BOPO, Steam Generator RC-2B Tube Rupture at 500 GPM on 10 Minute

+70 min CRS) Ramp Upon 3% to 5% Load Reduction.

8 I (BOPO) Diesel Generator DG-01 Start Failure on SIAS.

+70 min Manual Start Required.

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Technical Specifications Actual Target Quantitative Attributes 1 Malfunctions after EOP entry (1-2) 4 Abnormal events (2-4) 1 Major transients (1-2) 1 EOPs entered/requiring substantive actions (1-2) 0 EOP contingencies requiring substantive actions (0-2) 3 Critical tasks (2-3)

NRC Simulator Scenario 3 Outline Rev. 6

Scenario Event Description NRC Scenario 3 SCENARIO

SUMMARY

NRC 3 The crew will assume the shift at 100% power per OP-4, Load Change and Normal Power Operation.

The scheduled activity is to rotate Heater Drain Pumps by starting FW-5C and securing FW-5B per OI-VD-1, Feedwater Heater Vents and Drains Normal Operation, Attachment 2, Rotating Operating Heater Drain Pumps.

The next event is a low failure of Pressurizer Level Control Channel, LT-101X. Operator actions are per ARP-CB-1/2/3/A4, Window C PRESSURIZER LEVEL LO-LO CHANNEL X. The crew will transfer to the standby channel LT-101Y and restore Letdown per OI-RC-8, Reactor Coolant System Level Control Normal Operation, Attachment 8, Transferring Pressurizer Level Control Channel in CASCADE and Attachment 4, Transferring Letdown Controller from AUTOMATIC to MANUAL.

When plant conditions are stable, a high failure of Steam Generator RC-2A Steam Flow Transmitter FT-907 will occur. Initial operator actions are per ARP-DCS-FW, Feedwater DCS Annunciator Response Procedure and include verifying Feedwater Control is in Single Element Control, bypassing the failed input, and determining 3 Element Control is restored.

The next event is a trip of the running Charging Pump. Operator actions are per ARP-CB-1/2/3/A2, Window A-6L - CHARGING FLOW LO and include isolating of Letdown and verifying no system leaks exist. Charging Pump CH-1B is placed in service per OI-CH-1, Chemical and Volume Control System Normal Operation, Attachment 1, Startup of Charging and Letdown. The SRO will refer to Technical Specification LCO 2.2.4 - Charging Pumps - Operating.

When Charging flow is restored, a Steam Generator Tube Leak of greater than 150 gallons per day will occur on Steam Generator RC-2B. The crew will enter AOP-22, Reactor Coolant Leak, and implement Attachment B, Primary to Secondary Leak Rate Actions. RM-064, Main Steam Line Radiation Monitor, is placed in service to assist in determining leak size and location. Various Secondary Side valves are closed to minimize system contamination and HR-21, Blowdown Operation is performed to isolate blowdown flow from SG RC-2B. The SRO will refer to Technical Specification LCO 2.1.4 - Reactor Coolant System Leakage Limits.

Once blowdown is isolated, entry into AOP-05, Emergency Shutdown, is performed to bring the plant into MODE 4. When power has been reduced 3% to 5%, a Steam Generator Tube Rupture of 500 gpm will commence on a 10 minute ramp.

The crew enters EOP-00, Standard Post Trip Actions, and then transitions to EOP-04, Steam Generator Tube Rupture. Diesel Generator DG-01 fails to start upon SIAS and must be manually started. While in EOP-04, the Reactor Coolant System is cooled per HR-12, Secondary Heat Removal Operation, and the RCS is depressurized to less than 1000 psia per PC-11, Pressure Control, to allow isolating the affected Steam Generator. When SG RC-2B is isolated, the scenario is terminated.

Risk Significance:

Failure of risk important system prior to trip: Loss of Charging Pump Steam Generator Tube Leak

Risk significant core damage sequence: Steam Generator Tube Rupture

Risk significant operator actions: Stop RCPs Upon Loss of Subcooling Isolate Affected Steam Generator Cooldown and Depressurize RCS NRC Simulator Scenario 3 Outline Rev. 6

Scenario Event Description NRC Scenario 3 BOOTH OPERATOR INSTRUCTIONS for SIMULATOR SETUP RESET to IC-#103 (or any 100% MOL IC) and LOAD & EXECUTE NRC 3.sce for NRC Scenario 3.

Preset Item - CH-1C Removed from Service Type Item Value Condition Malfunction BUS_1B3B_4B_5_BKR_Trip True Scenario Event: CH-1C OOS Preset Item - Event 9 - Diesel Generator #1 Auto Start Failure Type Item Value Condition Expert H_PD1_033_3 Reset Scenario Event: DG-1 H_PD1_031_3 Reset Auto Start Failure Event 2 - Pressurizer Level Transmitter LT-101X Fails Low Type Item Value Condition Transmitter RCS_LT101X 0, ramp = 5 seconds When directed by examiner, trigger/activate this event.

Scenario Event: Pzr Level LT-101X Fail Low Event 3 - Steam Generator Flow Transmitter LT-907 Fails High Type Item Value Condition Transmitter FT-907 4000000, ramp = 5 sec When directed by examiner, trigger/activate this event.

FT-907 DCS Fail High Scenario Event: SG Flow FT-907-1 DCS Fail High FT-907 Fail High Event 4 - Charging Pump CH-1A trips Type Item Value Condition Malfunction BUS_1B3A_4_BKR_TRIP True When directed by examiner, trigger/activate this event.

Scenario Event: CH-1A Trip Event 5 - Primary-to-Secondary SG Tube Leak Develops in Steam Generator RC-2B Type Item Value Condition Malfunction RCS04B 0.001 When directed by examiner, trigger/activate this event.

Scenario Event: RC-2B S/G Tube Leak Event 7 - Steam Generator Tube Leak in RC-2B Grows to Tube Rupture Type Item Value Condition Malfunction RCS04B 1.4, ramp = 600 sec When directed by examiner, trigger/activate this event.

Scenario Event: RC-2B S/G Tube Rupture NRC Simulator Scenario 3 Outline Rev. 6

Scenario Event Description NRC Scenario 3 Booth Operator: INITIALIZE to IC-1 and LOAD NRC 3.sce.

ENSURE all Simulator Annunciator Alarms are ACTIVE.

ENSURE all Control Board Tags are removed.

ENSURE Charging Pump CH-1A in service.

ENSURE Charging Pump CH-1C OOS for emulsified oil replacement with Information Tag attached.

ENSURE Channel X Pressurizer Pressure and Level selected.

ENSURE Reactivity Briefing Sheet printout provided with Turnover.

ENSURE Middle-of-Life Thumb Rule Sheet provided with Turnover.

ENSURE ERF Computer System Display set to FWD for BOPO.

ENSURE procedures in progress provided to crew in Briefing Room:

- COPY of OI-VD-1, Feedwater Heater Vents and Drains Normal Operation, Attachment 2, Rotating Operating Heater Drains Pumps, INITIALED through Prerequisites and Procedure Step 2.

Control Room Annunciators in Alarm:

NONE 0B Procedure List Event 1: OP-4, Load Change and Normal Power Operation.

Event 1: OI-VD-1, Feedwater Heater Vents and Drains Normal Operation Event 2: ARP-CB-1/2/3/A4, Window C-8, PRESSURIZER LEVEL LO-LO CHANNEL X Event 3: ARP-DCS-FW, Feedwater DCS Annunciator Response Procedure Event 4: ARP-CB-1/2/3/A2, Window A-6L, CHARGING FLOW LO Event 4: OI-CH-1, Chemical and Volume Control System Normal Operation, Attachment 1, Startup of Charging and Letdown Event 5: AOP-22, Reactor Coolant Leak Event 5: HR-21, Blowdown Operation Event 6: AOP-05, Emergency Shutdown Event 7: EOP-00, Standard Post Trip Actions Event 7: EOP-04, Steam Generator Tube Rupture Event 8: HR-12, Secondary Heat Removal Operation Event 8: PC-11, Pressure Control NRC Simulator Scenario 3 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 1 Page 5 of 34 Event

Description:

Rotate Heater Drain Pumps Time Position Applicants Actions or Behavior Booth Operator: When directed, RESPOND to requests from Control Room.

Indications Available:

NONE Examiner Note: The following steps are from OI-VD-1, Feedwater Heater Vents and Drains Normal Operation, Attachment 2.

BOPO PERFORM the following at CB-10, 11: [Step 3]

PLACE 43/FW Switch in OFF. [Step 3.a]

VERIFY Annunciator CB-10,11/A10, Window B-6L - 43/FW TRANSFER SWITCH OFF AUTO in ALARM. [Step 3.b]

Examiner Note: XC105 is the Computer (DCS) generated value for Secondary Calorimetric.

CRS DECLARE XC105 invalid. [Step 4]

Make plant announcement, then:

BOPO PLACE FW-5C, Heater Drain Pump control switch to AFTER-START at CB-10, 11. [Step 5]

VERIFY FW-5C, Heater Drain Pump ammeter returns to less than 80 amps BOPO in less than 15 seconds and STABILIZES at ~ 66 amps. [Step 6]

Booth Operator: If contacted, REPORT FCV-1216C is closed.

VERIFY FCV-1216C, Heater Drain Pump FW-5C Recirculation Control Valve BOPO CLOSES. [Step 7]

PLACE FW-5B, Heater Drain Pump control switch to AFTER-STOP at BOPO CB-10, 11. [Step 8]

NOTE Verification of Cooling Water Flow to the Seal cooler will be used to ensure Stuffing Box pressure is < 250 psig when Pressure Gauge PI-1192A, B, or C is out of service.

NRC Simulator Scenario 3 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 1 Page 6 of 34 Event

Description:

Rotate Heater Drain Pumps Time Position Applicants Actions or Behavior Booth Operator: If contacted, REPORT FW-5C discharge and stuffing box pressures normal.

BOPO MONITOR the following parameters on Heater Drain Pump FW-5C: [Step 9]

Motor amperage at ~66 amps.

PI-1269C, Pump Discharge pressure at ~160 psig on ERF Computer.

Heater Drain Tank level ~54% on CB-10, 11.

Bearing temperatures on ERF Display FWD normal.

PI-1192C, Stuffing Box pressure < 250 psig read locally.

Booth Operator: If contacted, REPORT FW-5B is not rotating in reverse.

CONTACT Auxiliary Operator to VERIFY FW-5B, Heater Drain Pump NOT BOPO ROTATING in reverse direction. [Step 10]

BOPO PERFORM the following at CB-10, 11: [Step 11]

PLACE 43/FW Switch in AUTO. [Step 11.a]

VERIFY Annunciator CB-10, 11/A10, Window B-6L - 43/FW TRANSFER SWITCH OFF AUTO is CLEAR. [Step 11.b]

Booth Operator: If contacted, REPORT Shift Technical Advisor will restore GARDEL.

CONTACT Shift Technical Advisor to RESTORE GARDEL data feed per CRS OI-ERFCS-2. [Step 12]

When 12 minute validity period has passed and parameters are steady-state, STA DECLARE XC105 valid and ENTER in Control Room Log. [Step 13]

When restoration of XC105 is discussed, PROCEED to Event 2.

NRC Simulator Scenario 3 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 2 Page 7 of 34 Event

Description:

Pressurizer Level Channel Transmitter Failure Time Position Applicants Actions or Behavior Booth Operator: When directed, EXECUTE Event 2.

- Pressurizer Level Channel Transmitter LT-101X fails low.

Indications Available:

CB-1,2,3/A4 - PRESSURIZER LEVEL LO-LO CHANNEL X CB-1,2,3/A4 - PRESSURIZER LEVEL HI-LO CHANNEL X Charging Pump CH-1B starts Letdown flow to minimum (~26 gpm)

+30 sec ATCO RESPOND to Annunciator Response Procedures.

ATCO INFORM CRS of Pressurizer Level Channel LT-101X failure.

Examiner Note: ATCO may Operate to Mitigate per OPD 4-09 and TRANSFER to Channel Y.

REFER to ARP-CB-1,2,3/A4, Window C PRESSURIZER LEVEL LO-LO CRS CHANNEL X.

Examiner Note: During this event, pressurizer pressure may decrease to less than 2075 psia.

If this occurs, the crew should address TS 2.10.4.5 for pressurizer low pressure.

Examiner Note: The following steps are from ARP-CB-1,2,3/A4, Window C-8, PRESSURIZER LEVEL LO-LO CHANNEL X.

ATCO VERIFY Pressurizer Level on LR-101X/LR-101Y. [Step 1]

If Pressurizer level is NOT low, PERFORM the following: [Step 1.1]

PLACE HC-101 to Channel Y per OI-RC-8. [Step 1.1.1]

If desired, PLACE HIC-101-1/101-2, Letdown Throttle Valves Controller to MANUAL per OI-RC-8. [Step 1.1.2]

PLACE HC-101-1, Pzr Heater Cutout Channel Select Switch, to Channel Y. [Step 1.1.3]

Examiner Note: The following steps are from OI-RC-8, Reactor Coolant System Level Control Normal Operation, Attachment 8, Transferring Pressurizer Level Control Channel (X to Y or Y to X) in CASCADE.

ATCO ENSURE both Level Controllers are in (C) CASCADE: [Step 1]

NRC Simulator Scenario 3 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 2 Page 8 of 34 Event

Description:

Pressurizer Level Channel Transmitter Failure Time Position Applicants Actions or Behavior

LC-101X-1, Pressurizer Level Controller

LC-101Y-1, Pressurizer Level Controller If desired, PLACE Letdown Controller HIC-101-1/101-2, Letdown Throttle ATCO Valves Controller in MANUAL per Attachment 4. [Step 2]

Examiner Note: The following steps are from OI-RC-8, Attachment 4, Transferring Letdown Controller from AUTOMATIC to MANUAL.

ENSURE Letdown Controller HIC-101-1/101-2, Letdown Throttle Valves ATCO Controller in AUTO. [Step 1]

ATCO PLACE HIC-101-1/101-2 Manual/Auto Transfer Switch to BAL. [Step 2]

ADJUST Manual Control Knob on HIC-101-1/101-2 until TOP SCALE ATCO indicates 50% (zero deviation; red pointer aligned with the red dot). [Step 3]

ATCO PLACE HIC-101-1/101-2 Manual/Auto Transfer Switch to MAN. [Step 4]

If necessary, MAKE adjustments to HIC-101-1/101-2 Manual Control Knob to ATCO MAINTAIN desired Pressurizer Level. [Step 5]

Examiner Note: The following steps continue from OI-RC-8, Attachment 8.

CAUTION Transfer from the Selected Controller to the Non-Selected Controller should not be performed until both controller outputs are approximately equal.

VERIFY Controller LR-101Y has INDICATED Pressurizer Level and ATCO PROGRAMMED Pressurizer Level Setpoint MATCHED prior to transfer.

[Step 3]

PLACE HC-101, Pressurizer Level Channel Selector Switch, to Channel Y.

ATCO

[Step 4]

NRC Simulator Scenario 3 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 2 Page 9 of 34 Event

Description:

Pressurizer Level Channel Transmitter Failure Time Position Applicants Actions or Behavior ENSURE Controller LC-101Y-1 is controlling INDICATED Pressurizer Level ATCO at PROGRAMMED Setpoint. [Step 5]

PUSH LC-101-1 & LC-101-2, Charging Pump Bistable Reset buttons on ATCO Reactor Regulating System Panel AI-4B and VERIFY all bistables are RESET. [Step 6]

If required, PLACE Letdown Controller HIC-101-1/101-2, Letdown Throttle ATCO Valves Controller, in AUTO per Attachment 3. [Step 7]

Examiner Note: The following steps are from OI-RC-8, Attachment 3, Transferring Letdown Controller from MANUAL to AUTOMATIC.

ENSURE Letdown Controller HIC-101-1/101-2, Letdown Throttle Valves ATCO Controller is in (M) MANUAL. [Step 1]

Manually ADJUST HIC-101-1/101-2, Letdown Throttle Valves Controller and ATCO PIC-210, Letdown Press Controller until following parameters are met: [Step 2]

Indicated Pressurizer Level matches the Programmed Pressurizer Level Setpoint on LR-101X or LR-101Y, Pressurizer Level Recorder.

PIC-210 is maintaining 200 psi to 400 psi.

ADJUST bias knob on HIC-101-1/101-2 until the top scale indicates 50%

ATCO (zero deviation; red pointer aligned with the red dot). [Step 3]

PLACE HIC-101-1/101-2 Manual/Auto Transfer Switch to BAL, then to ATCO AUTO. [Step 4]

If necessary, ADJUST the bias knob of HIC-101-1/101-2 to ENSURE ATCO Indicated Pressurizer Level is maintained at Programmed Pressurizer Level setpoint. [Step 5]

Examiner Note: The following steps continue from ARP-CB-1,2,3/A4, Window C-8.

ATCO VERIFY RCS Pressure on PR-103X/PR-103Y > 1600 psia. [Step 2]

NRC Simulator Scenario 3 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 2 Page 10 of 34 Event

Description:

Pressurizer Level Channel Transmitter Failure Time Position Applicants Actions or Behavior ATCO ENSURE all Pressurizer Heaters DEENERGIZED. [Step 3]

DETERMINE RCS Cold Leg temperatures on A-D/TI-112C and A-D/TI-122C ATCO are NOT lowering. [Step 4]

CHECK VCT level on LI-219, for indication of lowering level. [Step 4.1]

DETERMINE VCT level is NOT lowering. [Step 4.2]

ATCO VERIFY the following CVCS parameters: [Step 5]

ENSURE Letdown at minimum flow of 26 gpm on FIC-212. [Step 5.1]

ENSURE Charging Pumps CH-1A & CH-1B are RUNNING. [Step 5.2]

ATCO NOTIFY Work Week Manager of Pressurizer level instrument failure. [Step 6]

When Pressurizer level is normal, PROCEED to Event 3.

NRC Simulator Scenario 3 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 3 Page 11 of 34 Event

Description:

Steam Generator Steam Flow Transmitter Failure Time Position Applicants Actions or Behavior Booth Operator: When directed, EXECUTE Event 3.

- Steam Generator RC-2A Steam Flow Transmitter FT-907 fails high.

Indications Available:

Feedwater Digital Control System Alarm

+30 sec BOPO RESPOND to Annunciator Response Procedures.

INFORM CRS Steam Generator RC-2A Steam Flow Transmitter FT-907 BOPO failed high.

CRS DIRECT actions of ARP-DCS-FW, FT-907.

Examiner Note: The following steps are from ARP-DCS-FW, Feedwater Digital Control System.

BOPO PERFORM the following for Steam Flow Instrument FT-907 failure: [Step 1]

VERIFY that FORCED TO 1 ELEM and 1 ELEM AUTO is displayed on Feedwater Regulating System display for RC-2A PT-907. [Step 1.1]

TOUCH display with the BAD process. [Step 1.2]

DETERMINE BAD input NOT automatically bypassed. [Step 1.3]

TOUCH Bypass on verification faceplate to BYPASS BAD input.

[Step 1.3.1]

VERIFY point displays GOOD status. [Step 1.3.2]

ENSURE control SHIFT to 3 ELEMENT AUTO. [Step 1.3.3]

CRS DETERMINE Steam Generator level instruments NOT affected. [Step 2]

CRS DETERMINE BAD input bypassed MANUALLY. [Step 3]

BOPO MONITOR Steam Generator levels. [Step 4]

CRS VERIFY XC-105, Secondary Calorimetric, is valid. [Step 5]

CRS DETERMINE LT-903 or LT-906 NOT cause of alarm. [Step 6]

NRC Simulator Scenario 3 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 3 Page 12 of 34 Event

Description:

Steam Generator Steam Flow Transmitter Failure Time Position Applicants Actions or Behavior BOPO NOTIFY Work Week Manager of FT-907 malfunction. [Step 7]

When Steam Generator levels are normal, PROCEED to Event 4.

NRC Simulator Scenario 3 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 4 Page 13 of 34 Event

Description:

Charging Pump Trip Time Position Applicants Actions or Behavior Booth Operator: When directed, EXECUTE Event 4.

- Charging Pump CH-1A trip.

Indications Available:

CB-1,2,3/A2 - CHARGING PUMPS TRIP CB-1,2,3/A2 - CHARGING FLOW LO

+30 sec ATCO RESPOND to Annunciator Response Procedures.

Examiner Note: ATCO may Operate to Mitigate per OPD 4-09 and START CH-1B to avoid losing Letdown flow. Charging Pump CH-1B does not AUTO START until a level deviation exists.

ATCO INFORM CRS of Charging Pump CH-1A trip.

CRS REFER to ARP-CB-1,2,3/A2, Window A-6L - CHARGING FLOW LO.

Examiner Note: The following steps are from ARP-CB-1,2,3/A2, Window A-6L - CHARGING FLOW LO.

ATCO OBSERVE Charging Header flow LOW. [Step 1]

If Charging flow is lost, CLOSE TCV-202 and HCV-204 to ISOLATE ATCO Letdown. [Step 2]

DETERMINE TCV-202, Letdown to Regenerative Heat Exchanger Isolation Valve AUTO CLOSED or manually CLOSE.

Manually CLOSE HCV-204, Reactor Coolant to Letdown Heat Exchanger Isolation Valve.

NOTE Based on plant conditions, XC-105 and GARDEL may be invalid.

Booth Operator: When contacted about the status of CH-1A, REPORT a breaker overcurrent trip. Investigation of CH-1A: The pump looks normal locally. If Maintenance or Work Week Manager is contacted, estimated time to restore CH-1C is 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

NRC Simulator Scenario 3 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 4 Page 14 of 34 Event

Description:

Charging Pump Trip Time Position Applicants Actions or Behavior If required, ROTATE Charging Pumps per OI-CH-1, CVCS Normal ATCO Operation, Attachment 1, Startup of Charging and Letdown. [Step 5]

EVALUATE Technical Specification LCO 2.2, Chemical and Volume Control CRS System

LCO 2.2.4 - Charging Pumps - Operating

CONDITION LCO 2.2.4.(1) - Two Charging Pumps inoperable

ACTION LCO 2.2.4.(1) - RESTORE to at least two OPERABLE Charging Pumps within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

When Charging and Letdown flows are restored, PROCEED to Event 5.

NRC Simulator Scenario 3 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 5 Page 15 of 34 Event

Description:

Steam Generator Tube Leak Time Position Applicants Actions or Behavior Booth Operator: When directed, EXECUTE Event 5.

- Steam Generator RC-2B Tube Leak greater than 150 gpd.

Indications Available:

RM-057, Condenser Off Gas Radiation Monitor in alarm and trending up RM-054B, Steam Generator RC-2B Blowdown Radiation Monitor in alarm and trending up

+30 sec ATCO RESPOND to Radiation Monitor Alarms.

ATCO INFORM CRS of indications of the tube leak on Steam Generator RC-2B.

CRS REFER to AOP-22, Reactor Coolant Leak.

Examiner Note: The following steps are from AOP-22, Reactor Coolant Leak,Section I, Leak Rate Determination and Leak Isolation.

CRS DETERMINE Shutdown Cooling is NOT in operation. [Step 4.1]

Booth Operator: When contacted as Shift Chemist, WAIT 2 minutes and REPORT Steam Generator RC-2B has increased activity and RC-2A has normal activity.

DETERMINE CIAS is NOT present and DIRECT Shift Chemist to PERFORM CRS the following: [Step 4.2]

PERFORM rapid activity analysis of both Steam Generators. [Step 4.2.a]

SAMPLE both SGs per CH-SMP-SE-0015, Steam Generator Sampling -

Room 60. [Step 4.2.b]

CRS IMPLEMENT the Emergency Plan. [Step 4.3]

CREW MONITOR the Floating Steps. [Step 4.4]

ATCO DETERMINE Pressurizer level is NOT below programmed level. [Step 4.5]

NRC Simulator Scenario 3 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 5 Page 16 of 34 Event

Description:

Steam Generator Tube Leak Time Position Applicants Actions or Behavior ATCO DETERMINE RCS leakage rate per IC-17, RCS Manual Leak Rate and/or Calculation. [Step 4.6]

BOPO CRS DETERMINE RCS leak rate is NOT greater than 40 gpm. [Step 4.7]

Booth Operator: When contacted as Shift Chemist, WAIT 10 minutes, then REPORT initial Steam Generator RC-2B leak rate is greater than 150 gallon per day.

DIRECT Shift Chemist to verify primary to secondary leak rate < 1 gpd per CRS CH-AD-0007, Primary to Secondary Leak Rate Determination. [Step 4.8]

[CA] If primary to secondary leak rate is > 1 gpd, IMPLEMENT Attachment B, Primary to Secondary Leak Rate Actions.

Examiner Note: The following steps are from AOP-22, Reactor Coolant Leak, Attachment B, Primary to Secondary Leak Rate Actions.

CRS IMPLEMENT SO-G-105, Steam Generator Tube Leakage. [Step 1]

Booth Operator: When contacted, REPORT Work Week Manager will implement SO-G-105.

Continuously MONITOR count rate trends for radiation monitors RM-054A, ATCO RM-054B and RM-057 on ERF Computer System. [Step 2]

PERFORM the following to PLACE RM-064, Main Steam Line Radiation ATCO Monitor, in service at AI-33C: [Step 3]

PLACE Main Steam Line A/B Enable Switch for HCV-921 and HCV-922 in ON. [Step 3.a]

PLACE Main Steam Line A/B Mode Selector Switch in AUTO. [Step 3.b]

CRS PERFORM the following to IDENTIFY SG with tube leak: [Step 4]

CRS

DIRECT Shift Chemist to continue sampling. [Step 4.a]

CRS

REVIEW Steam Generator analysis for activity. [Step 4.b]

MONITOR RM-057 & RM-064, Steam Line Radiation Monitors and ATCO DETERMINE both radiation levels RISING. [Step 4.c]

NRC Simulator Scenario 3 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 5 Page 17 of 34 Event

Description:

Steam Generator Tube Leak Time Position Applicants Actions or Behavior

MONITOR RM-054A & RM-054B, SG Blowdown Radiation Monitors and ATCO DETERMINE RM-054B is RISING [Step 4.d]

BOPO

MONITOR SG levels and DETERMINE no apparent change. [Step 4.e]

Booth Operator: When contacted, EXECUTE remote functions to position HC-2509 / HC-2508 /

FW-268 / FW-266 as required.

Direct Equipment Operators to PERFORM the following to MINIMIZE spread CREW of contamination: [Step 5]

ENSURE HC-2509, SAMPLE DRAIN TO DRAIN HEADER, is OPEN at AI-107 in Room 60. [Step 5.a]

ENSURE HC-2508, SAMPLE DRAIN TO CONDENSER C.W. TUNNEL, is CLOSED at AI-107 in Room 60. [Step 5.b]

ENSURE FW-268, CONDENSATE DUMP VALVE LCV-1193 OUTLET ISOLATION VALVE, is CLOSED at Turbine Building Mezzanine.

[Step 5.c]

ENSURE FW-266, CONDENSATE DUMP VALVE LCV-1193 BYPASS VALVE, is CLOSED at Turbine Building Mezzanine. [Step 5.d]

CRS

DETERMINE SG RC-2A is least affected Steam Generator. [Step 5.e]

DETERMINE SG RC-2B is most affected Steam Generator and BOPO PERFORM the following: [Step 5.f]

PLACE YCV-1045B, RC-2B to FW-10 Isolation Valve in OVERRIDE.

PLACE YCV-1045B, RC-2B to FW-10 Isolation Valve to CLOSE.

CONSIDER stopping Turbine Building Sump Pumps VD-1A & VD-1B.

CRS

[Step 5.g]

CRS

CONSIDER isolating steam generator blowdown. [Step 5.h]

BOPO

PLACE RCV-978, 6th Stage Extraction Isolation Valve to STOP. [Step 5.i]

Booth Operator: When contacted, EXECUTE remote function to align Condenser Evacuation Discharge to Auxiliary Building Stack.

CONTACT Auxiliary Operator to ALIGN Condenser Evacuation CRS Discharge to Auxiliary Building stack per OI-CE-1, Condenser Evacuation System Normal Operation. [Step 5.j]

DIRECT Radiation Protection to develop a method for processing CRS contaminated Condensate. [Step 5.k]

NRC Simulator Scenario 3 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 5 Page 18 of 34 Event

Description:

Steam Generator Tube Leak Time Position Applicants Actions or Behavior CRS DETERMINE primary to secondary leakage greater than 5 gpd. [Step 6]

CRS DETERMINE primary to secondary leakage greater than 30 gpd. [Step 7]

DETERMINE primary to secondary leakage greater than 30 gpd independent CRS of Xe-133 concentration. [Step 8]

DETERMINE primary to secondary leakage greater than 75 gpd independent CRS of Xe-133 concentration. [Step 9]

DETERMINE primary to secondary leakage greater than 75 gpd with a rate CRS increase greater than 30 gpd in any 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period. [Step 10]

DETERMINE primary to secondary leakage greater than 75 gpd with a rate CRS increase greater than 30 gpd in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . [Step 11]

DETERMINE primary to secondary leak rate greater than 150 gpd (0.10 gpm)

CRS and PERFORM the following: [Step 12]

ISOLATE blowdown from SG RC-2B per HR-21, Blowdown Operation.

[Step 12.a]

COMMENCE a Plant Shutdown to MODE 4 per AOP-05, Emergency Shutdown. [Step 12.b]

CRS EVALUATE Technical Specification LCO 2.1, Reactor Coolant System.

LCO 2.1.4 - Reactor Coolant System Leakage Limits

CONDITION LCO 2.2.4.(1).(d) - RCS operational LEAKAGE shall be limited to 150 gallons per day primary to secondary LEAKAGE through anyone Steam Generator.

ACTION LCO 2.1.4.(3) - Primary to secondary LEAKAGE is not within limits, then be in MODE 3, Hot Shutdown, within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months AND be in MODE 4, Cold Shutdown within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

When Technical Specifications have been addressed, PROCEED to Event 6.

NRC Simulator Scenario 3 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 6 Page 19 of 34 Event

Description:

Commence Plant Shutdown Time Position Applicants Actions or Behavior Booth Operator: When directed, RESPOND to requests from Control Room.

Indications Available:

NONE Examiner Note: The following steps are from AOP-05, Emergency Shutdown.

NOTE TDB-III-23a and the Power Ascension/Power Reduction Strategy (PAPRs) provide guidance for the shutdown.

CRS CONTACT Reactor Engineer if additional guidance is required. [Step 4.1]

NOTE Operation of more than one Charging Pump will raise the rate of the power reduction.

Examiner Note: Unless directed, boration will occur from the Safety Injection Refueling Water Tank (SIRWT) when in AOP-05 to avoid time constraints.

If borating from SIRWT, COMMENCE boration by performing the following:

CRS

[Step 4.2]

DETERMINE Charging Pump, CH-1B RUNNING.

ATCO

[Step 4.2.a]

OPEN LCV-218-3, Charging Pump Suction SIRWT Isolation Valve.

ATCO

[Step 4.2.b]

ATCO

CLOSE LCV-218-2, VCT Outlet Valve. [Step 4.2.c]

CRS DETERMINE Boration alignment from CVCS NOT required. [Step 4.3]

CRS NOTIFY Energy Marketing of power reduction. [Step 4.4]

NOTE During the power reduction, maintain TC PER TDB Figure III.1, Tave Program.

MAINTAIN RCS Temperature Control via Turbine Load per HR-12, BOPO Secondary Heat Removal Operation: [Step 4.5]

NRC Simulator Scenario 3 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 6 Page 20 of 34 Event

Description:

Commence Plant Shutdown Time Position Applicants Actions or Behavior

MAINTAIN TCOLD 527°F to 547°F AND

MAINTAIN TCOLD +0°F to -1°F of program.

MAINTAIN Pressurizer Level via Charging and Letdown per IC-11, Inventory ATCO Control: [Step 4.6]

MAINTAIN Pressurizer Level 45% to 60% AND

MAINTAIN Pressurizer Level within 4% of program.

PERFORM the following to MAINTAIN VCT level between 55% and 85%:

ATCO

[Step 4.7]

As required, PLACE LCV-218-1, VCT Inlet Valve to RWTS. [Step 4.7.a]

When diversion is complete, PLACE LCV-218-1, VCT Inlet Valve to AUTO. [Step 4.7.b]

PERFORM the following to MAXIMIZE Pressurizer Heaters and Spray:

ATCO

[Step 4.8]

As required, PLACE Backup Heater Control Switches to ON. [Step 4.8.a]

ADJUST PC-103X or PC-103Y, Pressurizer Pressure Controller Setpoint Pushbutton to maintain pressure between 2080 psia and 2145 psia.[Step 4.8.b]

CAUTION Do not insert CEAs below power dependent insertion limit.

As required, ADJUST Regulating Group 4 to CONTROL ASI per OI-RR-1, ATCO Attachment 4, Axial Shape Index (ASI) Control. [Step 4.9]

Examiner Note: The following steps are from HR-12, Secondary Heat Removal Operation.

BOPO ENSURE Turbine Control is in MANUAL. [Step 1]

NOTE Output will be highlighted by a yellow box when selected.

NRC Simulator Scenario 3 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 6 Page 21 of 34 Event

Description:

Commence Plant Shutdown Time Position Applicants Actions or Behavior BOPO PUSH the OUT button to select OUTPUT. [Step 2]

NOTES

1. Depressing the single arrow will adjust turbine load by 0.1%. Depressing the double arrow will adjust turbine load by 0.5%.

2. Tc should be maintained within (+)0°F, (-)1°F of program per TDB-III.1, Tave Program.

PRESS single or double UP[] or DOWN[] arrow to maintain Turbine BOPO Load: [Step 3]

MAINTAIN TCOLD 527°F to 547°F.

MAINTAIN TCOLD +0°F to -1°F of program.

Examiner Note: Do not proceed to the next event during electrical plant realignment to 161KV.

When Reactor power is reduced 3% to 5%, PROCEED to Events 7 and 8.

NRC Simulator Scenario 3 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 7&8 Page 22 of 34 Event

Description:

Steam Generator Tube Rupture / Diesel Generator Start Failure Time Position Applicants Actions or Behavior Booth Operator: When directed, EXECUTE Events 7 and 8.

- Steam Generator RC-2B Tube Rupture @ 500 gpm on 10 minute ramp.

- Diesel Generator DG-01 start failure on SIAS.

Indications Available:

Pressurizer pressure and level lowering.

RECOGNIZE Pressurizer pressure and level lowering, upward trending

+2 min ATCO Radiation Monitors and MANUALLY TRIP Reactor.

CRS DIRECT performance of EOP-00, Standard Post Trip Actions.

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

ATCO VERIFY Reactivity Control: [Step 1]

VERIFY ALL of the following: [Step 1.a]

VERIFY no more than one Regulating or Shutdown CEA NOT inserted.

VERIFY Reactor Power is LOWERING.

VERIFY Startup Rate is NEGATIVE.

MONITOR plant for an uncontrolled RCS Cooldown. [Step 1.b]

CRS DETERMINE Reactivity Control criteria SATISFIED.

BOPO VERIFY Turbine Trip: [Step 2]

VERIFY HP & LP Stop and Intercept Valves CLOSED.

BOPO ENSURE all Generator Breakers are tripped: [Step 3]

DETERMINE Generator Output Breaker 3451-4 tripped.

DETERMINE Generator Output Breaker 3451-5 tripped.

DETERMINE Generator Field Breaker 41E/G1F tripped.

BOPO VERIFY 4160 V Safeguards Buses 1A3 and 1A4 are ENERGIZED. [Step 4]

NRC Simulator Scenario 3 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 7&8 Page 23 of 34 Event

Description:

Steam Generator Tube Rupture / Diesel Generator Start Failure Time Position Applicants Actions or Behavior CRS DETERMINE Maintenance of Vital Auxiliaries criteria SATISFIED.

Examiner Note: The following step (Verify Diesel Generators running) is not required until Reactor Coolant System Pressure is less than 1600 psia and PPLS has actuated.

VERIFY both Diesel Generators RUNNING on Safety Injection Actuation BOPO Signal. [Step 5]

[CA] DEPRESS DG-01 Emergency Start pushbutton and VERIFY DG-01 running at 900 RPM.

VERIFY 4160 V Non-Safeguards Buses 1A1 and 1A2 are ENERGIZED.

BOPO

[Step 6]

BOPO VERIFY 125 VDC Buses 1 and 2 are ENERGIZED. [Step 7]

BOPO VERIFY Instrument Air is AVAILABLE: [Step 8]

DETERMINE Instrument Air pressure 90 psig.

DETERMINE Instrument Air Compressor CA-1C RUNNING.

ATCO VERIFY Component Cooling Water System operation NORMAL: [Step 9]

DETERMINE at least one CCW pump RUNNING. [Step 9.a]

DETERMINE CCW Pump discharge pressure 60 psig. [Step 9.b]

DETERMINE HCV-438A/B/C/D, CCW to RCP Coolers OPEN. [Step 9.c]

DETERMINE at least one Raw Water Pump RUNNING. [Step 9.d]

CRS VERIFY RCS Inventory Control criteria satisfied: [Step 10]

DETERMINE PZR level NOT between 30% and 70% and NOT TRENDING to between 45% and 60%.

[CA] RESTORE Inventory Control by manually controlling Charging and Letdown. [Step 10.1.a]

DETERMINE RCS subcooling > 20°F.

CRS DETERMINE RCS Inventory Control criteria NOT SATISFIED.

NRC Simulator Scenario 3 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 7&8 Page 24 of 34 Event

Description:

Steam Generator Tube Rupture / Diesel Generator Start Failure Time Position Applicants Actions or Behavior CRS VERIFY RCS Pressure Control criteria satisfied: [Step 11]

ATCO

DETERMINE RCS pressure less than 1600 psia.

[CA] VERIFY RCS pressure < 2300 psia and PORV NOT open.

[Step 11.1]

[CA] When RCS pressure < 1350 psia, PERFORM the following:

[Step 11.2]

ATCO * [CA] STOP one RCP in each Loop.

[CA] DETERMINE RCS pressure < 1600 psia and VERIFY Engineered Safeguards ACTUATED. [Step 11.3]

[CA] DETERMINE PPLS relays 86A/PPLS / 86B/PPLS /

86A1/PPLS / 86B1/PPLS have TRIPPED.

[Step 11.3.a]

[CA] DETERMINE all PPLS relays have TRIPPED.

[Step 11.3.b]

[CA] DETERMINE VIAS relays 86A/VIAS / 86B1/VIAS /

86B/VIAS / 86A1/VIAS have TRIPPED. [Step 11.3.c]

[CA] DETERMINE SIAS relays 86A/SIAS / 86AX/SIAS /

86B1/SIAS / 86B1X/SIAS / 86B/SIAS / 86BX/SIAS /

86A1/SIAS / 86A1X/SIAS have TRIPPED.

[Step 11.3.d]

[CA] DETERMINE CIAS relays 86A/CIAS / 86B1/CIAS /

86B/CIAS / 86A1/CIAS have TRIPPED. [Step 11.e]

[CA] ENSURE required pumps RUNNING [Step 11.3.f]

DETERMINE HPSI Pumps SI-2A & SI-2B RUNNING.

DETERMINE LPSI Pumps SI-1A and SI-1B RUNNING.

DETERMINE Charging Pumps CH-1B RUNNING.

[CA] ENSURE acceptable SI flow per Attachment IC-13, SI Flow vs. Pressurizer Pressure. [Step 11.3.g]

[CA] ENSURE Emergency Boration in progress.

ATCO

[Step 11.3.h]

CRS DETERMINE RCS Pressure Control criteria NOT SATISFIED.

NRC Simulator Scenario 3 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 7&8 Page 25 of 34 Event

Description:

Steam Generator Tube Rupture / Diesel Generator Start Failure Time Position Applicants Actions or Behavior Examiner Note: The following steps are from RC-11, Emergency Boration Verification.

ATCO ENSURE the following valves are CLOSED: [Step 1]

FCV-269X, Demin Water Makeup Valve

FCV-269Y, Boric Acid Makeup Valve

HCV-264, CH-4A Recirc Valve

HCV-257, CH-4B Recirc Valve ATCO VERIFY all the following valves OPEN: [Step 2]

HCV-268, Boric Acid Pump Header to Charging Pumps Isolation Valve

HCV-265, CH-11A Gravity Feed Valve

HCV-258, CH-11B Gravity Feed Valve ATCO ENSURE all available Boric Acid Pumps RUNNING: [Step 3]

CH-4A, Boric Acid Pump

CH-4B, Boric Acid Pump ATCO ENSURE all available Charging Pumps RUNNING: [Step 4]

CH-1A, Charging Pump is tripped.

CH-1B, Charging Pump is RUNNING.

CH-1C, Charging Pump is OOS.

ATCO ENSURE the following valves are CLOSED: [Step 5]

LCV-218-2, VCT Outlet Valve

LCV-218-3, Charging Pump Suction SIRWT Isolation Valve

HCV-257, CH-4B Recirculation Valve

HCV-264, CH-4A Recirculation Valve ATCO DETERMINE Emergency Boration is in progress. [Step 6]

NRC Simulator Scenario 3 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 7&8 Page 26 of 34 Event

Description:

Steam Generator Tube Rupture / Diesel Generator Start Failure Time Position Applicants Actions or Behavior Examiner Note: The following steps continue from EOP-00, Standard Post Trip Actions.

CRS VERIFY Core Heat Removal criteria satisfied: [Step 12]

ATCO

DETERMINE RCP NPSH requirements met.

DETERMINE at least one RCP operating.

DETERMINE Core T 10°F.

Stop One Reactor Coolant Pump in Each Loop when Reactor Coolant System CRITICAL TASK Pressure is < 1350 psia, Prior to losing Reactor Coolant Pump Net Positive STATEMENT Suction Head.

CRITICAL DETERMINE Reactor Coolant System pressure < 1350 psia and PERFORM TASK ATCO the following:

ATCO

STOP one RCP in each Loop.

CRS DETERMINE Core Heat Removal criteria SATISFIED.

NOTE If Instrument Air to valves HCV-1105, HCV-1106, HCV-1107A/B and HCV-1108A/B is not available, throttling of these valves is not possible. Open or close operation of these valves is possible for a minimum of three cycles.

CRS VERIFY RCS Heat Removal criteria satisfied:

VERIFY Main Feedwater is restoring SG levels to 35% to 80% NR and 73%

BOPO to 94% WR. [Step 13]

DETERMINE FCV-1101 & FCV-1102 Feed Regulating Valves CLOSED.

[Step 13.a]

DETERMINE HCV-1105 & HCV-1106 Feed Regulating Bypass Valves ramped to between 40% & 45% OPEN. [Step 13.b]

BOPO

PLACE 43/FW switch in OFF. [Step 13.c]

ENSURE no more than one Main Feedwater Pump RUNNING.

[Step 13.d]

ENSURE no more than one Condensate Pump RUNNING. [Step 13.e]

NRC Simulator Scenario 3 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 7&8 Page 27 of 34 Event

Description:

Steam Generator Tube Rupture / Diesel Generator Start Failure Time Position Applicants Actions or Behavior

STOP running Heater Drain Pumps FW-5A, FW-5B, and/or FW-5C.

BOPO

[Step 13.f]

ENSURE SG Blowdown Isolation Valves CLOSED. [Step 13.g]

HCV-1387A & HCV-1387B

HCV-1388A & HCV-1388B VERIFY Steam Dump and Bypass Valves controlling both of the following:

BOPO

[Step 14]

DETERMINE RCS TCOLD between 525°F and 535°F.

DETERMINE Steam Generator pressure between 850 psia & 925 psia.

CRS DETERMINE RCS Heat Removal criteria SATISFIED.

CRS VERIFY Containment Isolation criteria satisfied:

ATCO VERIFY Normal Containment conditions exist: [Step 15]

DETERMINE no unexpected rise in Containment Sump level. [Step 15.a]

DETERMINE Containment Area Radiation Monitors NOT in alarm.

[Step 15.b]

DETERMINE Containment Ventilation Radiation Monitors NOT in alarm.

[Step 15.c]

DETERMINE RM-054B, SG Blowdown Radiation Monitor ALARMING.

ATCO

[Step 15.d]

[CA] MINIMIZE spread of contamination: [Step 15.d.1]

[CA] VERIFY RCV-978, 6th Stage Extraction Isolation Valve BOPO CLOSED. [Step 15.d.1.1)]

[CA] VERIFY all Blowdown Isolation Valves CLOSED.

[Step 15.d.1.2)]

[CA] HCV-1387A & HCV-1387B

[CA] HCV-1388A & HCV-1388B

DETERMINE RM-054B, SG Blowdown Radiation Monitor and RM-057, ATCO Condenser Off Gas Radiation Monitor TRENDING upward. [Step 15.e]

[CA] DETERMINE if Steam Generator Tube Rupture is in progress:

CRS

[Step 15.e.1]

NRC Simulator Scenario 3 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 7&8 Page 28 of 34 Event

Description:

Steam Generator Tube Rupture / Diesel Generator Start Failure Time Position Applicants Actions or Behavior

[CA] DIRECT Shift Chemist to perform rapid activity analysis of both SGs. [Step 15.e.1.1)]

[CA] DETERMINE SG RC-2B has an abnormal rise in level.

BOPO

[Step 15.e.1.2)]

ATCO

VERIFY Containment conditions: [Step 15.f]

DETERMINE Containment pressure < 3 psig.

DETERMINE Containment temperature < 120°F.

CRS DETERMINE Containment Integrity criteria NOT SATISFIED.

CRS DIAGNOSE event in progress: [Step 16]

DETERMINE Reactivity Control requirements met.

DETERMINE both DC buses energized.

DETERMINE at least one Vital 4160 V Bus energized.

DETERMINE at least one Non-Vital 4160 V Bus energized.

DETERMINE at least one RCP running.

DETERMINE adequate Feedwater flow to at least one SG.

VERIFY Pressurizer pressure > 1800 psia with high subcooled margin, normal SG pressure, and no indications of primary to secondary leakage.

If not, CONSIDER EOP-04, Steam Generator Tube Rupture.

NOTE Certain events (i.e., LOCA, SGTR, UHE and Loss of All Feedwater) do not require offsite power in order to adequately, mitigate the effects of the accident. For this reason, the LOCA, SGTR, UHE or Loss of All Feedwater procedure may be implemented even if a Loss of Offsite Power has also occurred.

DETERMINE all Safety Function Acceptance Criteria NOT SATISFIED.

DETERMINE single event in progress and TRANSITION to EOP-04, Steam Generator Tube Rupture.

Examiner Note: The following steps are from EOP-04, Steam Generator Tube Rupture.

CRS CONFIRM Standard Post Trip Actions have been performed. [Step 1]

NRC Simulator Scenario 3 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 7&8 Page 29 of 34 Event

Description:

Steam Generator Tube Rupture / Diesel Generator Start Failure Time Position Applicants Actions or Behavior CRS CONFIRM Steam Generator Tube Rupture Diagnosis: [Step 2]

VERIFY Safety Function Status Check Acceptance Criteria being satisfied. [Step 2.a]

VERIFY CIAS is present and SAMPLE both SGs. [Step 2.c]

CRS IMPLEMENT the Emergency Plan. [Step 3]

Time: __________

CREW MONITOR the Floating Steps. [Step 4]

DETERMINE RCS pressure 1600 psia and VERIFY Engineered CRS Safeguards are ACTUATED: [Step 5]

DETERMINE PPLS relays 86A/PPLS / 86B/PPLS / 86A1/PPLS /

86B1/PPLS have TRIPPED. [Step 5.a]

DETERMINE SIAS relays 86A/SIAS / 86AX/SIAS / 86B1/SIAS /

86B1X/SIAS / 86B/SIAS / 86BX/SIAS / 86A1/SIAS / 86A1X/SIAS have TRIPPED. [Step 5.b]

DETERMINE CIAS relays 86A/CIAS / 86B1/CIAS / 86B/CIAS /

86A1/CIAS relays TRIPPED. [Step 5.c]

DETERMINE VIAS relays 86A/VIAS / 86B1/VIAS / 86B/VIAS /

86A1/VIAS relays TRIPPED. [Step 5.d]

OPTIMIZE Safety Injection and Charging flow and PERFORM the following:

ATCO

[Step 6]

ENSURE required Safety Injection Pumps RUNNING: [Step 6.a]

DETERMINE HPSI Pumps SI-2A & SI-2B RUNNING.

DETERMINE LPSI Pumps SI-1A and SI-1B RUNNING.

DETERMINE Charging Pumps CH-1B RUNNING.

DETERMINE Emergency Boration already in progress per RC-11, ATCO Emergency Boration Verification. [Step 6.b]

ENSURE adequate SI flow per IC-13, Safety Injection Flow vs.

Pressurizer Pressure. [Step 6.c]

NRC Simulator Scenario 3 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 7&8 Page 30 of 34 Event

Description:

Steam Generator Tube Rupture / Diesel Generator Start Failure Time Position Applicants Actions or Behavior NOTE Main PZR Spray flow will be reduced with less than four-pump operation. Pressure should be controlled using Main and Auxiliary PZR Spray whenever the Plant is placed in a two-pump configuration.

ATCO VERIFY RCP operating parameters: [Step 7]

ENSURE at least one RCP stopped if TCOLD < 500°F. [Step 7.a]

DETERMINE one RCP stopped in each loop when RCS pressure 1350 psia following SIAS. [Step 7.b]

DETERMINE all RCPs STOPPED on low subcooling. [Step 7.c]

Time: _______

DETERMINE Condenser vacuum greater than 10.92 inches Hg absolute or CRS 19 inches Hg. [Step 8]

NOTE Reducing RCS TH to less than or equal to 510°F will maintain adequate RCP NPSH and RCS subcooling when RCS pressure is reduced below SG safety valve setpoint of 1000 psia.

CAUTION When TC is 178°F or greater, the maximum RCS cooldown rate is 100°F/hr.

When TC is less than 178°F, the maximum RCS cooldown rate is 50°F/hr.

COMMENCE a cooldown using both SGs to reduce RCS THOT to 510°F per BOPO Attachment HR-12, Secondary Heat Removal Operation. [Step 9]

COMMENCE a depressurization of RCS to less than 1000 psia per ATCO Attachment PC-11, Pressure Control. [Step 10]

Examiner Note: The following steps are from HR-12, Secondary Heat Removal Operation.

BOPO ENSURE Turbine Control is in MANUAL. [Step 1]

BOPO * [CA] DETERMINE Turbine NOT online and GO TO Step 4.

NRC Simulator Scenario 3 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 7&8 Page 31 of 34 Event

Description:

Steam Generator Tube Rupture / Diesel Generator Start Failure Time Position Applicants Actions or Behavior NOTES

1. In MANUAL, single arrows are 1% / double arrows are 5% change in OUTPUT value.

2. While Steam Dump and Bypass Control is in Temperature/Pressure Mode, the controllers PC0910 and TC0909_PI will alternate between controls, depending on the higher output signal. A red square outlining the controlling function signify which parameter is in control.

3. Steps 4 through 11 may be performed as needed, and in any order.

CAUTIONS When TC is 178°F or greater, the maximum RCS cooldown rate is 100°F/hr.

When TC is less than 178°F, the maximum RCS cooldown rate is 50°F/hr.

CRITICAL TASK Reduce and Maintain RCS THOT 510°F to Maintain SG Pressure Below Safety STATEMENT Valve Setpoint of 1000 psia.

CRITICAL DETERMINE Steam Dump and Bypass (SD&B) available and CONTROL TASK BOPO RCS temperature with a single SD&B Valve. [Step 4]

DEPRESS Valve Toggle to SELECT valve to be operated: [Step 4.a]

PCV-910 / TCV-909-1 / TCV-909-2 / TCV-909-3 / TCV-909-4

PLACE Controller for selected valve in MANUAL. [Step 4.b]

PUSH UP and DOWN arrows to ADJUST Controller Output. [Step 4.c]

When no longer required, PLACE Controller for selected valve in AUTO.

[Step 4.d]

Examiner Note: The following steps are from PC-11, Pressure Control. PC-12, RCS Pressure-Temperature Limits (graph), is maintained on a Control Room hardcopy.

CAUTION A charging header flow path must be maintained at all times.

MAINTAIN RCS pressure per the PC-12, RCS Pressure-Temperature Limits ATCO graph: [Step 1]

DETERMINE Steps N/A due to RCS pressure. [Step 1.a to 1.d]

NRC Simulator Scenario 3 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 7&8 Page 32 of 34 Event

Description:

Steam Generator Tube Rupture / Diesel Generator Start Failure Time Position Applicants Actions or Behavior CRITICAL

OPERATE the following to CONTROL Auxiliary Spray flow and TASK ATCO REDUCE RCS pressure to < 1000 psia: [Step 1.e]

HCV-240, PZR Auxiliary Spray Isolation Valve

HCV-249, PZR Auxiliary Spray Isolation Valve

HCV-238, Loop 1 Charging Isolation Valve

HCV-239, Loop 2 Charging Isolation Valve

If HPSI Stop and Throttle criteria is met, CONTROL Pressurizer level using Charging, Letdown, and/or HPSI flow per IC-11, Inventory Control.

[Step 1.f]

CONTROL RCS heat removal per HR-12, Secondary Heat Removal Operation. [Step 1.g]

MAINTAIN RCS Pressure per PC-12, RCS Pressure-Temperature Limits by ATCO performing ANY of the following: [Step 11]

CONTROL RCS Heat Removal per HR-12, Secondary Heat Removal Operation. [Step 11.a]

CONTROL Pressurizer Heaters and Spray per PC-11 Pressure Control.

[Step 11.b]

If HPSI Stop and Throttle criteria are met, CONTROL Charging, Letdown, and/or HPSI flow per IC-11, Inventory Control. [Step 11.c]

If feeding through Feed Ring, MAINTAIN SG levels 44% to 85% NR (77% to BOPO 94% WR) using Main Feedwater or FW-54. [Step 12]

FEED SGs using HR-15, Main Feed Pump Operation or HR-16, FW-54 Operation. [Step 12.a]

CONTROL feed flow per HR-11, Manual Feed Control. [Step 12.b]

PERFORM the following to PLACE RM-064, Main Steam Line Radiation ATCO Monitor in service at AI-33C. [Step 13]

PLACE Main Steam Line A/B Enable Switch for HCV-921 and HCV-922 in ON. [Step 13.a]

PLACE Main Steam Line A/B Mode Selector Switch in AUTO. [Step 13.b]

CRS DETERMINE Steam Generator RC-2B has the tube rupture. [Step 14]

BOPO PERFORM the following to MINIMIZE spread of contamination: [Step 15]

NRC Simulator Scenario 3 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 7&8 Page 33 of 34 Event

Description:

Steam Generator Tube Rupture / Diesel Generator Start Failure Time Position Applicants Actions or Behavior

CONTACT Auxiliary Operator to POSITION following valves: [Step 15.a]

ENSURE HC-2509, SAMPLE DRAIN TO DRAIN HEADER is OPEN at AI-107, Room 60. [Step 15.a]

ENSURE HC-2508, SAMPLE DRAIN TO CONDENSER C.W.

TUNNEL is CLOSED at AI-107, Room 60. [Step 15.b]

ENSURE FW-268, CONDENSATE DUMP VALVE LCV-1193 OUTLET ISOLATION VALVE, is CLOSED at Turbine Building Mezzanine. [Step 15.c]

ENSURE FW-266, CONDENSATE DUMP VALVE LCV-1193 BYPASS VALVE, is CLOSED at Turbine Building Mezzanine.

[Step 15.d]

BOPO When RCS THOT is 510°F, ISOLATE SG RC-2B. [Step 16]

ISOLATE SG RC-2B per HR-20, Isolate/Restore Steam Generator B.

[Step 16.a]

Examiner Note: The following steps are from HR-20, Isolate/Restore Steam Generator B.

NOTE RCS Heat Removal takes precedence over isolation of a S/G with a tube rupture.

CRITICAL TASK Isolate the Affected Steam Generator with a Tube Rupture to Minimize Spread STATEMENT of Contamination.

CRITICAL TASK BOPO PERFORM the following to isolate Steam Generator RC-2B: [Step 1]

ENSURE all the following valves are CLOSED: [Step 1.a]

CLOSE HCV-1042A, RC-2B MSIV.

VERIFY HCV-1042C, RC-2B MSIV Bypass Valve CLOSED.

VERIFY MS-292, Air Assisted Main Steam Safety Valve CLOSED.

CLOSE FCV-1102, RC-2B Feed Regulating Valve.

CLOSE HCV-1106, Feed Regulating Bypass Valve.

BOPO

CLOSE HCV-1385, RC-2B Feed Header Isolation Valve.

CLOSE HCV-1104, Feed Regulating Block Valve.

VERIFY HCV-1387A, Blowdown Isolation Valve CLOSED.

VERIFY HCV-1387B, Blowdown Isolation Valve CLOSED.

CLOSE HCV-1108A, AFW Isolation Valve.

CLOSE HCV-1108B, AFW Isolation Valve.

NRC Simulator Scenario 3 Outline Rev. 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 3 Event # 7&8 Page 34 of 34 Event

Description:

Steam Generator Tube Rupture / Diesel Generator Start Failure Time Position Applicants Actions or Behavior

CONTACT Auxiliary Operator to CLOSE MS-298, Steam Valves HCV-1041A & 1042A Packing Leakoff Line Isolation Valve in Room 81.

[Step 1.b]

If sampling is NOT in progress, CLOSE both Sample Valves:

[Step 1.c]

HCV-2506A, RC-2A Blowdown Sample Isolation Valve

HCV-2506B, RC-2B Blowdown Sample Isolation Valve BOPO

PERFORM the following to CLOSE YCV-1045B: [Step 1.d]

DETERMINE Isolation Valve YCV-1045B OVERRIDE SW in OVERRIDE. [Step 1.d.1)]

DETERMINE SG RC-2B STM TO FW-10 HDR A ISOLATION VALVE YCV-1045B in CLOSE. [Step 1.d.2)]

NOTE Air accumulators will maintain the valve in a closed position for 30 minutes after a loss of Instrument Air.

CONTACT Auxiliary Operator HANDJACK YCV-1045B, MAIN STEAM LINE "B" TO AUX FEEDWATER PUMP FW-10 SUPPLY VALVE to CLOSE in Room 81. [Step 1.d.3)]

CRS

RECORD time Steam Generator RC-2B ISOLATED. [Step 1.e]

Time: ________

VERIFY RC-2B is most affected SG per Attachment HR-18, Most Affected CRS Steam Generator Determination. [Step 2]

When Steam Generator RC-2B is isolated, TERMINATE the scenario.

NRC Simulator Scenario 3 Outline Rev. 6

Appendix D Scenario Outline Form ES-D-1 Facility: Fort Calhoun Station Scenario No.: 4 Op Test No.: Dec 2015 NRC Examiners: Operators:

Initial Conditions: MODE 2 at ~1% power - RCS Boron is 959 ppm (by sample).

Turnover: Continue in OP-2A, Plant Startup and OI-RR-1, Reactor Regulating System Operation to raise Reactor power to 7% power. When MODE 1 is entered, place Steam Dump and Bypass Valves in AUTO per OI-MS-1A, Main Steam System Operation.

Critical Tasks:

Manually Trip Reactor to meet Core Design Criteria of Lowering Reactor Power and Negative Startup Rate to Verify Reactivity Control Established During ATWS Event. (Event 5)

Isolate the Affected Steam Generator to Prevent Excess Plant Cooldown and Reactivity Additions Prior to Steam Generator Level = 0% Wide Range Level. (Event 7)

Event No. Malf. No. Event Type* Event Description 1 R (ATCO) Raise Power Using Control Rods to 7% per OP-2A, Plant Startup.

+20 min N (BOPO, CRS) Place Steam Dump and Bypass Valves in AUTO per OI-MS-1A.

2 C (BOPO, CRS) Raw Water Pump Discharge Line Leak Upstream of HCV-2879A in

+30 min TS (CRS) the Auxiliary Building.

3 I (BOPO, CRS) Inadvertent Channel B Auxiliary Feedwater Actuation Signal On

+45 min TS (CRS) Steam Generator RC-2A.

4 C (ATCO, CRS) Loss of Instrument Bus AI-40A.

+60 min TS (CRS) Loss of Letdown and Pressurizer Level Control.

5 M (ATCO, BOPO, Reactor Coolant Pump RC-3A Trip.

+60 min CRS) Automatic Reactor Trip Failure, Manual Reactor Trip Required.

6 C (BOPO) Instrument Air Compressor CA-1B and CA-1C Trip.

+65 min Bearing Cooling Water Pump AC-9B Trip.

7 M (ATCO, BOPO, Steam Line Break inside Containment on RC-2A @ 1% Severity on

+70 min CRS) 5 Minute Ramp.

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Technical Specifications Actual Target Quantitative Attributes 1 Malfunctions after EOP entry (1-2) 3 Abnormal events (2-4) 2 Major transients (1-2) 1 EOPs entered/requiring substantive actions (1-2) 0 EOP contingencies requiring substantive actions (0-2) 2 Critical tasks (2-3)

NRC Simulator Scenario 4 Outline Rev 6

Scenario Event Description NRC Scenario 4 SCENARIO

SUMMARY

NRC 4 The crew will assume the shift at 1% power and raise Power to ~7% using CEAs per OP-2A, Plant Startup, Attachment 4, Hot Standby, MODE 2 to Minimum Load, MODE 1 and OI-RR-1, Reactor Regulating System Normal Operation, Attachment 7, Manual Sequential Mode Checklist. When MODE 1 is entered, temperature control is placed in AUTO per OI-MS-1A, Main Steam System Operation, , Steam Dump and Bypass Manual Control Function.

The next event is a Raw Water Pump AC-10C discharge line leak in the Auxiliary Building upstream of HCV-2879A. The crew enters AOP-18, Loss of Raw Water, and must observe Raw Water System indications in order to determine the location of the leak. Once identified, the leak is isolated per AOP-18, Attachment C, Equipment Isolation, and Raw Water flow is restored. The SRO will refer to Technical Specification LCO 2.4(1) - Raw Water Header.

The next event is an inadvertent Channel B Auxiliary Feedwater Actuation Signal (AFAS) on Steam Generator RC-2A. The crew responds per ARP-AI-66B/A66B, Window 41 and verifies Auxiliary Feedwater Pumps FW-6 and FW-10 are running. Once it is determined the AFAS was inadvertent, AOP-23, Reset of Engineered Safeguards,Section IX, Reset of Inadvertent AFAS, is performed. The SRO will refer to Technical Specification LCO 2.15.1(1) - Automatic Initiation Steam Generator Water Level Logic Subsystem B.

When plant conditions are stable, a loss of Instrument Bus AI-40A occurs. The crew enters AOP-16, Loss of Instrument Bus Power,Section I, Loss of Instrument Bus Power, then Section II, Loss of Instrument Bus AI-40A. Actions include isolating Letdown, transferring Pressurizer Level Control, and operating Charging Pumps as required. Electrical Maintenance is notified and the Plant remains in this configuration through the end of the Scenario. The SRO will refer to Technical Specification LCO 2.15.2

- Reactor Protective System Logic and Trip Initiation and LCO 2.7(1) - 120 VAC Instrument Bus A.

The next event is a trip of Reactor Coolant Pump RC-3A. The crew should recognize failure of the Reactor Protection System Low Flow trips and manually trip the Reactor and enter EOP-00, Standard Post Trip Actions. When the Reactor is tripped, a 1% severity Steam Line Break inside Containment initiates on a 5 minute ramp. Due to the small size of this break, RCS pressure remains above the SIAS initiation setpoint of 1600 psia. The crew will transition to EOP-05, Uncontrolled Heat Extraction, and identify and isolate the affected Steam Generator RC-2A.

The event is complicated by a trip of the running and standby Instrument Air Compressors CA-1B and CA-1C and a trip of Bearing Water Cooling Pump AC-9B. The crew must restore a Bearing Cooling Water Pump and Instrument Air Compressor while in EOP-00. The scenario is terminated when Steam Generator RC-2A is isolated per HR-19, Isolate/Restore Steam Generator A while in EOP-05.

Risk Significance:

Failure of risk important system prior to trip: Loss of Raw Water System Header Loss of Instrument Bus

Risk significant core damage sequence: Automatic Reactor Trip Failure Steam Line Break Inside Containment

Risk significant operator actions: Isolate Raw Water East Header Manually Trip Reactor Restore Instrument Air Isolate Affected Steam Generator NRC Simulator Scenario 4 Outline Rev 6

Scenario Event Description NRC Scenario 4 BOOTH OPERATOR INSTRUCTIONS for SIMULATOR SETUP RESET to IC-122 and LOAD & EXECUTE NRC 4.sce for NRC Scenario 4.

Preset item - Event 5 - Reactor Fails to Trip Automatically, CB-4 Trip Button Works Type Item Value Condition Expert RPS02 Energized Scenario Event: Rx Fail to RPS01 Energized Trip, CB-4 works RPS03 Energized RPS04 Energized P6A_026_1 True P6B_028_1 True ANN-P6A_0026R1C_Fail Alarm Off ANN-P6A_0027R1C_Fail Alarm Off ANN-P6B_0026R5C_Fail Alarm Off ANN-P6B_0027R5C_Fail Alarm Off ANN-P6B_0025R5C_Fail Alarm Off ANN-P6A_0025R1C_Fail Alarm Off H_P6A_022A_1 True H_P6B_024A_1 True Event 2 - Raw Water leak in the Auxiliary Building Type Item Value Condition Malfunction RWS02B 25 When directed by examiner, trigger/activate this event.

Scenario Event: Raw Water Leak in Aux Building Event 3 - Inadvertent AFAS on RC-2A Type Item Value Condition Expert B_RC_2A_AFWS True When directed by examiner, trigger/activate this event.

Scenario Event:

Inadvertent AFAS Event 4 - Loss of Instrument Bus AI-40A Type Item Value Condition Malfunction EDA08 10 When directed by examiner, trigger/activate this event.

Scenario Event: Loss of AI-40A Event 5 - A Reactor Coolant Pump Trips Type Item Value Condition Malfunction BUS_1A1_5_BKR_TRIP True When directed by examiner, trigger/activate this event.

Scenario Event: A RCP Trip NRC Simulator Scenario 4 Outline Rev 6

Scenario Event Description NRC Scenario 4 Event 6 - Following RX Trip, Loss of Instrument Air and Bearing Cooling Water Type Item Value Condition Remote BCW_AC9B_BRKR Trip Event is triggered Malfunction BUS_1B3A_4A_2_BKR_Trip True automatically after reactor BUS_1B4B_4_BKR_TRIP True trip. Scenario Event: Loss of Inst Air and Bearing Water Event 7 - Main Steam Break Inside Containment Type Item Value Condition Malfunction SGN01A 1%, ramp = 300 sec Event is triggered automatically after reactor trip. Scenario Event:

Steam Line Break in Containment NRC Simulator Scenario 4 Outline Rev 6

Scenario Event Description NRC Scenario 4 Booth Operator: INITIALIZE to IC-122 and LOAD NRC 4.sce.

ENSURE all Simulator Annunciator Alarms are ACTIVE.

ENSURE all Control Board Tags are removed.

ENSURE Bearing Water Pump AC-9B running.

ENSURE Air Compressors CA-1B & CA-1C alignment: 1 in Standby, 1 running.

PLACE Steam Dump & Bypass Controllers in Manual.

ENSURE Lead Examiner has AFAS Keys 55 & 57 for Event 3.

ENSURE Lead Examiner has RPS Trip Unit Keys 1-12 for Event 4.

ENSURE Operator Aid Tags reflect current boron conditions.

ENSURE Reactivity Briefing Sheet printout provided with Turnover.

ENSURE Middle-of-Life Thumb Rule Sheet provided with Turnover.

ENSURE Steam Dump and Turbine Bypass System in MANUAL control.

ENSURE Control Room hard copy for OI-RR-1 is CLEAN.

ENSURE CEA Regulating Group 4 @ 72.

ENSURE procedures in progress provided to crew in Briefing Room:

- COPY of ReMA Data for Reactor Power Ascension.

- COPY of OP-2A, Plant Startup, Attachment 4, Hot Standby, MODE 2 to Minimum Load, MODE 1, INITIALED through Step 6.b.

- COPY of OI-RR-1, Reactor Regulating System Normal Operation, Attachment 7, Manual Sequential Mode Checklist.

- Copy of OI-MS-1A, Main Steam System Operation, Attachment 5, Steam Dump and Bypass Manual Control Function, INITIALED through Prerequisites and Steps 1.a & 2.a.

Control Room Annunciators in Alarm:

A9-B-1(U) - TURBINE DIFFERENTIAL EXPANSION A10-A-1(U) - MOTOR SUCT PUMP RUNNING OR NOT IN AUTO A10-B-6(L) - 43/FW TRANSFER SWITCH OFF-AUTO A11-A-4(U) - HEATER 5A HEATER HI-LO A11-A-4(L) - HEATER 5B HEATER HI-LO A11-B-3(U) - HEATER DRAIN TANK LEVEL HI-LO A20-D LOSS OF LOAD CHANNEL TRIP BYPASSED A20-E HIGH POWER RATE OF CHANGE TRIP ENABLED A21-B-1(U) - HC-909 INHIBIT A21-C-6(U) - HEATING STEAM PRESS LO AI-66B/A66B-Window 3 - FW-10 TURBINE DRIVEN FEEDWATER PUMP RUNNING NRC Simulator Scenario 4 Outline Rev 6

Scenario Event Description NRC Scenario 4 Procedure List Event 1: OP-2A, Plant Startup, Attachment 4, Hot Standby, MODE 2 to Minimum Load, MODE 1 Event 1: OI-RR-1, Reactor Regulating System Normal Operation, Attachment 7, Manual Sequential Mode Checklist Event 1: OI-MS-1A, Main Steam System Operation, Attachment 5, Steam Dump and Bypass Manual Control Function Event 2: AOP-18, Loss of Raw Water Event 3: ARP-AI-66B/A66B, Window 41, AFWS STEAM GEN RC-2A CHANNEL B ACTUATED Event 3: AOP-23, Reset of Engineered Safeguards,Section IX, Reset of Inadvertent AFAS Event 3: OI-AFW-2, Auxiliary Feedwater System Actuation and Bypass, Attachment 1, Bypass of the Auxiliary Feedwater Actuation Signal (AFAS) (Modes 1 or 2)

Event 3: OI-AFW-2, Auxiliary Feedwater System Bypass, Table 2, AFAS Logic Subsystem Channel Bypass Switch Alignment Event 4: AOP-16, Loss of Instrument Bus Power,Section I - Loss of Instrument Bus Power Event 4: AOP-16, Loss of Instrument Bus Power,Section II, Loss of Instrument Bus AI-40A Event 5: EOP-00, Standard Post Trip Actions Event 7: EOP-05, Uncontrolled Heat Extraction NRC Simulator Scenario 4 Outline Rev 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 1 Page 7 of 36 Event

Description:

Raise Reactor Power Time Position Applicants Actions or Behavior Booth Operator: When directed, RESPOND to requests from Control Room.

Examiner Note: This Scenario Section contains guidance for the following Operator actions:

1. Raising power per OP-2A.

2. Withdrawing Control Rods per OI-RR-1.

3. Control of Steam Dumps and Bypass per OI-MS-1A.

Examiner Note: The following steps are from OP-2A, Plant Startup, Attachment 4, Hot Standby, MODE 2 to Minimum Load, MODE 1, Step 6.

RAISE Reactor power to ~ 10% while performing the following: [Step 6]

DETERMINE Main Feedwater Pump FW-4B is RUNNING. [Step 6.a]

DETERMINE Auxiliary Feedwater SECURED. [Step 6.b]

MAINTAIN RCS temperature 527°F to 535°F using Steam Dump and Bypass Valves. [Step 6.c]

Prior to exceeding 15% power, VERIFY Secondary Chemistry parameters. [Step 6.d]

Prior to exceeding 15% power, VERIFY Condensate Pump Discharge Suspended Solids within specification. [Step 6.e]

PERFORM daily grab samples for Secondary activity or DECLARE RM-057 Radiation Monitor in service. [Step 6.f]

NOTE This step is performed to ensure that the DVM NI indication is greater than or equal to actual power.

When power is stable at approximately 10% (as indicated by highest of NI and T power), ADJUST RPS power per OI-NI-1. [Step 6.g]

OPEN MFW Isolation Valves HCV-1103 & HCV-1104. [Step 6.h]

Examiner Note: The following steps are from OI-RR-1, Reactor Regulating System Normal Operation, Attachment 7, Manual Sequential Mode Checklist, and is maintained as a Control Room hard copy.

ENSURE an out-of-scan CEA is NOT selected as Target Rod on CB-4.

ATCO

[Step 1]

VERIFY alarm REGULATING GROUP WITHDRAWAL PROHIBIT is clear.

ATCO

[Step 2]

NRC Simulator Scenario 4 Outline Rev 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 1 Page 8 of 36 Event

Description:

Raise Reactor Power Time Position Applicants Actions or Behavior PLACE Rod Control Mode Selector Switch in Manual Sequential (MS).

ATCO

[Step 3]

NOTE Continuous CEA motion shall be avoided whenever possible. CEA motion should be stopped at least every 33 inches (43 seconds of continuous CEA motion) to check position of CEAs in Group and Reactor response.

MOVE Manual Rod Control Switch to RAISE or LOWER as required.

ATCO

[Step 4]

DETERMINE appropriate Group Overlap during WITHDRAWAL is N/A.

ATCO

[Step 5]

When CEAs are at desired position, RELEASE Manual Rod Control Switch.

ATCO

[Step 6]

ATCO VERIFY all CEA motion has stopped. [Step 7]

ATCO If additional movement is required, GO TO Step 3. [Step 10]

When completed, PLACE Rod Control Mode Selector Switch in OFF.

ATCO

[Step 11]

Examiner Note: The following steps are from OI-MS-1A, Main Steam System Operation, Attachment 5, Steam Dump and Bypass Manual Control Function.

If operating all Steam Dump and Bypass Valves via the Pressure Controller BOPO (PC0910), PERFORM the following (SEC/MS/SD&B Control): [Step 1]

DETERMINE PC0910, STM DMP & BYP PRESS CONTROL, in MANUAL control. [Step 1.a]

As required, ADJUST output to Steam Dump and Bypass Valves.

[Step 1.b]

NRC Simulator Scenario 4 Outline Rev 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 1 Page 9 of 36 Event

Description:

Raise Reactor Power Time Position Applicants Actions or Behavior Booth Operator: When power has been raised approximately 3%, and prior to transitioning to the next event, CONTACT the Control room as the Shift Manager and direct placing Steam Dump and Turbine Bypass System (pressure and temperature control) in AUTO.

If desired to transfer back to AUTO at Output that has been selected, BOPO COMPLETE the following on Digital Control System: [Step 1.c]

PLACE PC0910 in LOCAL. [Step 1.c.1)]

ADJUST PC0910 SPT to approximately match PC0910 MEAS value.

[Step 1.c.2)]

PLACE PC0910 back in AUTO. [Step 1.c.3)]

If operating all Steam Dump and Bypass Valves via the Temperature BOPO Controller (TC0909_PI), PERFORM the following (SEC/MS/SD&B Control):

[Step 2]

DETERMINE TC0909_PI, STM DMP & BYP TEMP CONTROL, in MANUAL control. [Step 2.a]

As required, ADJUST output to Steam Dump and Bypass Valves.

[Step 2.b]

If desired to transfer back to AUTO at Output that has been selected,

+20 min BOPO COMPLETE the following on Digital Control System: [Step 2.c]

PLACE TC0909_PI in LOCAL. [Step 2.c.1)]

ADJUST TC0909_PI SPT to approximately match TC0909_PI MEAS value. [Step 2.c.2)]

PLACE TC0909_PI back in AUTO. [Step 2.c.3)]

When Reactor power is raised 3% to 5%, PROCEED to Event 2.

NRC Simulator Scenario 4 Outline Rev 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 2 Page 10 of 36 Event

Description:

Raw Water Pump Discharge Line Leak Time Position Applicants Actions or Behavior Booth Operator: When directed, EXECUTE Event 2.

- Raw Water Pump discharge line leak upstream of HCV-2879A.

Indications Available:

CB-1,2,3/A1 - RAW WATER SUPPLY HEADER FLOW LO CB-1,2,3/A1 - RAW WATER SUPPLY HEADER PRESS LO All Raw Water System 10 psig and 25 psig pressure indicating lights OUT

+30 sec ATCO RESPOND to Annunciator Response Procedures.

ATCO INFORM CRS of Raw Water System low pressure and low flow.

Examiner Note: ATCO may Operate to Mitigate per OPD 4-09 and START another Raw Water Pump.

CRS REFER to AOP-18, Loss of Raw Water.

Examiner Note: The following steps are from AOP-18, Loss of Raw Water.

ATCO DETERMINE Raw Water Pump AC-10C is RUNNING. [Step 4.1]

Booth Operator: If not already contacted, 1 minute after Control Room Receipt of alarms, REPORT as Auxiliary Building Operator that he observed water flowing out of room 18, and he is going in to investigate.

WAIT 30 seconds and REPORT Raw Water System leak in room 18, upstream of HCV-2879A/B on the header side of the system.

If Raw Water System rupture is indicated, DIRECT Operators to identify ATCO location of leak: [Step 4.2]

OBSERVE East RW Header Flow FIC-2890 OSCILLATING.

OBSERVE West RW Header Flow FIC-2891 OSCILLATING.

OBSERVE RW Pump(s) Current OSCILLATING.

OBSERVE RW System Pressure PIC-2892 OSCILLATING.

OBSERVE RW Pump Room Water Level LIC-2889/LC-2825 Level NORMAL.

NRC Simulator Scenario 4 Outline Rev 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 2 Page 11 of 36 Event

Description:

Raw Water Pump Discharge Line Leak Time Position Applicants Actions or Behavior ATCO DETERMINE Raw Water vault flooding is NOT occurring. [Step 4.3]

DETERMINE Raw Water leak in Auxiliary Building and PERFORM the ATCO following: [Step 4.4]

ENSURE only one Raw Water Pump RUNNING. [Step 4.4.a]

IMPLEMENT Attachment C, Equipment Isolation. [Step 4.4.b]

ATCO DETERMINE CCW temperature 110°F. [Step 4.5]

CRS IMPLEMENT the Emergency Plan. [Step 4.6]

Examiner Note: The following steps are from AOP-18, Attachment C, Equipment Isolation.

CRS If leak is on Raw Water System, GO TO Step 8. [Step 1]

NOTE The leak isolation Steps 8 through 15 may be performed in any logical order.

ATCO DETERMINE leak is NOT on any of the following: [Step 8]

WEST RW header

AC-12A, Raw Water Strainer

AC-1C, RW Heat Exchanger DETERMINE leak is on East Raw Water Header and PERFORM the ATCO following to ISOLATE Header: [Step 9]

PLACE AC-10D, Raw Water Pump, in PULL-TO-LOCK. [Step 9.a]

ATCO

CLOSE all Raw Water Header Isolation Valves: [Step 9.b]

CLOSE HCV-2876A.

CLOSE HCV-2876B.

CLOSE HCV-2894.

CLOSE HCV-2879A.

CLOSE HCV-2879B.

CLOSE HCV-2883A.

CLOSE HCV-2883B.

NRC Simulator Scenario 4 Outline Rev 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 2 Page 12 of 36 Event

Description:

Raw Water Pump Discharge Line Leak Time Position Applicants Actions or Behavior Booth Operator: When contacted, REPORT RW-145 is CLOSED.

When contacted, EXECUTE local actions and report handjacks applied to Raw Water System Valves as directed.

Locally CLOSE RW-145, RAW WATER STRAINER AC-12B ATCO BACKWASH VALVE HCV-2805B OUTLET ISOLATION VALVE in RW Vault. [Step 9.c]

DETERMINE leak is isolated and one Raw Water Pump RUNNING.

CRS

[Step 9.d]

Examiner Note: The following steps continue from AOP-18.

CRS DETERMINE Raw Water System restored to service. [Step 4.8]

CRS EVALUATE Technical Specification LCO 2.4, Containment Cooling

LCO 2.4.(1).a.iv. - Raw Water Header

CONDITION 2.4.(2).d - Raw Water Header inoperable.

ACTION 2.4.(2).d - RESTORE Raw Water Header within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months OR PLACE Reactor in HOT SHUTDOWN condition within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

When Raw Water System is realigned and Technical Specifications have been addressed, PROCEED to Event 3.

NRC Simulator Scenario 4 Outline Rev 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 3 Page 13 of 36 Event

Description:

Inadvertent Auxiliary Feedwater Actuation Signal Time Position Applicants Actions or Behavior Booth Operator: When directed, EXECUTE Event 3.

- Inadvertent Auxiliary Feedwater Actuation Signal.

Indications Available:

AI-66B/A66B - AFWS STEAM GEN RC-2A CHANNEL B ACTUATED AI-66B/A66B - FW-10 TURBINE DRIVEN FEEDWATER PUMP RUNNING (~30 seconds later)

AI-66B/A66B - FW-10 TURBINE OIL PUMP RUNNING (~30 seconds later)

+30 sec BOPO RESPOND to Annunciator Response Procedures.

BOPO INFORM CRS of Auxiliary Feedwater Actuation Signal initiation.

Examiner Note: BOPO may Operate to Mitigate per OPD 4-09 and CLOSE HCV-1107A and HCV-1107B to stop FW-10, Turbine Driven Auxiliary Feedwater Pump.

REFER to ARP-AI-66B/A66B, Window 41, AFWS STEAM GEN RC-2A CRS CHANNEL B ACTUATED.

Examiner Note: The following steps are from ARP-AI-66B/A66B, Window 41 - AFWS STEAM GEN RC-2A CHANNEL B ACTUATED.

CHECK A/B/LI-911, Steam Generator RC-2A Level at AI-66A and AI-66B.

BOPO

[Step 1]

DETERMINE SG level LI-911A at Panel AI-66A DEENERGIZED.

DETERMINE SG level LI-911B at Panel AI-66B NORMAL.

Booth Operator: When contacted, REPORT LI-911D, RC-2A level at AI-179 is ~ 64% and LI-911C, RC-2A pressure is ~ 884 psia (or as indicated).

BOPO DISPATCH Operator to check C/D/LI-911, RC-2A Level at AI-179. [Step 2]

BOPO DETERMINE Steam Generator Wide Range level is > 32%. [Step 3]

DETERMINE AFAS initiation is inadvertent and IMPLEMENTS AOP-23, CRS Reset of Engineered Safeguards,Section IX, Reset of Inadvertent AFAS.

[Step 4]

NRC Simulator Scenario 4 Outline Rev 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 3 Page 14 of 36 Event

Description:

Inadvertent Auxiliary Feedwater Actuation Signal Time Position Applicants Actions or Behavior CRS REFER to Technical Specification LCOs 2.14 and 2.15. [Step 5]

EVALUATE Technical Specification LCO 2.15.1, Instrumentation and Control CRS Systems

LCO 2.15.1.(3) - Automatic Initiation Steam Generator Water Level Logic Subsystem B (Table 2.3 / Function 5).

CONDITION 2.15.1.(3) - Logic Subsystem B inoperable

ACTION 2.15.1.(3) - RESTORE inoperable channel within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months OR PLACE Reactor in HOT SHUTDOWN condition within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Examiner Note: The following steps are from AOP-23, Reset of Engineered Safeguards,Section IX, Reset of Inadvertent AFAS.

CRS DETERMINE the AFAS is inadvertent. [Step 4.1]

CRS REFER to the following Technical Specifications: [Step 4.2]

LCO 2.5, Steam and Feedwater Systems

LCO 2.15, Instrumentation and Control Systems Examiner Note: Entry into Technical Specification LCO 2.5.(1).d is required until FW-10, TDAFW Pump is reset and returned to AUTO at the end of this event.

EVALUATE Technical Specification LCO 2.5, Steam and Feedwater CRS Systems

LCO 2.5.(1) - Two AFW Trains OPERABLE

CONDITION 2.5.(1).d - Both AFW Trains inoperable

ACTION 2.5.(1).d - RESTORE one train to OPERABLE status immediately.

BOPO ENSURE both of the following valves in AUTO: [Step 4.3]

DETERMINE FCV-1368, FW-6 Recirc Valve in AUTO.

DETERMINE FCV-1369, FW-10 Recirc Valve in AUTO.

NRC Simulator Scenario 4 Outline Rev 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 3 Page 15 of 36 Event

Description:

Inadvertent Auxiliary Feedwater Actuation Signal Time Position Applicants Actions or Behavior PLACE control switches for the following AFW Isolation Valves in CLOSE:

BOPO

[Step 4.4]

PLACE HCV-1107A in CLOSE.

PLACE HCV-1107B in CLOSE.

PLACE HCV-1108A in CLOSE.

PLACE HCV-1108B in CLOSE.

BYPASS affected logic subsystem per OI-AFW-2, Auxiliary Feedwater CRS System Actuation and Bypass. [Step 4.5]

Examiner Note: The following steps are from OI-AFW-2, Auxiliary Feedwater System Actuation and Bypass, Attachment 1, Bypass of the Auxiliary Feedwater Actuation Signal (AFAS) (Modes 1 or 2).

BOPO DETERMINE AFAS is aligned for automatic initiation. [Step 2]

BOPO DETERMINE plant is in Mode 1. [Step 3]

DETERMINE if an Instrument Channel or a Logic Subsystem Channel is to CRS be bypassed. [Step 1]

DETERMINE an Instrument Channel will NOT be bypassed. [Step 1.a]

DETERMINE a Logic Subsystem Channel will be bypassed and GO TO Step 3. [Step 1.b]

If a Logic Subsystem Channel of AFAS is to be bypassed, COMPLETE the CRS following: [Step 3]

SM/CRS

LOG entry into Technical Specification 2.15.1(3), 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months LCO.

NRC Simulator Scenario 4 Outline Rev 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 3 Page 16 of 36 Event

Description:

Inadvertent Auxiliary Feedwater Actuation Signal Time Position Applicants Actions or Behavior NOTE The following alarms are expected depending on the Logic Subsystem Channel that is bypassed:

AFWS RC-2A CH A MATRIX TS-A/RC-2A/AFWS TEST SWITCH OFF NORM (AI-66A, Window 24)

AFWS RC-2B CH A MATRIX TS-A/RC-2B/AFWS TEST SWITCH OFF NORM (AI-66A, Window 25)

AFWS OVERRIDE SWITCH A/OR-RC-2A/AFWS OFF NORMAL (AI-66A, Window 29)

AFWS OVERRIDE SWITCH A/OR-RC-2B/AFWS OFF NORMAL (AI-66A, Window 30)

HCV-1107A & B AFWS OVERRIDE SWITCH CH A OR B OFF NORM (AI-66A, Window 35)

AFWS RC-2A CH B MATRIX TS-B/RC-2A AFWS TEST SWITCH OFF NORM (AI-66B, Window 21)

AFWS RC-2B CH B MATRIX TS-B/RC-2B/AFWS TEST SWITCH OFF NORM (AI-66B, Window 22)

AFWS OVERRIDE SWITCH B/OR-RC-2A/AFWS OFF NORMAL (AI-66B, Window 26)

AFWS OVERRIDE SWITCH B/OR-RC-2B/AFWS OFF NORMAL (AI-66B, Window 27)

HCV-1108A & B AFWS OVERRIDE SWITCH CHA OR B OFF NORMAL (AI-66A, Window 32)

BYPASS selected Logic Subsystem using Table 2, AFAS Logic Subsystem BOPO Bypass Switch Alignment, and RECORD as left information in appropriate slots. [Step 3.b]

Examiner Note: The following steps are from OI-AFW-2, Table 2, AFAS Logic Subsystem Channel Bypass Switch Alignment.

Table 2 - AFAS Logic Subsystem Channel Bypass Switch Alignment As-Left Switch Bypassing Channel Panel No. Switch Position Position RC-2A Channel B AI-66B S/G RC-2A Chan. B Auto Sig Bypass (Amber lamps S/G RC- Override Relay Test Sw 2A Chan B/B1)

S/G RC-2A Chan. B Auto Sig Override Override Sw AFW Pumps FW-6/FW-10 Chan. B AFW Auto Sig B/OR -1107 Override S/G Feed Valves AFWS Examiner Note: Acting as Shift Manager, PROVIDE Keys #55 and #57 when requested.

BOPO PERFORM the following at Panel AI-66B for RC-2A Channel B:

INSERT key #57 and PLACE S/G RC-2A Channel B Auto Signal Override Relay Test Switch in BYPASS.

NRC Simulator Scenario 4 Outline Rev 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 3 Page 17 of 36 Event

Description:

Inadvertent Auxiliary Feedwater Actuation Signal Time Position Applicants Actions or Behavior

INSERT key #55 and PLACE S/G RC-2A Channel B Auto Signal Override Switch AFW Pumps FW-6/FW-10 in OVERRIDE.

PLACE Channel B AFW Auto Signal Override S/G Feed Valves to B/OR

-1107 AFWS position.

Examiner Note: The following steps continue from AOP-23,Section IX, Reset of Inadvertent AFAS.

BOPO PERFORM the following to STOP all AFW Pumps: [Step 4.6]

CLOSE YCV-1045, FW-10 Steam Inlet Valve. [Step 4.6.a]

PLACE both Override Switches in OVERRIDE: [Step 4.6.b]

ISOLATION VALVE YCV-1045A OVERRIDE SW.

ISOLATION VALVE YCV-1045B OVERRIDE SW.

CLOSE both FW-10 Steam Supply Valves: [Step 4.6.c]

YCV-1045A, RC-2A to FW-10 Isolation Valve.

YCV-1045B, RC-2B to FW-10 Isolation Valve.

ENSURE FIC-1369, AUX FW PUMP FW-10 SUCTION FLOW drops to zero. [Step 4.6.d]

STOP FW-6, Electric AFW Pump, and PLACE HC-1367, FW-6 Control Switch, in PULL-TO-LOCK. [Step 4.6.e]

ENSURE FIC-1368, AUX FW PUMP FW-6 SUCTION FLOW drops to zero. [Step 4.6.f]

PERFORM the following to return the AFW System to automatic operation:

BOPO

[Step 4.7]

PLACE Control Switches for AFW Isolation Valves in RESET:

[Step 4.7.a]

PLACE HCV-1107A in RESET.

PLACE HCV-1107B in RESET.

PLACE HCV-1108A in RESET.

PLACE HCV-1108B in RESET.

PLACE Control Switches for AFW Isolation Valves in AUTO: [Step 4.7.b]

NRC Simulator Scenario 4 Outline Rev 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 3 Page 18 of 36 Event

Description:

Inadvertent Auxiliary Feedwater Actuation Signal Time Position Applicants Actions or Behavior

PLACE HCV-1107A in AUTO.

PLACE HCV-1107B in AUTO.

PLACE HCV-1108A in AUTO.

PLACE HCV-1108B in AUTO.

PLACE Control Switch for YCV-1045, FW-10 Steam Inlet Valve, in RESET. [Step 4.7.c]

PLACE Control Switch for YCV-1045, FW-10 Steam Inlet Valve, in AUTO. [Step 4.7.d]

PLACE both Override Switches in NORMAL. [Step 4.7.e]

ISOLATION VALVE YCV-1045A OVERRIDE SW

ISOLATION VALVE YCV-1045B OVERRIDE SW

PLACE HC-1367, FW-6 Control Switch, in AFTER-STOP. [Step 4.7.f]

Booth Operator: When contacted, EXECUTE remote functions to RESET FW-10 and Trip Latch Clamp is finger tight.

CONTACT Auxiliary Operator ENSURE FW-64-RL, AUX FEED PUMP BOPO FW-10 MANUAL TRIP LATCH RESET LEVER is latched: [Step 4.8]

VERIFY Reset Lever is seated.

ENSURE FW-64-C, AUX FEED PUMP FW-10 MANUAL TRIP LATCH CLAMP is installed finger tight.

CRS EXIT Technical Specification LCO 2.5, Steam and Feedwater. [Step 4.9]

When AFAS has been RESET, PROCEED to Event 4.

NRC Simulator Scenario 4 Outline Rev 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 4 Page 19 of 36 Event

Description:

Loss of Instrument Bus Time Position Applicants Actions or Behavior Booth Operator: When directed, EXECUTE Event 4.

- Loss of Instrument Bus AI-40A.

Indications Available:

CB-20/A15 - INVERTER A TROUBLE CB-20/A15 - INSTRUMENT BUS A LOW VOLTAGE/GROUND (~10 seconds later)

Multiple Loss of Instrument Bus alarms

+30 sec BOPO RESPOND to Annunciator Response Procedures.

CREW INFORM CRS of Loss of Instrument Bus AI-40A.

Booth Operator: When contacted, REPORT Inverter A Output Breaker is TRIPPED.

REFER to AOP-16, Loss of Instrument Bus Power,Section I, Loss of CRS Instrument Bus Power.

Examiner Note: The following steps are from AOP-16, Loss of Instrument Bus Power,Section I, Loss of Instrument Power.

CRS DETERMINE a Reactor Trip has NOT occurred: [Step 4.1]

CRS DETERMINE appropriate AOP-16 Section: [Step 4.2]

OBSERVE an INVERTER A TROUBLE alarm.

OBSERVE an INSTRUMENT BUS A LOW VOLTAGE/GROUND alarm.

CRS GO TO AOP-16,Section II, Loss of Instrument Bus AI-40A.

Examiner Note: The following steps are from AOP-16, Loss of Instrument Bus Power,Section II, Loss of Instrument Bus AI-40A.

CRS VERIFY Loss of Instrument Bus AI-40A by the following: [Step 4.1]

INVERTER A TROUBLE alarm.

INSTRUMENT BUS A LOW VOLTAGE/GROUND alarm.

NRC Simulator Scenario 4 Outline Rev 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 4 Page 20 of 36 Event

Description:

Loss of Instrument Bus Time Position Applicants Actions or Behavior NOTE

1. Upon loss of Instrument Bus A, ALL of the following instrumentation or equipment associated with the Reactivity Control Safety Function is affected as follows:

All RPS Channel A is in trip

Channel A "VARIABLE OVER POWER TRIP POWER MARGIN A/JI-007" meter is inoperable

Channel A Wide Range Log Power Meter and Rate Meter are inoperable

The Diverse Scram System is in half-trip

Three RPS Logic Matrix channels are inoperable

2. Loss of more than one RPS Logic Matrix channel requires entry into T.S. 2.15.2.

3. If the associated clutch power supply is selected to Instrument Bus A then two RPS Trip Initiation Logic channels (AB, AC, AD) are inoperable.

DETERMINE clutch power supply selected to AI-40A and VERIFY clutch ATCO power supply is DEENERGIZED: [Step 4.2]

OBSERVE AI-3-PS1 output current is 0.

OBSERVE AI-3-PS3 output current is 0.

OBSERVE AI-3-PS1 Indicating lights are out.

OBSERVE AI-3-PS3 Indicating lights are out.

OBSERVE clutch power supply breaker in half trip position.

Examiner Note: Acting as Shift Manager, PROVIDE Trip Unit Keys #1 to #12 when requested.

ATCO INSERT keys and BYPASS all RPS Channel A Bistable Trip Units. [Step 4.3]

CRS COMPLY with Technical Specification 2.15.2(5). [Step 4.4]

EVALUATE Technical Specification LCO 2.15, Instrumentation and Control CRS Systems

LCO 2.15.2 - Reactor Protective System Logic and Trip Initiation

CONDITION 2.15.2.(2) - One RPS Trip Initiation Logic channel inoperable.

ACTION 2.15.2.(2) - Deenergize the affected clutch power supply within one hour (in 1/2 trip).

ACTION 2.15.2.(5) - With the required actions of (2) not met, be in HOT SHUTDOWN and verify no more than one CEA is capable of being withdrawn within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

NRC Simulator Scenario 4 Outline Rev 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 4 Page 21 of 36 Event

Description:

Loss of Instrument Bus Time Position Applicants Actions or Behavior NOTE Upon loss of Instrument Bus A, ALL of the following instrumentation or equipment associated with the Vital Auxiliaries Safety Function are inoperable:

"WEST RW SUPPLY HEADER FLOW FIC-2891" indicator

"CC HT EXCH AC-1A RW OUTLET TEMP TIC-2885"

"CNTMT CLG COIL VA-1A OUTLT ISOL VLV CNTRLR HCV-400C"

"CNTMT CLG COIL VA-1B OUTLT ISOL VLV CNTRLR HCV-401C"

"CNTMT CLG COIL VA-8A OUTLT ISOL VLV CNTRLR HCV-402C"

"CNTMT CLG COIL VA-8B OUTLT ISOL VLV CNTRLR HCV-403C" ATCO ENSURE CCW System operation satisfactory: [Step 4.5]

DETERMINE one CCW Pump RUNNING.

DETERMINE CCW pressure 60 psig.

ATCO DETERMINE one Raw Water Pump RUNNING. [Step 4.6]

BOPO DETERMINE Instrument Air pressure 90 psig. [Step 4.7]

NOTE Upon loss of Instrument Bus A, ALL of the following instrumentation or equipment associated with the RCS Inventory Control Safety Function is affected as follows:

Letdown is isolated

Pressurizer Level Channel X is inoperable

CH-1A, Charging Pump, is inoperable

Charging Pump Backup Auto starts are disabled MAINTAIN Pressurizer level between 30% and 70% and TRENDING to ATCO between 45% percent by operating Charging Pumps CH-1B and/or CH-1C per IC-11, Inventory Control. [Step 4.8]

ATCO CLOSE TCV-202, Letdown Isolation Valve. [Step 4.9]

PLACE HC-101, Pressurizer Level Channel Selector Switch, in CHAN Y ATCO position. [Step 4.10]

NRC Simulator Scenario 4 Outline Rev 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 4 Page 22 of 36 Event

Description:

Loss of Instrument Bus Time Position Applicants Actions or Behavior NOTE Upon loss of Instrument Bus A, ALL of the following instrumentation or equipment associated with the RCS Pressure Control Safety Function is affected as follows:

RCS Pressure Control Channel X is inoperable

"PRESSURIZER PRESSURE A/PIA-102X AND A/PIA-102Y" indicators are inoperable

PZR Backup Heaters are on

PZR Heater Cutout is inoperable PLACE HC-103, Pressurizer Pressure Channel Selector Switch in CHAN Y ATCO position. [Step 4.11]

Manually CONTROL Pressurizer Heaters per PC-11, Pressure Control.

ATCO

[Step 4.12]

MAINTAIN RCS pressure per PC-12, RCS Pressure-Temperature Limits.

ATCO

[Step 4.13]

NOTE

1. Only one additional channel trip is needed to actuate the PORVs, even if the channel in trip is bypassed.

2. When RCS Heatup or Cooldown is in progress, the PORVs are the primary means of Low Temperature Overpressure Protection.

3. Closing the PORV block valves requires entry into Tech Spec 2.1.6.

CRS CONSIDER closing PORV Block Valves HCV-150 and HCV-151. [Step 4.14]

NOTE Upon loss of Instrument Bus A, ALL of the following instrumentation or equipment associated with the Core Heat Removal Safety Function are inoperable:

"SUBCOOLED MARGIN MONITOR A-168"

"RC LOOP TEMPERATURES LOOP 1A "T-COLD" A/TI-112C"

"RC LOOP TEMPERATURES LOOP 1 "T-HOT" A/TI-112H"

"RC LOOP TEMPERATURES LOOP 2A "T-COLD" A/TI-122C"

"RC LOOP TEMPERATURES LOOP 2 "T-HOT" A/TI-122H"

"SHTDN HT EXCH AC-4A OUTLET VALVE CNTRLR HCV-484" ATCO DETERMINE all RCPs are RUNNING. [Step 4.15]

NRC Simulator Scenario 4 Outline Rev 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 4 Page 23 of 36 Event

Description:

Loss of Instrument Bus Time Position Applicants Actions or Behavior NOTE Upon loss of Instrument Bus A, ALL of the following instrumentation or equipment associated with the RCS Heat Removal Safety Function is inoperable:

"EMGY FW STOR TNK LEVEL LIA-1183"

"STEAM GENERATOR PRESSURE S/G RC-2A A/PIC-902"

"STEAM GENERATOR PRESSURE S/G RC-2B A/PIC-905"

"AUX FW PUMP FW-6 SUCTION FLOW FIC-1368" BOPO DETERMIN Steam Generator NR levels steady at ~63%. [Step 4.16]

NOTE

Upon loss of Instrument Bus A, RM-091A, which is associated with the Containment Integrity Safety Function is inoperable.

ATCO PERFORM the following to CONFIRM Containment Integrity: [Step 4.17]

DETERMINE no unexpected rise in Containment Sump level.

[Step 4.17.a]

DETERMINE no Containment Area Radiation Monitor alarms.

[Step 4.17.b]

DETERMINE Radiation Monitors RM-051 / RM-052 / RM-062 NOT in alarm. [Step 4.17.c]

DETERMINE SG Blowdown or Condenser off Gas Radiation Monitors RM-054A / RM-054B / RM-057 NOT in alarm. [Step 4.17.d]

DETERMINE Containment conditions NORMAL. [Step 4.17.e]

DETERMINE Containment pressure < 3 psig.

DETERMINE Containment temperature <120°F.

ATCO PLACE the following switches in TEST: [Step 4.18]

HC-344/TEST, CNTMT SPRAY VLV HCV-344 TEST SWITCH

HC-345/TEST, CNTMT SPRAY VLV HCV-345 TEST SWITCH NRC Simulator Scenario 4 Outline Rev 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 4 Page 24 of 36 Event

Description:

Loss of Instrument Bus Time Position Applicants Actions or Behavior NOTE Upon loss of Instrument Bus A, ALL of the following instrumentation or equipment associated with the Engineered Safety Features Systems is affected as follows:

Safety Injection Tanks 6A and 6C level and pressure indicators are inoperable

OPLS is in half-trip

Sequencer S2-2 is inoperable

PPLS is in a two-out-of-three logic mode

SGLS is in a two-out-of-three logic mode CRS REFER to all the following Technical Specifications: [Step 4.19]

2.1.6, Pressurizer and Steam System Safety Valves

2.2, Chemical and Volume Control System

2.3, Emergency Core Cooling System

2.5, Steam and Feedwater Systems

2.7, Electrical Systems

2.15, Instrumentation and Control Systems

2.21, Post-Accident Monitoring Instrumentation CRS EVALUATE Technical Specification LCO 2.7, Electrical Systems

LCO 2.7.(1).h - 120 VAC Instrument Bus A (Panel AI-40A).

CONDITION 2.7.(2).h - 120 VAC Instrument Bus A (Panel AI-40A) inoperable

ACTION 2.7.(2).h - May remain inoperable for 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months provided RPS and ESF instrument channels supplied by the remaining 3 buses are all OPERABLE.

REFER to Electrical Load Distribution Listing Manual for a list of components CREW powered from AI-40A. [Step 4.20]

Examiner Note: Instrument Bus IA-40A will remain deenergized for duration of scenario.

NRC Simulator Scenario 4 Outline Rev 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 4 Page 25 of 36 Event

Description:

Loss of Instrument Bus Time Position Applicants Actions or Behavior Booth Operator: When contacted, REPORT Electrical Maintenance investigating issue with Inverter A.

When cause of power loss has been determined and corrected, RESTORE

+15 min CRS AI-40A to normal per Attachment 1 or 12 of OI-EE-4, 120 Volt AC System Normal Operation. [Step 4.21]

When Technical Specifications have been addressed, PROCEED to Events 5, 6, and 7.

NRC Simulator Scenario 4 Outline Rev 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 5, 6, & 7 Page 26 of 36 Event

Description:

RCP Trip / Automatic Reactor Trip Failure / Instrument Air Compressors Trip / Bearing Cooling Water Pump Trip / Steam Line Break inside Containment Time Position Applicants Actions or Behavior Booth Operator: When directed, EXECUTE Events 5, 6, and 7.

- Reactor Coolant Pump RC-3A trip.

- Instrument Air Compressors CA-1B and CA-1C trip.

- Bearing Cooling Water Pump AC-9B trip.

- Steam Line Break inside Containment on RC-2A @ 1% severity and 5 minute ramp.

Indications Available:

CB-1,2,3,4/A6 - REACTOR COOLANT PUMP RC-3A BREAKER O/L OR TRIP Low Flow Trip Unit lights lit on all RPS Channels B/C/D.

ERF Computer System alarms for low RCS flow

+30 sec ATCO RECOGNIZE RPS Low Flow lights lit and MANUALLY trip Reactor.

CRS DIRECT performance of EOP-00, Standard Post Trip Actions.

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

ATCO VERIFY Reactivity Control: [Step 1]

VERIFY ALL of the following: [Step 1.a]

DETERMINE more than one Regulating or Shutdown CEA NOT inserted.

[CA] If Reactor did NOT trip, ESTABLISH Reactivity Control by performing step a, b, c or d: [Step 1.1]

Manually Trip Reactor to meet Core Design Criteria of Lowering Reactor Power CRITICAL TASK and Negative Startup Rate to Verify Reactivity Control Established During STATEMENT ATWS Event.

CRITICAL TASK ATCO * [CA] Manually TRIP Reactor at CB-4. [Step 1.1.a]

VERIFY Reactor Power is LOWERING.

VERIFY Startup Rate is NEGATIVE.

NRC Simulator Scenario 4 Outline Rev 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 5, 6, & 7 Page 27 of 36 Event

Description:

RCP Trip / Automatic Reactor Trip Failure / Instrument Air Compressors Trip / Bearing Cooling Water Pump Trip / Steam Line Break inside Containment Time Position Applicants Actions or Behavior Examiner Note: An Emergency Boration is performed once the cooldown is recognized.

DETERMINE an uncontrolled RCS Cooldown in progress. [Step 1.b]

[CA] PERFORM Emergency Boration with uncontrolled cooldown in progress. [Step 1.2]

[CA] ENSURE both following valves CLOSED: [Step 1.2.a]

[CA] FCV-269X, Demin Water Makeup Valve

[CA] FCV-269Y, Boric Acid Makeup Valve

[CA] OPEN all the following valves: [Step 1.2.b]

[CA] HCV-268, Boric Acid Pump Header to Charging Pumps Isolation Valve

[CA] HCV-265, CH-11A Gravity Feed Valve

[CA] HCV-258, CH-11B Gravity Feed Valve

[CA] START all the following pumps: [Step 1.2.c]

[CA] Boric Acid Pump CH-4A

[CA] Boric Acid Pump CH-4B

[CA] Charging Pump CH-1A (inoperable)

[CA] Charging Pump CH-1B (running)

[CA] Charging Pump CH-1C (unavailable)

[CA] CLOSE LCV-218-2, VCT Outlet Valve. [Step 1.2.d]

[CA] ENSURE the following valves CLOSED: [Step 1.2.e]

[CA] LCV-218-3, Charging Pump Suction SIRWT Isolation Valve

[CA] HCV-257, CH-4B Recirc Valve

[CA] HCV-264, CH-4A Recirc Valve

[CA] BORATE until adequate Shutdown Margin is established.

[Step 1.2.f]

CRS DETERMINE Reactivity Control criteria SATISFIED.

BOPO VERIFY Turbine Trip: [Step 2]

VERIFY HP & LP Stop and Intercept Valves CLOSED.

NRC Simulator Scenario 4 Outline Rev 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 5, 6, & 7 Page 28 of 36 Event

Description:

RCP Trip / Automatic Reactor Trip Failure / Instrument Air Compressors Trip / Bearing Cooling Water Pump Trip / Steam Line Break inside Containment Time Position Applicants Actions or Behavior Examiner Note: The Generator Output Breakers are CLOSED due to back feeding.

BOPO ENSURE all Generator Breakers are tripped: [Step 3]

DETERMINE Generator Output Breaker 3451-4 CLOSED.

DETERMINE Generator Output Breaker 3451-5 CLOSED.

DETERMINE Generator Field Breaker 41E/G1F tripped.

BOPO VERIFY 4160 V Safeguards Buses 1A3 and 1A4 are ENERGIZED. [Step 4]

CRS DETERMINE Maintenance of Vital Auxiliaries criteria SATISFIED.

DETERMINE Safety Injection Actuation Signal has NOT occurred and both BOPO Diesel Generators are STOPPED. [Step 5]

VERIFY 4160 V Non-Safeguards Buses 1A1 and 1A2 are ENERGIZED.

BOPO

[Step 6]

BOPO VERIFY 125 VDC Buses 1 and 2 are ENERGIZED. [Step 7]

BOPO VERIFY Instrument Air is AVAILABLE: [Step 8]

DETERMINE Instrument Air pressure < 90 psig.

DETERMINE Instrument Air Compressors NOT RUNNING.

[CA] If Instrument Air pressure is < 90 psig, PERFORM the following to restore Instrument Air: [Step 8.1]

BOPO * [CA] START Bearing Water Pump AC-9A.

BOPO * [CA] START Air Compressor CA-1A.

ATCO VERIFY Component Cooling Water System operation NORMAL: [Step 9]

DETERMINE at least one CCW pump RUNNING. [Step 9.a]

DETERMINE CCW Pump discharge pressure 60 psig. [Step 9.b]

NRC Simulator Scenario 4 Outline Rev 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 5, 6, & 7 Page 29 of 36 Event

Description:

RCP Trip / Automatic Reactor Trip Failure / Instrument Air Compressors Trip / Bearing Cooling Water Pump Trip / Steam Line Break inside Containment Time Position Applicants Actions or Behavior

DETERMINE HCV-438A/B/C/D, CCW to RCP Coolers OPEN. [Step 9.c]

DETERMINE at least one Raw Water Pump RUNNING. [Step 9.d]

CRS VERIFY RCS Inventory Control criteria satisfied: [Step 10]

DETERMINE PZR level between 30% and 70% and NOT TRENDING to ATCO between 45% and 60%.

[CA] RESTORE Inventory Control by manually controlling Charging and Letdown. [Step 10.1.a]

DETERMINE RCS subcooling > 20°F.

CRS DETERMINE RCS Inventory Control criteria NOT SATISFIED.

CRS VERIFY RCS Pressure Control criteria satisfied: [Step 11]

ATCO

DETERMINE RCS pressure between 1800 psia and 2300 psia.

DETERMINE RCS pressure NOT TRENDING between 2050 psia and 2150 psia.

[CA] MANUALLY CONTROL PZR Heaters and Spray to restore RCS pressure.

DETERMINE PORVs are CLOSED.

CRS DETERMINE RCS Pressure Control criteria NOT SATISFIED.

CRS VERIFY Core Heat Removal criteria satisfied: [Step 12]

ATCO

DETERMINE RCP NPSH requirements met.

DETERMINE at least one RCP operating.

DETERMINE Core T 10°F.

CRS DETERMINE Core Heat Removal criteria SATISFIED.

CRS VERIFY RCS Heat Removal criteria satisfied:

NRC Simulator Scenario 4 Outline Rev 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 5, 6, & 7 Page 30 of 36 Event

Description:

RCP Trip / Automatic Reactor Trip Failure / Instrument Air Compressors Trip / Bearing Cooling Water Pump Trip / Steam Line Break inside Containment Time Position Applicants Actions or Behavior VERIFY Main Feedwater is restoring SG levels to 35% to 80% NR and 73%

BOPO to 94% WR. [Step 13]

DETERMINE FCV-1101 & FCV-1102 Feed Regulating Valves CLOSED.

[Step 13.a]

DETERMINE HCV-1105 & HCV-1106 Feed Regulating Bypass Valves ramped to between 40% & 45% OPEN. [Step 13.b]

BOPO

PLACE 43/FW switch in OFF. [Step 13.c]

ENSURE no more than one Main Feedwater Pump RUNNING.

[Step 13.d]

ENSURE no more than one Condensate Pump RUNNING. [Step 13.e]

STOP running Heater Drain Pumps FW-5A, FW-5B, and/or FW-5C.

[Step 13.f]

BOPO

ENSURE SG Blowdown Isolation Valves CLOSED. [Step 13.g]

HCV-1387A & HCV-1387B

HCV-1388A & HCV-1388B Booth Operator: When contacted, REPORT Air Assisted Main Steam Safety Valves CLOSED.

VERIFY Steam Dump and Bypass Valves controlling both of the following:

BOPO

[Step 14]

DETERMINE RCS TCOLD NOT between 525°F and 535°F.

DETERMINE Steam Generator RC-2A pressure < 700 psia.

[CA] If TCOLD less than 525°F, PERFORM the following: [Step 14.1]

BOPO * [CA] CLOSE Steam Dump and Bypass Valves. [Step 14.1.a]

[CA] VERIFY HCV-1040, Atmospheric Dump Valve CLOSED.

[Step 14.1.b]

[CA] CHECK Air Assisted Main Steam Safety Valves CLOSED.

[Step 14.1.c]

[CA] If Steam Generator pressure < 700 psia, ISOLATE Main Steam Header. [Step 14.1.d]

[CA] CLOSE HCV-1041A, MSIV. [Step 14.1.d.1)]

[CA] CLOSE HCV-1042A, MSIV. [Step 14.1.d.1)]

[CA] VERIFY HCV-1041A, MSIV Bypass CLOSED.

[Step 14.1.d.2)]

NRC Simulator Scenario 4 Outline Rev 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 5, 6, & 7 Page 31 of 36 Event

Description:

RCP Trip / Automatic Reactor Trip Failure / Instrument Air Compressors Trip / Bearing Cooling Water Pump Trip / Steam Line Break inside Containment Time Position Applicants Actions or Behavior

[CA] VERIFY CLOSE HCV-1042A, MSIV Bypass CLOSED.

[Step 14.1.d.2)]

[CA] DETERMINE Steam Generator pressure > 500 psia.

[Step 14.1.e]

CRS DETERMINE RCS Heat Removal criteria NOT SATISFIED.

CRS VERIFY Containment Isolation criteria satisfied:

ATCO VERIFY Normal Containment conditions exist: [Step 15]

DETERMINE rise in Containment Sump level in progress. [Step 15.a]

DETERMINE Containment Area Radiation Monitors NOT in alarm.

[Step 15.b]

DETERMINE Containment Ventilation Radiation Monitors NOT in alarm.

[Step 15.c]

DETERMINE SG Blowdown and Condenser Off Gas Radiation Monitors ATCO NOT alarming. [Step 15.d]

DETERMINE SG Blowdown and Condenser Off Gas Radiation Monitors ATCO NOT TRENDING to alarm. [Step 15.e]

ATCO

VERIFY Containment conditions: [Step 15.f]

DETERMINE Containment pressure > 3 psig.

DETERMINE Containment temperature > 120°F.

[CA] INITIATE Containment Cooling. [Step 15.f.1]

ATCO * [CA] ENSURE CCW flow to Containment Vent Fan coils.

[CA] PLACE HCV-402B/D to OPEN.

[CA] PLACE HCV-403B/D to OPEN.

[CA] PLACE HCV-402A/C to OPEN.

[CA] PLACE HCV-403A/C to OPEN.

ATCO * [CA] START all Containment Vent Fans.

[CA] VERIFY Containment Vent Fans VA-3A & VA-3B RUNNING.

[CA] START Containment Vent Fans VA-7C & VA-7D.

[CA] DETERMINE Containment pressure < 5 psig. [Step 15.f.2]

NRC Simulator Scenario 4 Outline Rev 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 5, 6, & 7 Page 32 of 36 Event

Description:

RCP Trip / Automatic Reactor Trip Failure / Instrument Air Compressors Trip / Bearing Cooling Water Pump Trip / Steam Line Break inside Containment Time Position Applicants Actions or Behavior CRS DETERMINE Containment Integrity criteria NOT SATISFIED.

CRS DIAGNOSE event in progress: [Step 16]

DETERMINE Reactivity Control requirements met.

DETERMINE both DC buses energized.

DETERMINE at least one Vital 4160 V Bus energized.

DETERMINE at least one Non-Vital 4160 V Bus energized.

DETERMINE at least one RCP running.

DETERMINE adequate Feedwater flow to at least one SG.

VERIFY Pressurizer pressure > 1800 psia with high subcooled margin, normal SG pressure, and no indications of primary to secondary leakage.

If not, CONSIDER EOP-05, Uncontrolled Heat Extraction.

NOTE Certain events (i.e., LOCA, SGTR, UHE and Loss of All Feedwater) do not require offsite power in order to adequately, mitigate the effects of the accident.

For this reason, the LOCA, SGTR, UHE or Loss of All Feedwater procedure may be implemented even if a Loss of Offsite Power has also occurred.

DETERMINE all Safety Function Acceptance Criteria NOT SATISFIED.

DETERMINE single event in progress and TRANSITION to EOP-05, Uncontrolled Heat Extraction.

NRC Simulator Scenario 4 Outline Rev 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 5, 6, & 7 Page 33 of 36 Event

Description:

RCP Trip / Automatic Reactor Trip Failure / Instrument Air Compressors Trip / Bearing Cooling Water Pump Trip / Steam Line Break inside Containment Time Position Applicants Actions or Behavior Examiner Note: The following steps are from EOP-05, Uncontrolled Heat Extraction.

CRS CONFIRM Standard Post Trip Actions have been performed. [Step 1]

CRS CONFIRM Uncontrolled Heat Extraction Diagnosis: [Step 2]

VERIFY Safety Function Status Check Acceptance Criteria being satisfied. [Step 2.a]

DETERMINE CIAS is NOT present and DIRECT Shift Chemist to SAMPLE both SGs for activity. [Step 2.c]

CRS IMPLEMENT the Emergency Plan. [Step 3]

Time: __________

CREW MONITOR the Floating Steps. [Step 4]

DETERMINE RCS pressure > 1600 psia, Containment pressure < 5 psig, CRS with Steam Generator 500 psia. [Step 5]

BOPO

ENSURE SGIS closes all the following valves: [Step 5.d]

DETERMINE HCV-1041A, RC-2A MSIV CLOSED.

DETERMINE HCV-1041C, RC-2A MSIV Bypass Valve CLOSED.

DETERMINE HCV-1042A, RC-2B MSIV CLOSED.

DETERMINE HCV-1042C, RC-2B MSIV Bypass Valve CLOSED.

DETERMINE HCV-1105, RC-2A Feed Regulating Bypass Valve CLOSED.

DETERMINE HCV-1106, RC-2B Feed Regulating Bypass Valve CLOSED.

DETERMINE HCV-1386, RC-2A Feed Header Isolation Valve CLOSED.

DETERMINE HCV-1385, RC-2B Feed Header Isolation Valve CLOSED.

DETERMINE HCV-1103, RC-2A Feed Regulating Block Valve CLOSED.

DETERMINE HCV-1104, RC-2B Feed Regulating Block Valve CLOSED.

NRC Simulator Scenario 4 Outline Rev 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 5, 6, & 7 Page 34 of 36 Event

Description:

RCP Trip / Automatic Reactor Trip Failure / Instrument Air Compressors Trip / Bearing Cooling Water Pump Trip / Steam Line Break inside Containment Time Position Applicants Actions or Behavior CRS DETERMINE RCS pressure 1600 psia. [Step 6]

CRS DETERMINE Containment pressure < 5 psig. [Step 7]

CRS DETERMINE SIAS has NOT actuated. [Step 8]

ATCO VERIFY RCP operating parameters: [Step 9]

DETERMINE RCP RC-3A TRIPPED and TCOLD < 500°F. [Step 9.a]

DETERMINE RCS pressure ~1900 psia. [Step 9.b]

DETERMINE RCPs subcooling > 20°F. [Step 9.c]

ATCO VERIFY normal CCW/RW System operation: [Step 10]

DETERMINE at least 2 CCW Pumps are RUNNING. [Step 10.a]

DETERMINE CCW Pump discharge pressure 60 psig. [Step 10.b]

ENSURE at least two Raw Water Pumps operating. [Step 10.c]

ATCO

START at least one Raw Water Pump.

DETERMINE at least three RW/CCW Heat Exchangers in service.

[Step 10.d]

DETERMINE all RCP cooler CCW Valves OPEN. [Step 10.e]

CRS DETERMINE affected SG is RC-2A and SG pressure is < 700 psia. [Step 11]

CRS DETERMINE Uncontrolled Heat Extraction has NOT been isolated. [Step 12]

[CA] DETERMINE Emergency Boration already in progress. [Step 12.1]

BOPO DETERMINE SG RC-2A < 500 psia and SG RC-2B > 500 psia. [Step 13]

BOPO DETERMINE Steam Generator RC-2A is most affected SG. [Step 14]

CRS DETERMINE Uncontrolled Heat Extraction has NOT been isolated. [Step 15]

NRC Simulator Scenario 4 Outline Rev 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 5, 6, & 7 Page 35 of 36 Event

Description:

RCP Trip / Automatic Reactor Trip Failure / Instrument Air Compressors Trip / Bearing Cooling Water Pump Trip / Steam Line Break inside Containment Time Position Applicants Actions or Behavior IF RC-2A is most affected, ISOLATE RC-2A by performing HR-19, CRS Isolate/Restore Steam Generator A. [Step 16]

Examiner Note: The following steps are from HR-19, Isolate/Restore Steam Generator A.

CRITICAL TASK Isolate the Affected Steam Generator to Prevent Excess Plant Cooldown and STATEMENT Reactivity Additions Prior to Steam Generator Level = 0% Wide Range Level.

CRITICAL TASK BOPO PERFORM the following to isolate Steam Generator RC-2A: [Step 1]

ENSURE all the following valves are CLOSED: [Step 1.a]

VERIFY HCV-1041A, RC-2A MSIV CLOSED.

VERIFY HCV-1041C, RC-2A MSIV Bypass Valve CLOSED.

VERIFY MS-291, Air Assisted Main Steam Safety Valve CLOSED.

VERIFY FCV-1101, RC-2A Feed Regulating Valve CLOSED.

VERIFY HCV-1105, Feed Regulating Bypass Valve CLOSED.

BOPO

VERIFY HCV-1386, RC-2A Feed Header Isolation Valve CLOSED.

VERIFY HCV-1103, Feed Regulating Block Valve CLOSED.

VERIFY HCV-1388A, Blowdown Isolation Valve CLOSED.

VERIFY HCV-1388B, Blowdown Isolation Valve CLOSED.

CLOSE HCV-1107A, AFW Isolation Valve.

CLOSE HCV-1107B, AFW Isolation Valve.

CONTACT Auxiliary Operator to CLOSE MS-298, Steam Valves HCV-1041A & 1042A Packing Leakoff Line Isolation Valve in Room 81.

[Step 1.b]

If sampling is NOT in progress, CLOSE both Sample Valves: [Step 1.c]

HCV-2506A, RC-2A Blowdown Sample Isolation Valve

HCV-2506B, RC-2B Blowdown Sample Isolation Valve

PERFORM the following to CLOSE YCV-1045A: [Step 1.d]

PLACE ISOLATION VALVE YCV-1045A OVERRIDE SW in BOPO OVERRIDE. [Step 1.d.1)]

PLACE control switch for S/G RC-2A STM TO FW-10 HDR A BOPO ISOLATION VALVE YCV-1045A in CLOSE. [Step 1.d.2)]

NRC Simulator Scenario 4 Outline Rev 6

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 4 Event # 5, 6, & 7 Page 36 of 36 Event

Description:

RCP Trip / Automatic Reactor Trip Failure / Instrument Air Compressors Trip / Bearing Cooling Water Pump Trip / Steam Line Break inside Containment Time Position Applicants Actions or Behavior

CONTACT Auxiliary Operator HANDJACK YCV-1045A, MAIN STEAM LINE "A" TO AUX FEEDWATER PUMP FW-10 SUPPLY VALVE to CLOSE in Room 81. [Step 1.d.3)]

VERIFY RC-2A is most affected SG per Attachment HR-18, Most Affected CRS Steam Generator Determination. [Step 2]

When Steam Generator RC-2A is isolated, TERMINATE the scenario.

NRC Simulator Scenario 4 Outline Rev 6

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

Title:

Classify an Emergency Plan Event Examinee (Print):

Testing Method:

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

Alternate Path: Plant:

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

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

Initial Conditions: Given the following conditions:

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

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

Aircraft arrival time is estimated at 60 minutes.

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

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

Emergency Classification Level __________

IC/EAL Classification __________

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

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

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

Validation Time: 5 minutes Completion Time: ________ minutes Critical Time limit: 15 minutes Comments:

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

Page 1 of 4 NRC Admin JPM SA5 Rev. 4

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

PROVIDE the examinee with a copy of:

EPIP-OSC-1, Emergency Plan.

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

NOTE:

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

Page 2 of 4 NRC Admin JPM SA5 Rev. 4

Appendix C JPM STEPS Form ES-C-1

- Check Mark Denotes Critical Step CRITICAL START TIME:

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

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

Perform Step: 1 DETERMINE the Event Category.

Standard: REFERRED to FCS Emergency Action Levels:

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

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

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

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

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

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

Comment: SAT UNSAT Examiner Note: Declaration shall be made within 15 minutes of start time of JPM.

Perform Step: 4 Classify the event.

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

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

Terminating Cue: This JPM is complete.

Comment: SAT UNSAT CRITICAL STOP TIME:

Page 3 of 4 NRC Admin JPM SA5 Rev. 4

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

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

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

Aircraft arrival time is estimated at 60 minutes.

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

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

Emergency Classification Level __________

IC/EAL Classification __________

THIS IS A TIME CRITICAL JPM Page 4 of 4 NRC Admin JPM SA5 Rev. 4

Appendix D Scenario Outline Form ES-D-1 Facility: Fort Calhoun Station Scenario No.: 1 Op Test No.: Dec 2015 NRC Examiners: Operators:

Initial Conditions: 100% power MOL - RCS Boron is 482 ppm (by sample).

Turnover: Maintain steady-state power conditions. Chemistry requests two Charging Pumps be placed in service per OI-CH-1, CVCS System Normal Operation.

Critical Tasks:

Manually Actuate Pressurizer Pressure Low Signal (PPLS) when RCS Pressure 1600 psia and before 1350 psia to ensure SIAS / VIAS / CIAS Activation. (Event 8)

Stop All Reactor Coolant Pumps (RCPs) when Subcooling is less than 20°F due to Loss of Net Positive Suction Head (NPSH) per RCP NPSH Curve. (Event 6)

Commence a Cooldown and Depressurization of the Reactor Coolant System to Reestablish RCS Inventory Control while maintaining RCS Heat Removal. (Event 6)

Event No. Malf. No. Event Type* Event Description 1 N (ATCO) Raise Charging and Letdown Flow per OI-CH-1, CVCS System

+15 min Normal Operation, Attachment 3.

2 C (ATCO, CRS) Component Cooling Water (CCW) Pump Trip.

+25 min TS (CRS) Start Either Standby CCW Pump.

3 C (BOPO, CRS) Plant Air System Leak.

+35 min Start Instrument Air Compressors.

4 I (ATCO, CRS) Pressurizer Pressure Control Channel PT-103X Fails to 2150 psia

+45 min TS (CRS) on 15 Minute Ramp. Transfer Pressure Control to PT-103Y.

5 R (ATCO) Condenser Evacuation Pump Trip with Auto Start Failure.

+55 min C (BOPO, CRS) Partial Loss of Condenser Vacuum. Reduce Turbine Load.

6 M (ATCO, BOPO, Inadvertent Main Turbine Trip.

+55 min CRS) Pressurizer Safety Valve Fails 50% Open on Reactor Trip.

7 C (BOPO) Total Loss of Condenser Vacuum.

+55 min Place HCV-1040, Atmospheric Dump Valve in Service.

8 I (ATCO) Pressurizer Pressure Low Signal Actuation Failure.

+65 min Manually Initiate Safety Injection.

9 C (ATCO) Low Pressure Safety Injection (LPSI) Pumps Start Failure.

+65 min Manually Start LPSI Pumps.

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Technical Specifications Actual Target Quantitative Attributes 3 Malfunctions after EOP entry (1-2) 5 Abnormal events (2-4) 1 Major transients (1-2) 1 EOPs entered/requiring substantive actions (1-2) 0 EOP contingencies requiring substantive actions (0-2) 3 Critical tasks (2-3)